Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, Richard L.; Folsom, Charles Pearson; Pastore, Giovanni

2016-05-01

One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag

2012-04-01

The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather thanmore » graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.« less

Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan

NASA Astrophysics Data System (ADS)

Fujimura, Koji; Sasahira, Akira; Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi

According to the Japan's Framework for Nuclear Energy Policy(1), a basic scenario for fast breeder reactors (FBRs) is that they will be introduced on a commercial basis starting around 2050 replacing light water reactors (LWRs). During the FBR introduction period, the Pu from LWR spent fuel is used for FBR startup. Howerver, the FBR core loaded with this Pu has a larger burnup reactivity due to its larger isotopic content of Pu-241 than a core loaded with Pu from an FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of an FBR. We investigated, an FBR transitional core concept to confront the issues of the FBR introductory period in Japan. Core specifications are based on the compact-type sodium-cooled mixed oxide (MOX)-fueled core designed from the Japanese FBR cycle feasibility studies, because lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up FBRs. The reference specifications were selected as follows. Output of 1500MWe and average discharge fuel burnup of about 150GWd/t. Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBRs are loaded to the initial loading fuels and exchanged fuels during few cycles until equilibrium. We made the MA content of the initial loading fuel four kinds like 0%, 3%, 4%, 5%. The average of the initial loading fuel is assumed to be 3%, and that of the exchange fuel is set as 5%. This 5% maximum of the MA content is based on the irradiation results of the experimental fast reactor Joyo. We evaluated the core performances including burnup characteristics and the reactivity coefficient and confirmed that transitional core from initial loading until equilibrium cycle with loaded Pu from LWR spent fuel performs similary to an FBR multi-recycling core.

Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

DOE Office of Scientific and Technical Information (OSTI.GOV)

Katoh, Yutai; Terrani, Kurt A.

2015-08-01

Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

Current and anticipated uses of thermal-hydraulic codes in NFI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsuda, K.; Takayasu, M.

1997-07-01

This paper presents the thermal-hydraulic codes currently used in NFI for the LWR fuel development and licensing application including transient and design basis accident analyses of LWR plants. The current status of the codes are described in the context of code capability, modeling feature, and experience of code application related to the fuel development and licensing. Finally, the anticipated use of the future thermal-hydraulic code in NFI is briefly given.

TREAT Neutronics Analysis of Water-Loop Concept Accommodating LWR 9-rod Bundle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hill, Connie M.; Woolstenhulme, Nicolas E.; Parry, James R.

Abstract. Simulation of a variety of transient conditions has been successfully achieved in the Transient Reactor Test (TREAT) facility during operation between 1959 and 1994 to support characterization and safety analysis of nuclear fuels and materials. A majority of previously conducted tests were focused on supporting sodium-cooled fast reactor (SFR) designs. Experiments evolved in complexity. Simulation of thermal-hydraulic conditions expected to be encountered by fuels and materials in a reactor environment was realized in the development of TREAT sodium loop experiment vehicles. These loops accommodated up to 7-pin fuel bundles and served to simulate more closely the reactor environment whilemore » safely delivering large quantities of energy into the test specimen. Some of the immediate TREAT restart operations will be focused on testing light water reactor (LWR) accident tolerant fuels (ATF). Similar to the sodium loop objectives, a water loop concept, developed and analyzed in the 1990’s, aimed at achieving thermal-hydraulic conditions encountered in commercial power reactors. The historic water loop concept has been analyzed in the context of a reactivity insertion accident (RIA) simulation for high burnup LWR 2-pin and 3-pin fuel bundles. Findings showed sufficient energy could be deposited into the specimens for evaluation. Similar results of experimental feasibility for the water loop concept (past and present) have recently been obtained using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. The old water loop concept required only two central TREAT core grid spaces. Preparation for future experiments has resulted in a modified water loop conceptual design designated the TREAT water environment recirculating loop (TWERL). The current TWERL design requires nine TREAT core grid spaces in order to place the water recirculating pump under the TREAT core. Due to the effectiveness of water moderation, neutronics analysis shows that removal of seven additional TREAT fuel elements to facilitate the experiment will not inhibit the ability to successfully simulate a RIA for the 2-pin or 3-pin bundle. This new water loop design leaves room for accommodating a larger fuel pin bundle than previously analyzed. The 7-pin fuel bundle in a hexagonal array with similar spacing of fuel pins in a SFR fuel assembly was considered the minimum needed for one central fuel pin to encounter the most correct thermal conditions. The 9-rod fuel bundle in a square array similar in spacing to pins in a LWR fuel assembly would be considered the LWR equivalent. MCNP analysis conducted on a preliminary LWR 9-rod bundle design shows that sufficient energy deposition into the central pin can be achieved well within range to investigate fuel and cladding performance in a simulated RIA. This is achieved by surrounding the flow channel with an additional annulus of water. Findings also show that a highly significant increase in TREAT to specimen power coupling factor (PCF) within the central pin can be achieved by surrounding the experiment with one to two rings of TREAT upgrade fuel assemblies. The experiment design holds promise for the performance evaluation of PWR fuel at extremely high burnup under similar reactor environment conditions.« less

Modeling and analysis of UN TRISO fuel for LWR application using the PARFUME code

NASA Astrophysics Data System (ADS)

Collin, Blaise P.

2014-08-01

The Idaho National Laboratory (INL) PARFUME (PARticle FUel ModEl) code was used to assess the overall fuel performance of uranium nitride (UN) tristructural isotropic (TRISO) ceramic fuel under irradiation conditions typical of a Light Water Reactor (LWR). The dimensional changes of the fuel particle layers and kernel were calculated, including the formation of an internal gap. The survivability of the UN TRISO particle was estimated depending on the strain behavior of the constituent materials at high fast fluence and burn-up. For nominal cases, internal gas pressure and representative thermal profiles across the kernel and layers were determined along with stress levels in the inner and outer pyrolytic carbon (IPyC/OPyC) and silicon carbide (SiC) layers. These parameters were then used to evaluate fuel particle failure probabilities. Results of the study show that the survivability of UN TRISO fuel under LWR irradiation conditions might only be guaranteed if the kernel and PyC swelling rates are limited at high fast fluence and burn-up. These material properties have large uncertainties at the irradiation levels expected to be reached by UN TRISO fuel in LWRs. Therefore, a large experimental effort would be needed to establish material properties, including kernel and PyC swelling rates, under these conditions before definitive conclusions can be drawn on the behavior of UN TRISO fuel in LWRs.

Irradiation effects on thermal properties of LWR hydride fuel

NASA Astrophysics Data System (ADS)

Terrani, Kurt; Balooch, Mehdi; Carpenter, David; Kohse, Gordon; Keiser, Dennis; Meyer, Mitchell; Olander, Donald

2017-04-01

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

Power ramp induced iodine and cesium redistribution in LWR fuel rods

NASA Astrophysics Data System (ADS)

Sontheimer, F.; Vogl, W.; Ruyter, I.; Markgraf, J.

1980-01-01

Volatile fission product migration in LWR fuel rods which are power ramped above a certain threshold beyond the envelope of their previous power history, plays an important role in stress corrosion cracking of Zircaloy. This may cause fuel rods to fail already at stresses below the yield strength. In the HFR, Petten, many power ramp experiments have been performed with subsequent examination of the ramped rods for fission product distribution. This study describes the measurement of iodine and cesium distribution using γ-spectroscopy of I-131 and Cs-137. An evaluation method is presented which makes the determination of absolute amounts of I/Cs feasible. It is shown that a threshold for I/Cs redistribution exists beyond which it depends strongly on local fuel rod power and fuel type.

Early implementation of SiC cladding fuel performance models in BISON

DOE Office of Scientific and Technical Information (OSTI.GOV)

Powers, Jeffrey J.

2015-09-18

SiC-based ceramic matrix composites (CMCs) [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation duemore » to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.« less

Safety and Regulatory Issues of the Thorium Fuel Cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian; Worrall, Andrew; Powers, Jeffrey

2014-02-01

Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2),more » add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.« less

Converting Maturing Nuclear Sites to Integrated Power Production Islands

DOE PAGES

Solbrig, Charles W.

2011-01-01

Nuclear islands, which are integrated power production sites, could effectively sequester and safeguard the US stockpile of plutonium. A nuclear island, an evolution of the integral fast reactor, utilizes all the Transuranics (Pu plus minor actinides) produced in power production, and it eliminates all spent fuel shipments to and from the site. This latter attribute requires that fuel reprocessing occur on each site and that fast reactors be built on-site to utilize the TRU. All commercial spent fuel shipments could be eliminated by converting all LWR nuclear power sites to nuclear islands. Existing LWR sites have the added advantage ofmore » already possessing a license to produce nuclear power. Each could contribute to an increase in the nuclear power production by adding one or more fast reactors. Both the TRU and the depleted uranium obtained in reprocessing would be used on-site for fast fuel manufacture. Only fission products would be shipped to a repository for storage. The nuclear island concept could be used to alleviate the strain of LWR plant sites currently approaching or exceeding their spent fuel pool storage capacity. Fast reactor breeding ratio could be designed to convert existing sites to all fast reactors, or keep the majority thermal.« less

The fractalline properties of experimentally simulated PWR fuel crud

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Mishra, V. K.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Mota, G.; Hussey, D.; Wang, G.; Byers, W. A.; Short, M. P.

2018-02-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud may cause serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each makes critical assumptions regarding how to treat the complex, porous microstructure of crud and its resultant effects on temperature, pressure, and crud chemistry. In this study, we demonstrate that crud is indeed a fractalline porous medium using flowing loop experiments, validating the most recent models of its effects on LWR fuel cladding. This crud is shown to match that in other LWR-prototypical facilities through a porosity-fractal dimension scaling law. Implications of this result range from post-mortem analysis of the effects of crud on reactor fuel performance, to utilizing crud's fractalline dimensions to quantify the effectiveness of anti-fouling measures.

Characteristics of potential repository wastes. Volume 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-07-01

The LWR spent fuels discussed in Volume 1 of this report comprise about 99% of all domestic non-reprocessed spent fuel. In this report we discuss other types of spent fuels which, although small in relative quantity, consist of a number of diverse types, sizes, and compositions. Many of these fuels are candidates for repository disposal. Some non-LWR spent fuels are currently reprocessed or are scheduled for reprocessing in DOE facilities at the Savannah River Site, Hanford Site, and the Idaho National Engineering Laboratory. It appears likely that the reprocessing of fuels that have been reprocessed in the past will continuemore » and that the resulting high-level wastes will become part of defense HLW. However, it is not entirely clear in some cases whether a given fuel will be reprocessed, especially in cases where pretreatment may be needed before reprocessing, or where the enrichment is not high enough to make reprocessing attractive. Some fuels may be canistered, while others may require special means of disposal. The major categories covered in this chapter include HTGR spent fuel from the Fort St. Vrain and Peach Bottom-1 reactors, research and test reactor fuels, and miscellaneous fuels, and wastes generated from the decommissioning of facilities.« less

Validating the BISON fuel performance code to integral LWR experiments

DOE PAGES

Williamson, R. L.; Gamble, K. A.; Perez, D. M.; ...

2016-03-24

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to datemore » for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Our results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. In the initial rod diameter comparisons they were unsatisfactory and have lead to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hellesen, C.; Grape, S.; Haakanson, A.

2013-07-01

Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miao, Yinbin; Ye, Bei; Mei, Zhigang

Uranium silicide (U 3Si 2) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U 3Si 2, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U 3Si 2 at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuelmore » material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U 3Si 2 in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U 3Si 2 at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U 3Si 2 fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U 3Si 2 fuel as an accident-tolerant alternative for uranium dioxide.« less

Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

NASA Astrophysics Data System (ADS)

Permana, Sidik; Saputra, Geby; Suzuki, Mitsutoshi; Saito, Masaki

2017-01-01

Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

New approaches for MOX multi-recycling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gain, T.; Bouvier, E.; Grosman, R.

Due to its low fissile content after irradiation, Pu from used MOX fuel is considered by some as not recyclable in LWR (Light Water Reactors). The point of this paper is hence to go back to those statements and provide a new analysis based on AREVA extended experience in the fields of fissile and fertile material management and optimized waste management. This is done using the current US fuel inventory as a case study. MOX Multi-recycling in LWRs is a closed cycle scenario where U and Pu management through reprocessing and recycling leads to a significant reduction of the usedmore » assemblies to be stored. The recycling of Pu in MOX fuel is moreover a way to maintain the self-protection of the Pu-bearing assemblies. With this scenario, Pu content is also reduced repetitively via a multi-recycling of MOX in LWRs. Simultaneously, {sup 238}Pu content decreases. All along this scenario, HLW (High-Level Radioactive Waste) vitrified canisters are produced and planned for deep geological disposal. Contrary to used fuel, HLW vitrified canisters do not contain proliferation materials. Moreover, the reprocessing of used fuel limits the space needed on current interim storage. With MOX multi-recycling in LWR, Pu isotopy needs to be managed carefully all along the scenario. The early introduction of a limited number of SFRs (Sodium Fast Reactors) can therefore be a real asset for the overall system. A few SFRs would be enough to improve the Pu isotopy from used LWR MOX fuel and provide a Pu-isotopy that could be mixed back with multi-recycled Pu from LWRs, hence increasing the Pu multi-recycling potential in LWRs.« less

Initial experimental evaluation of crud-resistant materials for light water reactors

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Carlson, M. B.; Hussey, D.; Short, M. P.

2018-01-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud causes serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each assumes that it will always be present. In this study, we report on the development of crud-resistant materials as fuel cladding coatings, to reduce or eliminate these problems altogether. Integrated loop testing experiments at flowing LWR conditions show significantly reduced crud adhesion and surface crud coverage, respectively, for TiC and ZrN coatings compared to ZrO2. The loop testing results roughly agree with the London dispersion component of van der Waals force predictions, suggesting that they contribute most significantly to the adhesion of crud to fuel cladding in out-of-pile conditions. These results motivate a new look at ways of reducing crud, thus avoiding many expensive LWR operational issues.

Sustained Recycle in Light Water and Sodium-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven J. Piet; Samuel E. Bays; Michael A. Pope

2010-11-01

From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in freshmore » fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.« less

An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bathke, Charles Gary; Wallace, Richard K; Hase, Kevin R

2010-01-01

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled untilmore » consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blaise Collin

The Idaho National Laboraroty (INL) PARFUME (particle fuel model) code was used to assess the overall fuel performance of uranium nitride (UN) tristructural isotropic (TRISO) ceramic fuel under irradiation conditions typical of a Light Water Reactor (LWR). The dimensional changes of the fuel particle layers and kernel were calculated, including the formation of an internal gap. The survivability of the UN TRISO particle was estimated depending on the strain behavior of the constituent materials at high fast fluence and burn up. For nominal cases, internal gas pressure and representative thermal profiles across the kernel and layers were determined along withmore » stress levels in the inner and outer pyrolytic carbon (IPyC/OPyC) and silicon carbide (SiC) layers. These parameters were then used to evaluate fuel particle failure probabilities. Results of the study show that the survivability of UN TRISO fuel under LWR irradiation conditions might only be guaranteed if the kernel and PyC swelling rates are limited at high fast fluence and burn up. These material properties have large uncertainties at the irradiation levels expected to be reached by UN TRISO fuel in LWRs. Therefore, a large experimental effort would be needed to establish material properties, including kernel and PyC swelling rates, under these conditions before definitive conclusions can be drawn on the behavior of UN TRISO fuel in LWRs.« less

Advanced Fuels Campaign FY 2015 Accomplishments Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Braase, Lori Ann; Carmack, William Jonathan

2015-10-29

The mission of the Advanced Fuels Campaign (AFC) is to perform research, development, and demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This report is a compilation of technical accomplishment summaries for FY-15. Emphasis is on advanced accident-tolerant LWR fuel systems, advanced transmutation fuels technologies, and capability development.

Advanced multiphysics coupling for LWR fuel performance analysis

DOE PAGES

Hales, J. D.; Tonks, M. R.; Gleicher, F. N.; ...

2015-10-01

Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics,more » particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is possible to use lower length scale models such as those used in the mesoscale MARMOT code to compute average properties, e.g. swelling or thermal conductivity. These may then be used by an engineering-scale model. Examples of this type of multiscale, multiphysics modeling are shown.« less

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Jain, Prashant K.; Powers, Jeffrey J.

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments usingmore » equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.« less

Validation of the analytical methods in the LWR code BOXER for gadolinium-loaded fuel pins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paratte, J.M.; Arkuszewski, J.J.; Kamboj, B.K.

1990-01-01

Due to the very high absorption occurring in gadolinium-loaded fuel pins, calculations of lattices with such pins present are a demanding test of the analysis methods in light water reactor (LWR) cell and assembly codes. Considerable effort has, therefore, been devoted to the validation of code methods for gadolinia fuel. The goal of the work reported in this paper is to check the analysis methods in the LWR cell/assembly code BOXER and its associated cross-section processing code ETOBOX, by comparison of BOXER results with those from a very accurate Monte Carlo calculation for a gadolinium benchmark problem. Initial results ofmore » such a comparison have been previously reported. However, the Monte Carlo calculations, done with the MCNP code, were performed at Los Alamos National Laboratory using ENDF/B-V data, while the BOXER calculations were performed at the Paul Scherrer Institute using JEF-1 nuclear data. This difference in the basic nuclear data used for the two calculations, caused by the restricted nature of these evaluated data files, led to associated uncertainties in a comparison of the results for methods validation. In the joint investigations at the Georgia Institute of Technology and PSI, such uncertainty in this comparison was eliminated by using ENDF/B-V data for BOXER calculations at Georgia Tech.« less

Closed DTU fuel cycle with Np recycle and waste transmutation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beller, D.E.; Sailor, W.C.; Venneri, F.

1999-09-01

A nuclear energy scenario for the 21st century that included a denatured thorium-uranium-oxide (DTU) fuel cycle and new light water reactors (LWRs) supported by accelerator-driven transmutation of waste (ATW) systems was previously described. This coupled system with the closed DTU fuel cycle provides several improvements beyond conventional LWR (CLWR) (once-through, UO{sub 2} fuel) nuclear technology: increased proliferation resistance, reduced waste, and efficient use of natural resources. However, like CLWR fuel cycles, the spent fuel in the first one-third core discharged after startup contains higher-quality Pu than the equilibrium fuel cycle. To eliminate this high-grade Pu, Np is separated and recycledmore » with Th and U--rather than with higher actinides [(HA) including Pu]. The presence of Np in the LWR feed greatly increases the production of {sup 238}Pu so that a few kilograms of Pu generated enough alpha-decay heat that the separated Pu is highly resistant to proliferation. This alternate process also simplifies the pyrochemical separation of fuel elements (Th and U) from HAs. To examine the advantages of this concept, the authors modeled a US deployment scenario for nuclear energy that includes DTU-LWRs plus ATW`s to burn the actinides produced by these LWRs and to close the back-end of the DTU fuel cycle.« less

Modelling of LOCA Tests with the BISON Fuel Performance Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, Richard L; Pastore, Giovanni; Novascone, Stephen Rhead

2016-05-01

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculationsmore » are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.« less

An Example of an INPRO Assessment of an INS in the Area of Waste Management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Allan, C.; Busurin, Y.; Depisch, F.

2006-07-01

Following a resolution of the General Conference of the IAEA in the year 2000 the International Project on Innovative Nuclear Reactors and Fuel Cycles, referred to as INPRO, was initiated. INPRO has defined requirements organized in a hierarchy of Basic Principles, User Requirements and Criteria (consisting of an indicator and an acceptance limit) to be met by innovative nuclear reactor systems (INS) in six areas, namely: economics, safety, waste management, environment, proliferation resistance, and infrastructure. If an INS meets all requirements in all areas it represents a sustainable system for the supply of energy, capable of making a significant contributionmore » to meeting the energy needs of the 21. century. Draft manuals have been developed, for each INPRO area, to provide guidance for performing an assessment of whether an INS meets the INPRO requirements in a given area. The manuals set out the information that needs to be assembled to perform an assessment and provide guidance on selecting the acceptance limits and, for a given INS, for determining the value of the indicators for comparison with the associated acceptance limits. Each manual also includes an example of a specific assessment to illustrate the guidance. This paper discusses the example presented in the manual for performing an INPRO assessment in the area of waste management. The example, chosen solely for the purpose of illustrating the INPRO methodology, describes an assessment of an INS based on the DUPIC fuel cycle. It is assumed that uranium is mined, milled, converted, enriched, and fabricated into LWR fuel in Canada. The LWR fuel is assumed to be leased to a utility in the USA. The spent LWR fuel is assumed to be returned to Canada where it is processed into CANDU DUPIC fuel, which is then burned in CANDU reactors. The assessment steps and the results are presented in detail in the paper. The example illustrates an assessment performed for an INS at an early stage of development. (authors)« less

Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella J van Rooyen

2012-09-01

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella J van Rooyen

2013-01-01

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

Closed Fuel Cycle Waste Treatment Strategy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vienna, J. D.; Collins, E. D.; Crum, J. V.

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significantmore » additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.« less

Modeling and Analysis of FCM UN TRISO Fuel Using the PARFUME Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blaise Collin

2013-09-01

The PARFUME (PARticle Fuel ModEl) modeling code was used to assess the overall fuel performance of uranium nitride (UN) tri-structural isotropic (TRISO) ceramic fuel in the frame of the design and development of Fully Ceramic Matrix (FCM) fuel. A specific modeling of a TRISO particle with UN kernel was developed with PARFUME, and its behavior was assessed in irradiation conditions typical of a Light Water Reactor (LWR). The calculations were used to access the dimensional changes of the fuel particle layers and kernel, including the formation of an internal gap. The survivability of the UN TRISO particle was estimated dependingmore » on the strain behavior of the constituent materials at high fast fluence and burn-up. For nominal cases, internal gas pressure and representative thermal profiles across the kernel and layers were determined along with stress levels in the pyrolytic carbon (PyC) and silicon carbide (SiC) layers. These parameters were then used to evaluate fuel particle failure probabilities. Results of the study show that the survivability of UN TRISO fuel under LWR irradiation conditions might only be guaranteed if the kernel and PyC swelling rates are limited at high fast fluence and burn-up. These material properties are unknown at the irradiation levels expected to be reached by UN TRISO fuel in LWRs. Therefore, more effort is needed to determine them and positively conclude on the applicability of FCM fuel to LWRs.« less

One-way implodable tag capsule with hemispherical beaded end cap for LWR fuel manufacturing

DOEpatents

Gross, K.; Lambert, J.

1999-04-06

A capsule is disclosed containing a tag gas in a zircaloy body portion having a hemispherical top curved toward the bottom of the body portion. The hemispherical top has a rupturable portion upon exposure to elevated gas pressure and the capsule is positioned within a fuel element in a nuclear reactor. 3 figs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vienna, John D.; Todd, Terry A.; Gray, Kimberly D.

The U.S. Department of Energy, Office of Nuclear Energy has chartered an effort to develop technologies to enable safe and cost effective recycle of commercial used nuclear fuel (UNF) in the U.S. Part of this effort includes the evaluation of exiting waste management technologies for effective treatment of wastes in the context of current U.S. regulations and development of waste forms and processes with significant cost and/or performance benefits over those existing. This study summarizes the results of these ongoing efforts with a focus on the highly radioactive primary waste streams. The primary streams considered and the recommended waste formsmore » include: •Tritium separated from either a low volume gas stream or a high volume water stream. The recommended waste form is low-water cement in high integrity containers. •Iodine-129 separated from off-gas streams in aqueous processing. There are a range of potentially suitable waste forms. As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals. •Carbon-14 separated from LWR fuel treatment off-gases and immobilized as a CaCO3 in a cement waste form. •Krypton-85 separated from LWR and SFR fuel treatment off-gases and stored as a compressed gas. •An aqueous reprocessing high-level waste (HLW) raffinate waste which is immobilized by the vitrification process in one of three forms: a single phase borosilicate glass, a borosilicate based glass ceramic, or a multi-phased titanate ceramic [e.g., synthetic rock (Synroc)]. •An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel that is either included in the borosilicate HLW glass or is immobilized in the form of a metal alloy in the case of glass ceramics or titanate ceramics. •Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware that are washed and super-compacted for disposal or as an alternative Zr purification and reuse (or disposal as low-level waste, LLW) by reactive gas separations. •Electrochemical process salt HLW which is immobilized in a glass bonded Sodalite waste form known as the ceramic waste form (CWF). •Electrochemical process UDS and SS cladding hulls which are melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported.« less

Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

NASA Astrophysics Data System (ADS)

Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

2005-02-01

A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

Pyroprocessing of Light Water Reactor Spent Fuels Based on an Electrochemical Reduction Technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ohta, Hirokazu; Inoue, Tadashi; Sakamura, Yoshiharu

A concept of pyroprocessing light water reactor (LWR) spent fuels based on an electrochemical reduction technology is proposed, and the material balance of the processing of mixed oxide (MOX) or high-burnup uranium oxide (UO{sub 2}) spent fuel is evaluated. Furthermore, a burnup analysis for metal fuel fast breeder reactors (FBRs) is conducted on low-decontamination materials recovered by pyroprocessing. In the case of processing MOX spent fuel (40 GWd/t), UO{sub 2} is separately collected for {approx}60 wt% of the spent fuel in advance of the electrochemical reduction step, and the product recovered through the rare earth (RE) removal step, which hasmore » the composition uranium:plutonium:minor actinides:fission products (FPs) = 76.4:18.4:1.7:3.5, can be applied as an ingredient of FBR metal fuel without a further decontamination process. On the other hand, the electroreduced alloy of high-burnup UO{sub 2} spent fuel (48 GWd/t) requires further decontamination of residual FPs by an additional process such as electrorefining even if RE FPs are removed from the alloy because the recovered plutonium (Pu) is accompanied by almost the same amount of FPs in addition to RE. However, the amount of treated materials in the electrorefining step is reduced to {approx}10 wt% of the total spent fuel owing to the prior UO{sub 2} recovery step. These results reveal that the application of electrochemical reduction technology to LWR spent oxide fuel is a promising concept for providing FBR metal fuel by a rationalized process.« less

Light-water-reactor safety research program. Quarterly progress report, July--September 1975

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1975-01-01

Progress is summarized in the following research and development areas: (1) loss-of-coolant accident research; heat transfer and fluid dynamics; (2) transient fuel response and fission-product release; and (3) mechanical properties of Zircaloy containing oxygen. Also included is an appendix on Kinetics of Fission Gas and Volatile Fission-product Behavior under Transient Conditions in LWR Fuel.

Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

NASA Astrophysics Data System (ADS)

Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

2017-04-01

A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

NASA Astrophysics Data System (ADS)

Metzger, Kathryn E.

Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it difficult to predict fuel-cladding mechanical behavior. This information is essential for designing accident tolerant fuel systems where ceramic claddings, like silicon carbide (SiC) are proposed. This research provides a model for both the thermal and irradiation creep behavior for U3Si2. This body of research is comprised of both experimental and modeling components. Characterization of the fuel microstructure includes; optical microscopy with pore and grain size analysis, helium pycnometry for density determination, mercury intrusion porosimetry, compositional analysis in the form of XRD, second phase identification using EDX, electrical resistance measurement via four point probe, determination of hardness and toughness through Vickers indentation testing, and determination of elastic properties using the impulse excitation method. Post-sintering grain size data allowed for the determination of grain boundary activation energy and diffusion coefficients, which were used to develop creep models. This was extended to lattice and irradiation enhanced diffusion in order to develop a U3Si2 creep model over thermal and irradiation creep regimes. In addition to the creep model, thermal and swelling behavior models for U3Si2 were implemented into the BISON fuel performance code. A series of simulations evaluated the performance and behavior of U3Si2 under typical light water reactor conditions with advanced SiC ceramic cladding. Simulation results show that fuel creep relieves stress in the ceramic cladding and postpones the. moment of fuel-clad contact. However, the stress reduction to the cladding is minimal because the fuel creep rate is low while the swelling rate is high. Future work should include the investigation of monolithic U3Si2 irradiation swelling since the current model relies upon the swelling data of U3Si2 particles in a metallic dispersion fuel. Additionally, planned thermal creep testing at the University of South Carolina can provide confirmation of the U3Si2 creep model contained herein.

PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. T. Khericha

2007-04-01

The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed tomore » achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.« less

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. Khericha

2010-12-01

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn thesemore » actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3 to 5 years to construct.« less

Technologies for Upgrading Light Water Reactor Outlet Temperature

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessmentmore » of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.« less

Top Ten Reasons for DEOX as a Front End to Pyroprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; K.J. Bateman; S.D. Herrmann

A front end step is being considered to augment chopping during the treatment of spent oxide fuel by pyroprocessing. The front end step, termed DEOX for its emphasis on decladding via oxidation, employs high temperatures to promote the oxidation of UO2 to U3O8 via an oxygen carrier gas. During oxidation, the spent fuel experiences a 30% increase in lattice structure volume resulting in the separation of fuel from cladding with a reduced particle size. A potential added benefit of DEOX is the removal of fission products, either via direct release from the broken fuel structure or via oxidation and volatilizationmore » by the high temperature process. Fuel element chopping is the baseline operation to prepare spent oxide fuel for an electrolytic reduction step. Typical chopping lengths range from 1 to 5 mm for both individual elements and entire assemblies. During electrolytic reduction, uranium oxide is reduced to metallic uranium via a lithium molten salt. An electrorefining step is then performed to separate a majority of the fission products from the recoverable uranium. Although DEOX is based on a low temperature oxidation cycle near 500oC, additional conditions have been tested to distinguish their effects on the process.[1] Both oxygen and air have been utilized during the oxidation portion followed by vacuum conditions to temperatures as high as 1200oC. In addition, the effects of cladding on fission product removal have also been investigated with released fuel to temperatures greater than 500oC.« less

Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

2012-07-01

A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

Full-scale hot cell test of an acoustic sensor dedicated to measurement of the internal gas pressure and composition of a LWR nuclear fuel rod

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferrandis, J. Y.; Rosenkrantz, E.; Leveque, G.

2011-07-01

A full-scale hot cell test of the internal gas pressure and composition measurement by an acoustic sensor was carried on successfully between 2008 and 2010 on irradiated fuel rods in the LECA-STAR facility at Cadarache Centre. The acoustic sensor has been specially designed in order to provide a nondestructive technique to easily carry out the measurement of the internal gas pressure and gas composition of a LWR nuclear fuel rod. This sensor has been achieved in 2007 and is now covered by an international patent. The first positive result, concerning the device behaviour, is that the sensor-operating characteristics have notmore » been altered by a two-year exposure in the hot cell ambient. We performed the gas characterisation contained in irradiated fuel rods. The acoustic method accuracy is now {+-}5 bars on the pressure measurement result and {+-}0.3% on the evaluated gas composition. The results of the acoustic method were compared to puncture results. Another significant conclusion is that the efficiency of the acoustic method is not altered by the irradiation time, and possible modification of the cladding properties. These results make it possible to demonstrate the feasibility of the technique on irradiated fuel rods. The transducer and the associated methodology are now operational. (authors)« less

Uranium nitride as LWR TRISO fuel: Thermodynamic modeling of U-C-N

NASA Astrophysics Data System (ADS)

Besmann, Theodore M.; Shin, Dongwon; Lindemer, Terrence B.

2012-08-01

TRISO coated particle fuel is envisioned as a next generation replacement for current urania pellet fuel in LWR applications. To obtain adequate fissile loading the kernel of the TRISO particle will likely need to be UN instead of UO2. In support of the necessary development effort for this new fuel system, an assessment of phase regions of interest in the U-C-N system was undertaken as the fuel will be prepared by the carbothermic reduction of the oxide followed by nitriding, will be in equilibrium with carbon within the TRISO particle, and will react with minor actinides and fission products. The phase equilibria and thermochemistry of the U-C-N system is reviewed, including nitrogen pressure measurements above various phase fields. Measurements were used to confirm an ideal solution model of UN and UC adequately represents the UC1-xNx phase. Agreement with the data was significantly improved by effectively adjusting the Gibbs free energy of UN by +12 kJ/mol. This also required adjustment of the value for the sesquinitride by +17 kJ/mol to obtain agreement with phase equilibria. The resultant model together with reported values for other phases in the system was used to generate isothermal sections of the U-C-N phase diagram. Nitrogen partial pressures were also computed for regions of interest.

Fusion Applications and Market Evaluation (FAME) Study

DTIC Science & Technology

1988-02-01

fuel from the breeder. Pyrochemical reprocessing is identified as having the potential for low cost, but needs development . The fast-fission designs... Development Administration, "Alternatives for Man- aging Wastes from Reactors and Post-Fission Operations in the LWR Fuel Cycle," ERDA-76-43 (1976). 5...of the ICF program to produce pulsed radiation for military development applications. X-rays can be converted into UV at about 50% energy efficiency

NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL ASSOCIATED WITH A CLOSED FUEL CYCLE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bathke, C. G.; Ebbinghaus, B.; Sleaford, Brad W.

2010-06-11

This paper examines the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the various processing steps required for a closed fuel cycle. This paper combines the results from earlier studies that examined the attractiveness of SNM associated with the processing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR with new results for the final, repeated burning of SNM in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). The results of this paper suggest that all reprocessing products evaluated so farmore » need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, or undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed.« less

Validation of the U.S. NRC NGNP evaluation model with the HTTR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saller, T.; Seker, V.; Downar, T.

2012-07-01

The High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditional light water reactor (LWR) homogenization methods rely on the short mean free paths of neutrons in LWR. In gas-cooled, graphite-moderated reactors like the HTTR neutrons have much longer mean free paths and penetrate further into neighboring assemblies than in LWRs. Because of this, conventional lattice calculations with a single assembly may not be valid. In addition to difficulties caused by the longer mean free paths, the HTTR presents unique axial and radial heterogeneities that require additional modifications to the single assembly homogenization method. To handle these challenges, the homogenizationmore » domain is decreased while the computational domain is increased. Instead of homogenizing a single hexagonal fuel assembly, the assembly is split into six triangles on the radial plane and five blocks axially in order to account for the placement of burnable poisons. Furthermore, the radial domain is increased beyond a single fuel assembly to account for spectrum effects from neighboring fuel, reflector, and control rod assemblies. A series of five two-dimensional cases, each closer to the full core, were calculated to evaluate the effectiveness of the homogenization method and cross-sections. (authors)« less

Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

NASA Astrophysics Data System (ADS)

Lin, Jun-Li; Zhong, Weicheng; Bilheux, Hassina Z.; Heuser, Brent J.

2017-12-01

High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.

Economic Analysis of Complex Nuclear Fuel Cycles with NE-COST

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ganda, Francesco; Dixon, Brent; Hoffman, Edward

The purpose of this work is to present a new methodology, and associated computational tools, developed within the U.S. Department of Energy (U.S. DOE) Fuel Cycle Option Campaign to quantify the economic performance of complex nuclear fuel cycles. The levelized electricity cost at the busbar is generally chosen to quantify and compare the economic performance of different baseload generating technologies, including of nuclear: it is the cost of electricity which renders the risk-adjusted discounted net present value of the investment cash flow equal to zero. The work presented here is focused on the calculation of the levelized cost of electricitymore » of fuel cycles at mass balance equilibrium, which is termed LCAE (Levelized Cost of Electricity at Equilibrium). To alleviate the computational issues associated with the calculation of the LCAE for complex fuel cycles, a novel approach has been developed, which has been called the “island approach” because of its logical structure: a generic complex fuel cycle is subdivided into subsets of fuel cycle facilities, called islands, each containing one and only one type of reactor or blanket and an arbitrary number of fuel cycle facilities. A nuclear economic software tool, NE-COST, written in the commercial programming software MATLAB®, has been developed to calculate the LCAE of complex fuel cycles with the “island” computational approach. NE-COST has also been developed with the capability to handle uncertainty: the input parameters (both unit costs and fuel cycle characteristics) can have uncertainty distributions associated with them, and the output can be computed in terms of probability density functions of the LCAE. In this paper NE-COST will be used to quantify, as examples, the economic performance of (1) current Light Water Reactors (LWR) once-through systems; (2) continuous plutonium recycling in Fast Reactors (FR) with driver and blanket; (3) Recycling of plutonium bred in FR into LWR. For each fuel cycle, the contributions to the total LCAE of the main cost components will be identified.« less

Grain growth in uranium nitride prepared by spark plasma sintering

NASA Astrophysics Data System (ADS)

Johnson, Kyle D.; Lopes, Denise Adorno

2018-05-01

Uranium mononitride (UN) has long been considered a potential high density, high performance fuel candidate for light water reactor (LWR) and fast reactor (FR) applications. However, deployability of this fuel has been limited by the notable resistance to sintering and subsequent difficulty in producing a desirable microstructure, the high costs associated with 15N enrichment, as well as the known proclivity to oxidation and interaction with steam. In this study, the stimulation of grain growth in UN pellets sintered using SPS has been investigated. The results reveal that by using SPS and controlling temperature, time, and holding pressure, grain growth can be stimulated and controlled to produce a material featuring both a desired porosity and grain size, at least within the range of interest for nuclear fuel candidates. Grain sizes up to 31 μm were obtained using temperatures of 1650 °C and hold times of 15 min. Evaluation by EBSD reveal grain rotation and coalescence as the dominant mechanism in grain growth, which is suppressed by the application of higher external pressure. Moreover, complete closure of the porosity of the material was observed at relative densities of 96% TD, resulting in a material with sufficient porosity to accommodate LWR burnup. These results indicate that a method exists for the economic fabrication of an 15N-bearing uranium mononitride fuel with favorable microstructural characteristics compatible with use in a light water-cooled nuclear reactor.

Experimental studies at the Idaho Chemical Processing Plant on actinide partitioning from acidic nuclear wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

McIssaac, L. D.; Baker, J. D.; Meikrantz, D. H.

1980-01-01

Wastes generated at ICPP and in the reprocessing of LWR fuel is discussed separately. DHDECMP is used as extractant. Studies on DHDECMP purification and toxicity, diluent effects, reaction kinetics, radioloysis, mixer-settler performance, etc. are reported. 10 tables, 3 figures. (DLC)

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

DOE PAGES

Menlove, Howard Olsen; Belian, Anthony P.; Geist, William H.; ...

2017-10-07

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2O 3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate themore » 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).« less

Towards the reanalysis of void coefficients measurements at proteus for high conversion light water reactor lattices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hursin, M.; Koeberl, O.; Perret, G.

2012-07-01

High Conversion Light Water Reactors (HCLWR) allows a better usage of fuel resources thanks to a higher breeding ratio than standard LWR. Their uses together with the current fleet of LWR constitute a fuel cycle thoroughly studied in Japan and the US today. However, one of the issues related to HCLWR is their void reactivity coefficient (VRC), which can be positive. Accurate predictions of void reactivity coefficient in HCLWR conditions and their comparisons with representative experiments are therefore required. In this paper an inter comparison of modern codes and cross-section libraries is performed for a former Benchmark on Void Reactivitymore » Effect in PWRs conducted by the OECD/NEA. It shows an overview of the k-inf values and their associated VRC obtained for infinite lattice calculations with UO{sub 2} and highly enriched MOX fuel cells. The codes MCNPX2.5, TRIPOLI4.4 and CASMO-5 in conjunction with the libraries ENDF/B-VI.8, -VII.0, JEF-2.2 and JEFF-3.1 are used. A non-negligible spread of results for voided conditions is found for the high content MOX fuel. The spread of eigenvalues for the moderated and voided UO{sub 2} fuel are about 200 pcm and 700 pcm, respectively. The standard deviation for the VRCs for the UO{sub 2} fuel is about 0.7% while the one for the MOX fuel is about 13%. This work shows that an appropriate treatment of the unresolved resonance energy range is an important issue for the accurate determination of the void reactivity effect for HCLWR. A comparison to experimental results is needed to resolve the presented discrepancies. (authors)« less

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Menlove, Howard Olsen; Belian, Anthony P.; Geist, William H.

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2O 3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate themore » 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).« less

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

NASA Astrophysics Data System (ADS)

Menlove, Howard; Belian, Anthony; Geist, William; Rael, Carlos

2018-01-01

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd2O3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate the 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. We have named the system the fast-neutron passive collar (FNPC).

Analyzing the impact of reactive transport on the repository performance of TRISO fuel

NASA Astrophysics Data System (ADS)

Schmidt, Gregory

One of the largest determiners of the amount of electricity generated by current nuclear reactors is the efficiency of the thermodynamic cycle used for power generation. Current light water reactors (LWR) have an efficiency of 35% or less for the conversion of heat energy generated by the reactor to electrical energy. If this efficiency could be improved, more power could be generated from equivalent volumes of nuclear fuel. One method of improving this efficiency is to use a coolant flow that operates at a much higher temperature for electricity production. A reactor design that is currently proposed to take advantage of this efficiency is a graphite-moderated, helium-cooled reactor known as a High Temperature Gas Reactor (HTGR). There are significant differences between current LWR's and the proposed HTGR's but most especially in the composition of the nuclear fuel. For LWR's, the fuel elements consist of pellets of uranium dioxide or plutonium dioxide that are placed in long tubes made of zirconium metal alloys. For HTGR's, the fuel, known as TRISO (TRIstructural-ISOtropic) fuel, consists of an inner sphere of fissile material, a layer of dense pyrolytic carbon (PyC), a ceramic layer of silicon carbide (SiC) and a final dense outer layer of PyC. These TRISO particles are then compacted with graphite into fuel rods that are then placed in channels in graphite blocks. The blocks are then arranged in an annular fashion to form a reactor core. However, this new fuel form has unanswered questions on the environmental post-burn-up behavior. The key question for current once-through fuel operations is how these large irradiated graphite blocks with spent fuel inside will behave in a repository environment. Data in the literature to answer this question is lacking, but nevertheless this is an important question that must be answered before wide-spread adoption of HTGR's could be considered. This research has focused on answering the question of how the large quantity of graphite surrounding the spent HTGR fuel will impact the release of aqueous uranium from the TRISO fuel. In order to answer this question, the sorption and partitioning behavior of uranium to graphite under a variety of conditions was investigated. Key systematic variables that were analyzed include solution pH, dissolved carbonate concentration, uranium metal concentration and ionic strength. The kinetics and desorption characteristics of uranium/graphite partitioning were studied as well. The graphite used in these experiments was also characterized by a variety of techniques and conclusions are drawn about the relevant surface chemistry of graphite. This data was then used to generate a model for the reactive transport of uranium in a graphite matrix. This model was implemented with the software code CXTFIT and validated through the use of column studies mirroring the predicted system.

Microstructure and hydrothermal corrosion behavior of NITE-SiC with various sintering additives in LWR coolant environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parish, Chad M.; Terrani, Kurt A.; Kim, Young -Jin

Nano-infiltration and transient eutectic phase (NITE) sintering was developed for fabrication of nuclear grade SiC composites. We produced monolithic SiC ceramics using NITE sintering, as candidates for accident-tolerant fuels in light-water reactors (LWRs). In this work, we exposed three different NITE chemistries (yttria-alumina [YA], ceria-zirconia-alumina [CZA], and yttria-zirconia-alumina [YZA]) to autoclave conditions simulating LWR coolant loops. The YZA was most corrosion resistant, followed by CZA, with YA being worst. High-resolution elemental analysis using scanning transmission electron microscopy (STEM) X-ray mapping combined with multivariate statistical analysis (MVSA) datamining helped explain the differences in corrosion. YA-NITE lost all Al from the corrodedmore » region and the ytttria reformed into blocky precipitates. The CZA material lost all Al from the corroded area, and the YZA – which suffered the least corrosion –retained some Al in the corroded region. Lastly, the results indicate that the YZA-NITE SiC is most resistant to hydrothermal corrosion in the LWR environment.« less

Imminent: Irradiation Testing of (Th,Pu)O{sub 2} Fuel - 13560

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, Julian F.; Franceschini, Fausto

2013-07-01

Commercial-prototype thorium-plutonium oxide (Th-MOX) fuel pellets have been loaded into the material test reactor in Halden, Norway. The fuel is being operated at full power - with instrumentation - in simulated LWR / PHWR conditions and its behaviour is measured 'on-line' as it operates to high burn-up. This is a vital test on the commercialization pathway for this robust new thoria-based fuel. The performance data that is collected will support a fuel modeling effort to support its safety qualification. Several different samples of Th-MOX fuel will be tested, thereby collecting information on ceramic behaviours and their microstructure dependency. The fuel-cyclemore » reasoning underpinning the test campaign is that commercial Th- MOX fuels are an achievable intermediate / near-term SNF management strategy that integrates well with a fast reactor future. (authors)« less

Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions

NASA Astrophysics Data System (ADS)

Ott, L. J.; Robb, K. R.; Wang, D.

2014-05-01

Following the severe accidents at the Japanese Fukushima Daiichi Nuclear Power Station in 2011, the US Department of Energy initiated research and development on the enhancement of the accident tolerance of light water reactors by the development of fuels/cladding that, in comparison with the standard UO2/Zircaloy (Zr) system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations. Analyses are presented that illustrate the impact of these new candidate fuel/cladding materials on the fuel performance at normal operating conditions and on the reactor system under DB and BDB accident conditions.

Evaluation Metrics Applied to Accident Tolerant Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shannon M. Bragg-Sitton; Jon Carmack; Frank Goldner

2014-10-01

The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and have yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. One of the current missions of the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) is to develop nuclear fuelsmore » and claddings with enhanced accident tolerance for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+). Accident tolerance became a focus within advanced LWR research upon direction from Congress following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness and economics of commercial nuclear power. Enhanced accident tolerant fuels would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The U.S. DOE is supporting multiple teams to investigate a number of technologies that may improve fuel system response and behavior in accident conditions, with team leadership provided by DOE national laboratories, universities, and the nuclear industry. Concepts under consideration offer both evolutionary and revolutionary changes to the current nuclear fuel system. Mature concepts will be tested in the Advanced Test Reactor at Idaho National Laboratory beginning in Summer 2014 with additional concepts being readied for insertion in fiscal year 2015. This paper provides a brief summary of the proposed evaluation process that would be used to evaluate and prioritize the candidate accident tolerant fuel concepts currently under development.« less

The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, R.R.

1986-01-01

This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

Fission product release from fuel under LWR accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, M.F.; Lorenz, R.A.; Norwood, K.S.

Three tests have provided additional data on fission product release under LWR accident conditions in a temperature range (1400 to 2000/sup 0/C). In the release rate data are compared with curves from a recent NRC-sponsored review of available fission product release data. Although the iodine release in test HI-3 was inexplicably low, the other data points for Kr, I, and Cs fall reasonably close to the corresponding curve, thereby tending to verify the NRC review. The limited data for antimony and silver release fall below the curves. Results of spark source mass spectrometric analyses were in agreement with the gammamore » spectrometric results. Nonradioactive fission products such as Rb and Br appeared to behave like their chemical analogs Cs and I. Results suggest that Te, Ag, Sn, and Sb are released from the fuel in elemental form. Analysis of the cesium and iodine profiles in the thermal gradient tube indicates that iodine was deposited as CsT along with some other less volatile cesium compound. The cesium profiles and chemical reactivity indicate the presence of more than one cesium species.« less

Experimental critical loadings and control rod worths in LWR-PROTEUS configurations compared with MCNPX results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Plaschy, M.; Murphy, M.; Jatuff, F.

2006-07-01

The PROTEUS research reactor at the Paul Scherrer Inst. (PSI) has been operating since the sixties and has already permitted, due to its high flexibility, investigation of a large range of very different nuclear systems. Currently, the ongoing experimental programme is called LWR-PROTEUS. This programme was started in 1997 and concerns large-scale investigations of advanced light water reactors (LWR) fuels. Until now, the different LWR-PROTEUS phases have permitted to study more than fifteen different configurations, each of them having to be demonstrated to be operationally safe, in particular, for the Swiss safety authorities. In this context, recent developments of themore » PSI computer capabilities have made possible the use of full-scale SD-heterogeneous MCNPX models to calculate accurately different safety related parameters (e.g. the critical driver loading and the shutdown rod worth). The current paper presents the MCNPX predictions of these operational characteristics for seven different LWR-PROTEUS configurations using a large number of nuclear data libraries. More specifically, this significant benchmarking exercise is based on the ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JENDL3.2, and JENDL3.3 libraries. The results highlight certain library specific trends in the prediction of the multiplication factor k{sub eff} (e.g. the systematically larger reactivity calculated with JEF2.2 and the smaller reactivity associated with JEFF3.0). They also confirm the satisfactory determination of reactivity variations by all calculational schemes, for instance, due to the introduction of a safety rod pair, these calculations having been compared with experiments. (authors)« less

Advanced Small Modular Reactor Economics Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, Thomas J.

2014-10-01

This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic andmore » nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial operation. Advanced fuel materials and fabrication costs have large uncertainties based on complexities of operation, such as contact-handled fuel fabrication versus remote handling, or commodity availability. Thus, this analytical work makes a good faith effort to quantify uncertainties and provide qualifiers, caveats, and explanations for the sources of these uncertainties. The overall result is that this work assembles the necessary information and establishes the foundation for future analyses using more precise data as nuclear technology advances.« less

Development of self-interrogation neutron resonance densitometry (sinrd) to measure the fissile content in nuclear fuel

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne Marie

The development of non-destructive assay (NDA) capabilities to directly measure the fissile content in spent fuel is needed to improve the timely detection of the diversion of significant quantities of fissile material. Currently, the International Atomic Energy Agency (IAEA) does not have effective NDA methods to verify spent fuel and recover continuity of knowledge in the event of a containment and surveillance systems failure. This issue has become increasingly critical with the worldwide expansion of nuclear power, adoption of enhanced safeguards criteria for spent fuel verification, and recent efforts by the IAEA to incorporate an integrated safeguards regime. In order to address these issues, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been developed to improve existing nuclear safeguards and material accountability measurements. The following characteristics of SINRD were analyzed: (1) ability to measure the fissile content in Light Water Reactors (LWR) fuel assemblies and (2) sensitivity and penetrability of SINRD to the removal of fuel pins from an assembly. The Monte Carlo Neutral Particle eXtended (MCNPX) transport code was used to simulate SINRD for different geometries. Experimental measurements were also performed with SINRD and were compared to MCNPX simulations of the experiment to verify the accuracy of the MCNPX model of SINRD. Based on the results from these simulations and measurements, we have concluded that SINRD provides a number of improvements over current IAEA verification methods. These improvements include: (1) SINRD provides absolute measurements of burnup independent of the operator's declaration. (2) SINRD is sensitive to pin removal over the entire burnup range and can verify the diversion of 6% of fuel pins within 3o from LWR spent LEU and MOX fuel. (3) SINRD is insensitive to the boron concentration and initial fuel enrichment and can therefore be used at multiple spent fuel storage facilities. (4) The calibration of SINRD at one reactor facility carries over to reactor sites in different countries because it uses the ratio of fission chambers (FCs) that are not facility dependent. (5) SINRD can distinguish fresh and 1-cycle spent MOX fuel from 3- and 4-cycles spent LEU fuel without using reactor burnup codes.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kristine Barrett; Shannon Bragg-Sitton

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system thatmore » would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.« less

Nuclear fuels - Present and future

NASA Astrophysics Data System (ADS)

Olander, D.

2009-06-01

The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert O.; Bennett, Wendy D.; Henager, Charles H.

2016-09-12

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impactmore » of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.« less

Establishment of the roadmap for chlorination process development for zirconium recovery and recycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, E.D.; Del Cul, G.D.; Spencer, B.B.

Process development studies are being conducted to recover, purify, and reuse the zirconium (about 98.5% by mass) in used nuclear fuel (UNF) zirconium alloy cladding. Feasibility studies began in FY 2010 using empty cladding hulls that were left after fuel dissolution or after oxidation to a finely divided oxide powder (voloxidation). In FY 2012, two industrial teams (AREVA and Shaw-Westinghouse) were contracted by the Department of Energy Office of Nuclear Energy (NE) to provide technical assistance to the project. In FY 2013, the NE Fuel Cycle Research and Development Program requested development of a technology development roadmap to guide futuremore » work. The first step in the roadmap development was to assess the starting point, that is, the current state of the technology and the end goal. Based on previous test results, future work is to be focused on first using chlorine as the chlorinating agent and secondly on the use of a process design that utilizes a chlorination reactor and dual ZrCl{sub 4} product salt condensers. The likely need for a secondary purification step was recognized. Completion of feasibility testing required an experiment on the chemical decladding flowsheet option. This was done during April 2013. The roadmap for process development will continue through process chemistry optimization studies, the chlorinated reactor design configuration, product salt condensers, and the off-gas trapping of tritium or other volatile fission products from the off-gas stream. (authors)« less

Transient Testing of Nuclear Fuels and Materials in the United States

NASA Astrophysics Data System (ADS)

Wachs, Daniel M.

2012-12-01

The United States has established that transient irradiation testing is needed to support advanced light water reactors fuel development. The U.S. Department of Energy (DOE) has initiated an effort to reestablish this capability. Restart of the Transient Testing Reactor (TREAT) facility located at the Idaho National Laboratory (INL) is being considered for this purpose. This effort would also include the development of specialized test vehicles to support stagnant capsule and flowing loop tests as well as the enhancement of postirradiation examination capabilities and remote device assembly capabilities at the Hot Fuel Examination Facility. It is anticipated that the capability will be available to support testing by 2018, as required to meet the DOE goals for the development of accident-tolerant LWR fuel designs.

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Root, M. A.; Menlove, H. O.; Lanza, R. C.

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE PAGES

Root, M. A.; Menlove, H. O.; Lanza, R. C.; ...

2018-03-21

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Fission Product Release and Survivability of UN-Kernel LWR TRISO Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Besmann, Theodore M; Ferber, Mattison K; Lin, Hua-Tay

2014-01-01

A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from range calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 m diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated with a TRISO particle as a function of fluence. Creep and swelling of the innermore » and outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by measuring the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers as a function of fluence. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.« less

Fission product release and survivability of UN-kernel LWR TRISO fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

T. M. Besmann; M. K. Ferber; H.-T. Lin

2014-05-01

A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from fission product recoil calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 um diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated within a TRISO particle undergoing burnup. Creep and swelling of the inner andmore » outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by computing the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers from internal pressure and thermomechanics of the layers. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.« less

Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella J van Rooyen

2012-09-01

The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrixmore » composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing« less

Significance of and prospects for fuel recycle in Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Otsuka, K.; Ikeda, K.

Japan's nuclear power plant capacity ranks fourth in the world at around 20 GW. But nuclear fuel cycle industries (enrichment, reprocessing and radioactive waste management) are still in their infancy compared with the size and stage of the power plants. Thus it is a matter of urgency to establish a nuclear fuel cycle in Japan which can promote nuclear energy as a quasi-indigenous energy source. Some moves toward establishing a nuclear fuel cycle have been observed recently. As a case in point, in July 1984, the Federation of Electric Power Companies has formally requested Aomori Prefecture to locate nuclear fuelmore » cycle facilities in the Shimokita Peninsula region. Plutonium recovered from spent fuel will be utilized in LWR, ATR, and FBR. Research and development activities on these technologies are in progress.« less

The underwater coincidence counter (UWCC) for plutonium measurements in mixed oxide fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eccleston, G.W.; Menlove, H.O.; Abhold, M.

1998-12-31

The use of fresh uranium-plutonium mixed oxide (MOX) fuel in light-water reactors (LWR) is increasing in Europe and Japan and it is necessary to verify the plutonium content in the fuel for international safeguards purposes. The UWCC is a new instrument that has been designed to operate underwater and nondestructively measure the plutonium in unirradiated MOX fuel assemblies. The UWCC can be quickly configured to measure either boiling-water reactor (BWR) or pressurized-water reactor (PWR) fuel assemblies. The plutonium loading per unit length is measured using the UWCC to precisions of less than 1% in a measurement time of 2 tomore » 3 minutes. Initial calibrations of the UWCC were completed on measurements of MOX fuel in Mol, Belgium. The MCNP-REN Monte Carlo simulation code is being benchmarked to the calibration measurements to allow accurate simulations for extended calibrations of the UWCC.« less

Neutron source, linear-accelerator fuel enricher and regenerator and associated methods

DOEpatents

Steinberg, Meyer; Powell, James R.; Takahashi, Hiroshi; Grand, Pierre; Kouts, Herbert

1982-01-01

A device for producing fissile material inside of fabricated nuclear elements so that they can be used to produce power in nuclear power reactors. Fuel elements, for example, of a LWR are placed in pressure tubes in a vessel surrounding a liquid lead-bismuth flowing columnar target. A linear-accelerator proton beam enters the side of the vessel and impinges on the dispersed liquid lead-bismuth columns and produces neutrons which radiate through the surrounding pressure tube assembly or blanket containing the nuclear fuel elements. These neutrons are absorbed by the natural fertile uranium-238 elements and are transformed to fissile plutonium-239. The fertile fuel is thus enriched in fissile material to a concentration whereby they can be used in power reactors. After use in the power reactors, dispensed depleted fuel elements can be reinserted into the pressure tubes surrounding the target and the nuclear fuel regenerated for further burning in the power reactor.

Opportunities for the Multi Recycling of Used MOX Fuel in the US - 12122

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, P.; Bailly, F.; Bouvier, E.

Over the last 50 years the US has accumulated an inventory of used nuclear fuel (UNF) in the region of 64,000 metric tons in 2010, and adds an additional 2,200 metric tons each year from the current fleet of 104 Light Water Reactors. This paper considers a fuel cycle option that would be available for a future pilot U.S. recycling plant that could take advantage of the unique opportunities offered by the age and size of the large U.S. UNF inventory. For the purpose of this scenario, recycling of UNF must use the available reactor infrastructure, currently LWR's, and themore » main product of recycling is considered to be plutonium (Pu), recycled into MOX fuel for use in these reactors. Use of MOX fuels must provide the service (burn-up) expected by the reactor operator, with the required level of safety. To do so, the fissile material concentration (Pu-239, Pu-241) in the MOX must be high enough to maintain criticality, while, in current recycle facilities, the Pu-238 content has to be kept low enough to prevent excessive heat load, neutron emission, and neutron capture during recycle operations. In most countries, used MOX fuel (MOX UNF) is typically stored after one irradiation in an LWR, pending the development of the GEN IV reactors, since it is considered difficult to directly reuse the recycled MOX fuel in LWRs due to the degraded Pu fissile isotopic composition. In the US, it is possible to blend MOX UNF with LEUOx UNF from the large inventory, using the oldest UNF first. Blending at the ratio of about one MOX UNF assembly with 15 LEUOx UNF assemblies, would achieve a fissile plutonium concentration sufficient for reirradiation in new MOX fuel. The Pu-238 yield in the new fuel will be sufficiently low to meet current fuel fabrication standards. Therefore, it should be possible in the context of the US, for discharged MOX fuel to be recycled back into LWR's, using only technologies already industrially deployed worldwide. Building on that possibility, two scenarios are assessed where current US inventory is treated; Pu recycled in LWR MOX fuels, and used MOX fuels themselves are treated in a continuous partitioning-transmutation mode (case 2a) or until the whole current UNF inventory (64,000 MT in 2010) has been treated followed by disposal of the MOX UNF to a geologic repository (case 2b). In the recycling scenario, two cases (2a and 2b) are considered. Benefits achieved are compared with the once through scenario (case 1) where UNF in the current US inventory are disposed directly to a geologic repository. For each scenario, the heat load and radioactivity of the high activity wastes disposed to a geologic repository are calculated and the savings in natural resources quantified, and compared with the once-through fuel cycle. Assuming an initial pilot recycling facility with a capacity of 800 metric tons a year of heavy metal begins operation in 2030, ∼8 metric tons per year of Pu is recovered from the LEUOx UNF inventory, and is used to produce fresh MOX fuels. At a later time, additional treatment and recycling capacities are assumed to begin operation, to accommodate blending and recycling of used MOX Pu, up to 2,400 MT/yr treatment capacity to enable processing UNF slightly faster than the rate of generation. Results of this scenario analysis study show the flexibility of the recycling scenarios so that Pu is managed in a way that avoids accumulating used MOX fuels. If at some future date, the decision is made to dispose of the MOX UNF to a geologic repository (case 2b), the scenario is neutral to final repository heat load in comparison to the direct disposal of all UNF (case 1), while diminishing use of natural uranium, enrichment, UNF accumulation, and the volume of HLW. Further recycling of Pu at the end of the scenario (case 2a) would exhibit further benefits. As expected, Pu-241 and Am-241 are the source of long term HLW heat load and Am-241 and Np-237 are the source of long term radiotoxicity. When advanced technology is available, introduction of minor actinide recycling, in addition to Pu recycling, by the end of this scenario, or sooner, would have a major impact on final repository heat load and long term radiotoxicity of the HLW. This scenario is also compatible with a gradual introduction of a small number of FR's for Pu management. (authors)« less

Phased Development of Accident Tolerant Fue

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bragg-Sitton, Shannon M.; Carmack, W. Jon

2016-09-01

The United States Department of Energy (U.S. DOE) Advanced Fuels Campaign (AFC) has adopted a three-phase approach for the development and eventual commercialization of enhanced, accident tolerant fuel (ATF) for light water reactors (LWRs). Extending from 2012 to 2016, AFC is currently coming to the end of Phase 1 research that has entailed Feasibility Assessment and Prioritization for a large number of proposed fuel systems (fuel and cladding) that could provide improved performance under accident conditions. Phase 1 activities will culminate with a prioritization of concepts for both near-term and long-term development based on the available experimental data and modelingmore » predictions. This process will provide guidance to DOE on what concepts should be prioritized for investment in Phase 2 Development/Qualification activities based on technical performance improvements and probability of meeting the aggressive schedule to insert a lead fuel rod (LFR) in a commercial power reactor by 2022. While Phase 1 activities include small-scale fabrication work, materials characterization, and limited irradiation of samples, Phase 2 will require development teams to expand to industrial fabrication methods, conduct irradiation tests under more prototypic reactor conditions (i.e. in contact with reactor primary coolant at LWR conditions and in-pile transient testing), conduct additional characterization and post-irradiation examination, and develop a fuel performance code for the candidate ATF. Phase 2 will culminate in the insertion of an LFR (or lead fuel assembly) in a commercial power reactor. The Phase 3 Commercialization work will extend past 2022. Following post-irradiation examination of LFRs, partial-core reloads will be demonstrated. The commercialization phase will further entail the establishment of commercial fabrication capabilities and the transition of LWR cores to the new fuel. The three development phases described roughly correspond to the technology readiness levels (TRL) defined for nuclear fuel development. TRL 1–3 corresponds to the “proof-of-concept” stage (Phase 1), TRL 4–6 to “proof-of-principle” (Phase 2), and TRL 7–9 to “proof-of-performance” (Phase 3). This paper will provide an overview of the anticipated activities within each phase of development and will provide an update on the current ATF development status.« less

Ceria-thoria pellet manufacturing in preparation for plutonia-thoria LWR fuel production

NASA Astrophysics Data System (ADS)

Drera, Saleem S.; Björk, Klara Insulander; Sobieska, Matylda

2016-10-01

Thorium dioxide (thoria) has potential to assist in niche roles as fuel for light water reactors (LWRs). One such application for thoria is its use as the fertile component to burn plutonium in a mixed oxide fuel (MOX). Thor Energy and an international consortium are currently irradiating plutonia-thoria (Th-MOX) fuel in an effort to produce data for its licensing basis. During fuel-manufacturing research and development (R&D), surrogate materials were utilized to highlight procedures and build experience. Cerium dioxide (ceria) provides a good surrogate platform to replicate the chemical nature of plutonium dioxide. The project's fuel manufacturing R&D focused on powder metallurgical techniques to ensure manufacturability with the current commercial MOX fuel production infrastructure. The following paper highlights basics of the ceria-thoria fuel production including powder milling, pellet pressing and pellet sintering. Green pellets and sintered pellets were manufactured with average densities of 67.0% and 95.5% that of theoretical density respectively.

Study of fission-product segregation in used CANDU fuel by X-ray photoelectron spectroscopy (XPS) II

NASA Astrophysics Data System (ADS)

Hocking, William H.; Duclos, A. Michael; Johnson, Lawrence H.

1994-03-01

A thorough investigation of the grain-boundary chemistry of used CANDU fuel from one intact element has been conducted by X-ray photoelectron spectroscopy (XPS). Selected findings from more extensive XPS measurements on other used CANDU fuels exposed to storage conditions are included for comparison. Cesium, rubidium, tellurium and barium have been commonly observed, often reaching high degrees of surface enrichment, although their relative abundances can vary widely with a complex dependence on the fuel irradiation history. Lower concentrations of cadmium, molybdenum, strontium and iodine have also been occasionally detected. Except for iodine, chemical-shift data are indicative of oxidized species, possibly uranates. Segregation at monolayer-level coverages has been demonstrated by sequential XPS analysis and argon-ion sputtering. Calculations based on an idealized thin-film model are consistent with the depth profiles. The interpretation of these results is discussed in the context of previous studies, especially on LWR fuels.

High-temperature Gas Reactor (HTGR)

NASA Astrophysics Data System (ADS)

Abedi, Sajad

2011-05-01

General Atomics (GA) has over 35 years experience in prismatic block High-temperature Gas Reactor (HTGR) technology design. During this period, the design has recently involved into a modular have been performed to demonstrate its versatility. This versatility is directly related to refractory TRISO coated - particle fuel that can contain any type of fuel. This paper summarized GA's fuel cycle studies individually and compares each based upon its cycle sustainability, proliferation-resistance capabilities, and other performance data against pressurized water reactor (PWR) fuel cycle data. Fuel cycle studies LEU-NV;commercial HEU-Th;commercial LEU-Th;weapons-grade plutonium consumption; and burning of LWR waste including plutonium and minor actinides in the MHR. results show that all commercial MHR options, with the exception of HEU-TH, are more sustainable than a PWR fuel cycle. With LEU-NV being the most sustainable commercial options. In addition, all commercial MHR options out perform the PWR with regards to its proliferation-resistance, with thorium fuel cycle having the best proliferation-resistance characteristics.

Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium-Plutonium Carbide based fuel and 300 MWt Power

NASA Astrophysics Data System (ADS)

Clief Pattipawaej, Sandro; Su'ud, Zaki

2017-01-01

A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

Overview of experimental support for fission-product transport analyses at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.

The program was designed to determine fission product and aerosol release rates from irradiated fuel under accident conditions, to identify the chemical forms of the released material, and to correlate the results with experimental and specimen conditions with the data from related experiments. These tests of PWR fuel were conducted and fuel specimen and test operating data are presented. The nature and rate of fission product vapor interaction with aerosols were studied. Aerosol deposition rates and transport in the reactor vessel during LWR core-melt accidents were studied. The Nuclear Safety Pilot Plant is dedicated to developing an expanded data basemore » on the behavior of aerosols generated during a severe accident.« less

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

NASA Astrophysics Data System (ADS)

Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-01

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-29

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclearmore » waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.« less

Predicting thermo-mechanical behaviour of high minor actinide content composite oxide fuel in a dedicated transmutation facility

NASA Astrophysics Data System (ADS)

Lemehov, S. E.; Sobolev, V. P.; Verwerft, M.

2011-09-01

The European Facility for Industrial Transmutation (EFIT) of the minor actinides (MA), from LWR spent fuel is being developed in the integrated project EUROTRANS within the 6th Framework Program of EURATOM. Two composite uranium-free fuel systems, containing a large fraction of MA, are proposed as the main candidates: a CERCER with magnesia matrix hosting (Pu,MA)O 2-x particles, and a CERMET with metallic molybdenum matrix. The long-term thermal and mechanical behaviour of the fuel under the expected EFIT operating conditions is one of the critical issues in the core design. To make a reliable prediction of long-term thermo-mechanical behaviour of the hottest fuel rods in the lead-cooled version of EFIT with thermal power of 400 MW, different fuel performance codes have been used. This study describes the main results of modelling the thermo-mechanical behaviour of the hottest CERCER fuel rods with the fuel performance code MACROS which indicate that the CERCER fuel residence time can safely reach at least 4-5 effective full power years.

Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. L. Williamson

A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete andmore » smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.« less

FRAPCON analysis of cladding performance during dry storage operations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richmond, David J.; Geelhood, Kenneth J.

There is an increasing need in the U.S. and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations (ISFSI) or interim storage sites. The NRC limits cladding temperature to 400°C while maintaining cladding hoop stress below 90 MPa in an effort to avoid radial hydride reorientation. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at 400 °C. Results were representative of the majority of U.S. LWR fuel. They conservativelymore » showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.« less

Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Howard, Richard H.

FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiationmore » tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO 2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO 2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys, hence promoting FCCI between the fuel-clad systems. The other factor was to develop a test bed where multiple candidate alloys could be evaluated within a single irradiation test train, thereby reducing overall costs and increasing efficiency in alloy screening efforts. A collaboration between ORNL and INL was developed to facilitate the completion of the test bed for FCCI testing. The report highlights the activities related to the development of the ATF-1 ORNL FCCI rodlets for irradiation in INL’s ATR as part of the on-going ATF-1 experiments.« less

Method and apparatus for controlled size distribution of gel microspheres formed from aqueous dispersions

DOEpatents

Ryon, Allen D.; Haas, Paul A.; Vavruska, John S.

1984-01-01

The present invention is directed to a method and apparatus for making a population of dense, closely size-controlled microspheres by sol-gel procedures wherein said microspheres are characterized by a significant percentage of said population being within a predetermined, relatively narrow size range. Microsphere populations thus provided are useful in vibratory-packed processes for nuclear fuels to be irradiated in LWR- and FBR-type nuclear reactors.

Fission products and nuclear fuel behaviour under severe accident conditions part 2: Fuel behaviour in the VERDON-1 sample

NASA Astrophysics Data System (ADS)

Geiger, E.; Le Gall, C.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

2017-11-01

Within the framework of the International Source Term Programme (ISTP), the VERDON programme aims at quantifying the source term of radioactive materials in case of a hypothetical severe accident in a light water reactor (LWR). Tests were performed in a new experimental laboratory (VERDON) built in the LECA-STAR facility (CEA Cadarache). The VERDON-1 test was devoted to the study of a high burn-up UO2 fuel and FP releases at very high temperature (≈2873 K) in a reducing atmosphere. Post-test qualitative and quantitative characterisations of the VERDON-1 sample led to the proposal of a scenario explaining the phenomena occurring during the experimental sequence. Hence, the fuel and the cladding may have interacted which led to the melting of UO2-ZrO2 alloy. Although no relocation was observed during the test, it may have been imminent.

Generic repository design concepts and thermal analysis (FY11).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Howard, Robert; Dupont, Mark; Blink, James A.

2011-08-01

Reference concepts for geologic disposal of used nuclear fuel and high-level radioactive waste in the U.S. are developed, including geologic settings and engineered barriers. Repository thermal analysis is demonstrated for a range of waste types from projected future, advanced nuclear fuel cycles. The results show significant differences among geologic media considered (clay/shale, crystalline rock, salt), and also that waste package size and waste loading must be limited to meet targeted maximum temperature values. In this study, the UFD R&D Campaign has developed a set of reference geologic disposal concepts for a range of waste types that could potentially be generatedmore » in advanced nuclear FCs. A disposal concept consists of three components: waste inventory, geologic setting, and concept of operations. Mature repository concepts have been developed in other countries for disposal of spent LWR fuel and HLW from reprocessing UNF, and these serve as starting points for developing this set. Additional design details and EBS concepts will be considered as the reference disposal concepts evolve. The waste inventory considered in this study includes: (1) direct disposal of SNF from the LWR fleet, including Gen III+ advanced LWRs being developed through the Nuclear Power 2010 Program, operating in a once-through cycle; (2) waste generated from reprocessing of LWR UOX UNF to recover U and Pu, and subsequent direct disposal of used Pu-MOX fuel (also used in LWRs) in a modified-open cycle; and (3) waste generated by continuous recycling of metal fuel from fast reactors operating in a TRU burner configuration, with additional TRU material input supplied from reprocessing of LWR UOX fuel. The geologic setting provides the natural barriers, and establishes the boundary conditions for performance of engineered barriers. The composition and physical properties of the host medium dictate design and construction approaches, and determine hydrologic and thermal responses of the disposal system. Clay/shale, salt, and crystalline rock media are selected as the basis for reference mined geologic disposal concepts in this study, consistent with advanced international repository programs, and previous investigations in the U.S. The U.S. pursued deep geologic disposal programs in crystalline rock, shale, salt, and volcanic rock in the years leading up to the Nuclear Waste Policy Act, or NWPA (Rechard et al. 2011). The 1987 NWPA amendment act focused the U.S. program on unsaturated, volcanic rock at the Yucca Mountain site, culminating in the 2008 license application. Additional work on unsaturated, crystalline rock settings (e.g., volcanic tuff) is not required to support this generic study. Reference disposal concepts are selected for the media listed above and for deep borehole disposal, drawing from recent work in the U.S. and internationally. The main features of the repository concepts are discussed in Section 4.5 and summarized in Table ES-1. Temperature histories at the waste package surface and a specified distance into the host rock are calculated for combinations of waste types and reference disposal concepts, specifying waste package emplacement modes. Target maximum waste package surface temperatures are identified, enabling a sensitivity study to inform the tradeoff between the quantity of waste per disposal package, and decay storage duration, with respect to peak temperature at the waste package surface. For surface storage duration on the order of 100 years or less, waste package sizes for direct disposal of SNF are effectively limited to 4-PWR configurations (or equivalent size and output). Thermal results are summarized, along with recommendations for follow-on work including adding additional reference concepts, verification and uncertainty analysis for thermal calculations, developing descriptions of surface facilities and other system details, and cost estimation to support system-level evaluations.« less

Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heuser, Brent; Stubbins, James; Kozlowski, Tomasz

The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys.more » The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be. International fabrication options were explored in Europe and Asia, but this proved to be impractical, if not impossible. Consequently, experimental investigation of the Zr-Be binary system was dropped and exploration binary Zr-Y binary system was initiated. The motivation behind the Zr-Y system is the known thermodynamic stability of yttria over zirconia.« less

Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

NASA Astrophysics Data System (ADS)

Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

2017-12-01

FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

NASA Astrophysics Data System (ADS)

Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

2018-02-01

FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

Microstructural Characterization of High Burn-up Mixed Oxide Fast Reactor Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Melissa C. Teague; Brian P. Gorman; Steven L. Hayes

2013-10-01

High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column weremore » observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.« less

Fast, quantitative, and nondestructive evaluation of hydrided LWR fuel cladding by small angle incoherent neutron scattering of hydrogen

DOE PAGES

Yan, Y.; Qian, S.; Littrell, K.; ...

2015-02-13

A non-destructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentration were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distributionmore » of circumferential hydrides across the wall. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. This study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount ( 20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor will be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for absolute hydrogen concentration determination.« less

In-pile tests at Karlsruhe of LWR fuel-rod behavior during the heatup phase of a LOCA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Karb, E.H.

1980-01-01

In order to investigate the influence of a nuclar environment on the mechanisms of fuel-rod failure, in-pile tests simulating the heatup phase of a loss-of-coolant accident in a pressurized-water reactor are being conducted with irradiated and unirradiated short-length single rods in the FR2 reactor at Kernforschungszentrum karlsruhe (Karlsruhe Nuclear Reasearch Center), Federal Republic of Germany, within the Project Nuclear Safety. With nearly 70% of the scheduled tests completed, no such influences have been found. The in-pile burst and deformation data are in good agreement with results from nonnuclear tests with electrically heated fuel-rod simulators. The phenomenon of pellet disintegration, whichmore » has been observed in all tests with previously irradiated rods, needs further investigation.« less

A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andress, D.; Joy, D.S.; McLeod, N.B.

The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elementsmore » as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs.« less

Verification and Validation of the BISON Fuel Performance Code for PCMI Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle Allan Lawrence; Novascone, Stephen Rhead; Gardner, Russell James

2016-06-01

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described. Validation for application to light water reactor (LWR) PCMI problems is assessed by comparing predicted and measured rod diameter following base irradiation andmore » power ramps. Results indicate a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

Prioritized List of Research Needs to support MRWFD Case Study Flowsheet Advancement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Law, Jack Douglas; Soelberg, Nicholas Ray

In FY-13, a case study evaluation was performed of full recycle technologies for both the processing of light-water reactor (LWR) used nuclear fuels as well as fast reactor (FR) fuel in the full recycle option. This effort focused on the identification of the case study processes and the initial preparation of material balance flowsheets for the identified technologies. In identifying the case study flowsheets, it was decided that two cases would be developed: one which identifies the flowsheet as currently developed and another near-term target flowsheet which identifies the flowsheet as envisioned within two years, pending the results of ongoingmore » research. The case study focus is on homogeneous aqueous recycle of the U/TRU resulting from the processing of LWR fuel as feed for metal fuel fabrication. The metal fuel is utilized in a sodium-cooled fast reactor, and the used fast reactor fuel is processed using electrochemical separations. The recovered U/TRU from electrochemical separations is recycled to fuel fabrication and the fast reactor. Waste streams from the aqueous and electrochemical processing are treated and prepared for disposition. Off-gas from the separations and waste processing are also treated. As part of the FY-13 effort, preliminary process unknowns and research needs to advance the near-term target flowsheets were identified. In FY-14, these research needs were updated, expanded and prioritized. This report again updates the prioritized list of research needs based upon results to date in FY-15. The research needs are listed for each of the main portions of the flowsheet: 1) Aqueous headend, 2) Headend tritium pretreatment off-gas, 3) Aqueous U/Pu/Np recovery, 4) Aqueous TRU product solidification, 5) Aqueous actinide/lanthanide separation, 6) Aqueous off-gas treatment, 7) Aqueous HLW management, 8) Treatment of aqueous process wastes, 9) E-chem actinide separations, 10) E-chem off-gas, 11) E-chem HLW management. The identified research needs were prioritized within each of these areas. No effort was made to perform an overall prioritization. This information will be used by the MRWFD Campaign leadership in research planning for FY-16. Additionally, this information will be incorporated into the next version of the Case Study Report scheduled to be issued September 2015.« less

Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

DOE Office of Scientific and Technical Information (OSTI.GOV)

Akimoto, Hajime; Kukita; Ohnuki, Akira

1997-07-01

The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

Understanding EUV mask blank surface roughness induced LWR and associated roughness requirement

NASA Astrophysics Data System (ADS)

Yan, Pei-Yang; Zhang, Guojing; Gullikson, Eric M.; Goldberg, Ken A.; Benk, Markus P.

2015-03-01

Extreme ultraviolet lithography (EUVL) mask multi-layer (ML) blank surface roughness specification historically comes from blank defect inspection tool requirement. Later, new concerns on ML surface roughness induced wafer pattern line width roughness (LWR) arise. In this paper, we have studied wafer level pattern LWR as a function of EUVL mask surface roughness via High-NA Actinic Reticle Review Tool. We found that the blank surface roughness induced LWR at current blank roughness level is in the order of 0.5nm 3σ for NA=0.42 at the best focus. At defocus of ±40nm, the corresponding LWR will be 0.2nm higher. Further reducing EUVL mask blank surface roughness will increase the blank cost with limited benefit in improving the pattern LWR, provided that the intrinsic resist LWR is in the order of 1nm and above.

Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahman, Fariz Abdul; Lee, John C.; Franceschini, Fausto

2012-07-01

As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning themore » legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and burn TRU in a thermal spectrum, while satisfying top-level operational and safety constraints. Various assembly designs have been proposed to assess the TRU burning potential of Th-based fuel in PWRs. In addition to typical homogeneous loading patterns, heterogeneous configurations exploiting the breeding potential of thorium to enable multiple cycles of TRU irradiation and burning have been devised. The homogeneous assembly design, with all pins featuring TRU in Th, has the benefit of a simple loading pattern and the highest rate of TRU transmutation, but it can be used only for a few cycles due to the rapid rise in the TRU content of the recycled fuel, which challenges reactivity control, safety coefficients and fuel handling. Due to its simple loading pattern, such assembly design can be used as the first step of Th implementation, achieving up to 3 times larger TRU transmutation rate than conventional U-MOX, assuming same fraction of MOX assemblies in the core. As the next step in thorium implementation, heterogeneous assemblies featuring a mixed array of Th-U and Th-U-TRU pins, where the U is in-bred from Th, have been proposed. These designs have the potential to enable burning an external supply of TRU through multiple cycles of irradiation, recovery (via reprocessing) and recycling of the residual actinides at the end of each irradiation cycle. This is achieved thanks to a larger breeding of U from Th in the heterogeneous assemblies, which reduces the TRU supply and thus mitigates the increase in the TRU core inventory for the multi-recycled fuel. While on an individual cycle basis the amount of TRU burned in the heterogeneous assembly is reduced with respect to the homogeneous design, TRU burning rates higher than single-pass U-MOX fuel can still be achieved, with the additional benefits of a multi-cycle transmutation campaign recycling all TRU isotopes. Nitride fuel, due its higher density and U breeding potential, together with its better thermal properties, ideally suits the objectives and constraints of the heterogeneous assemblies. However, significant technological advancements must be made before nitride fuels can be employed in an LWR: its water resistance needs to be improved and a viable technology to enrich N in N-15 must be devised. Moreover, for the nitride heterogeneous configurations examined in this study, the enhancement in TRU burning performance is achieved not only by replacing oxide with nitride fuel, but also by increasing the fuel rod size. This latter modification, allowed by the high thermal conductivity of nitride fuel, leads however to a very tight lattice, which may challenge reactor coolant pumps and assembly hold-down mechanisms, the former through an increase in core pressure drop and the latter through an increase in assembly lift-off forces. To alleviate these issues, while still achieving the large fuel-to-moderator ratios resulting from using tight lattices, wire wraps could be used in place of grid spacers. For tight lattices, typical grid spacers are hard to manufacture and their replacement with wire wraps is known to allow for a pressure drop reduction by at least 2 times. The studies, while certainly very preliminary, provide a starting point to devise an optimum strategy for TRU transmutation in Th-based PWR fuel. The viability of the scheme proposed depends on the timely phasing in of the associated technologies, with proper lead time and to solve the many challenges. These challenges are certainly substantial, and make the current once-through U-based scheme pursued in the US by far a more practical (and cheaper) option. However, when compared to other transmutation schemes, the proposed one has arguably similar challenges and unknowns with potentially bigger rewards. (authors)« less

MODELLING OF FUEL BEHAVIOUR DURING LOSS-OF-COOLANT ACCIDENTS USING THE BISON CODE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pastore, G.; Novascone, S. R.; Williamson, R. L.

2015-09-01

This work presents recent developments to extend the BISON code to enable fuel performance analysis during LOCAs. This newly developed capability accounts for the main physical phenomena involved, as well as the interactions among them and with the global fuel rod thermo-mechanical analysis. Specifically, new multiphysics models are incorporated in the code to describe (1) transient fission gas behaviour, (2) rapid steam-cladding oxidation, (3) Zircaloy solid-solid phase transition, (4) hydrogen generation and transport through the cladding, and (5) Zircaloy high-temperature non-linear mechanical behaviour and failure. Basic model characteristics are described, and a demonstration BISON analysis of a LWR fuel rodmore » undergoing a LOCA accident is presented. Also, as a first step of validation, the code with the new capability is applied to the simulation of experiments investigating cladding behaviour under LOCA conditions. The comparison of the results with the available experimental data of cladding failure due to burst is presented.« less

Th/U-233 multi-recycle in pressurized water reactors : feasibility study of multiple homogeneous and heterogeneous assembly designs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yun, D.; Taiwo, T. A.; Kim, T. K.

2010-10-01

The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle. The possibility for thorium utilization in a multi-recycle system has also been considered in past literature, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this study is to evaluatemore » the potential of Th/U-233 fuel multi-recycle in current LWRs, focusing on pressurized water reactors (PWRs). Approaches for sustainable multi-recycle without the need for external fissile material makeup have been investigated. The intent is to obtain a design that allows existing PWRs to be used with minimal modifications.« less

Silicon carbide composite for light water reactor fuel assembly applications

NASA Astrophysics Data System (ADS)

Yueh, Ken; Terrani, Kurt A.

2014-05-01

The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

NASA Astrophysics Data System (ADS)

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

2015-06-01

FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

Coupling procedure for TRANSURANUS and KTF codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jimenez, J.; Alglave, S.; Avramova, M.

2012-07-01

The nuclear industry aims to ensure safe and economic operation of each single fuel rod introduced in the reactor core. This goal is even more challenging nowadays due to the current strategy of going for higher burn-up (fuel cycles of 18 or 24 months) and longer residence time. In order to achieve that goal, fuel modeling is the key to predict the fuel rod behavior and lifetime under thermal and pressure loads, corrosion and irradiation. In this context, fuel performance codes, such as TRANSURANUS, are used to improve the fuel rod design. The modeling capabilities of the above mentioned toolsmore » can be significantly improved if they are coupled with a thermal-hydraulic code in order to have a better description of the flow conditions within the rod bundle. For LWR applications, a good representation of the two phase flow within the fuel assembly is necessary in order to have a best estimate calculation of the heat transfer inside the bundle. In this paper we present the coupling methodology of TRANSURANUS with KTF (Karlsruhe Two phase Flow subchannel code) as well as selected results of the coupling proof of principle. (authors)« less

Analysis and Implementation of Accident Tolerant Nuclear Fuels

NASA Astrophysics Data System (ADS)

Prewitt, Benjamin Joseph

To improve the reliability and robustness of LWR, accident tolerant nuclear fuels and cladding materials are being developed to possibly replace the current UO2/zirconium system. This research highlights UN and U3Si 2, two of the most favorable accident tolerant fuels being developed. To evaluate the commercial feasiblilty of these fuels, two areas of research were conducted. Chemical fabrication routes for both fuels were investigated in detail, considering UO2 and UF6 as potential starting materials. Potential pathways for industrial scale fabrication using these methods were discussed. Neutronic performance of 70%UN-30%U3Si2 composite was evaluated in MNCP using PWR assembly and core models. The results showed comparable performance to an identical UO2 fueled simulation with the same configuration. The parameters simulated for composite and oxide fuel include the following: fuel to moderator ratio curves; energy dependent flux spectra; temperature coefficients for fuel and moderator; delayed neutron fractions; power peaking factors; axial and radial flux profiles in 2D and 3D; burnup; critical boron concentration; and shutdown margin. Overall, the neutronic parameters suggest that the transition from UO2 to composite in existing nuclear systems will not require significant changes in operating procedures or modifications to standards and regulations.

Dry-vault storage of spent fuel at the CASCAD facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baillif, L.; Guay, M.

A new modular dry storage vault concept using vertical metallic wells cooled by natural convection has been developed by the Commissariat a l'Energie Atomique and Societe Generale pour les Techniques Nouvelles to accommodate special fuels for high-level wastes. Basic specifications and design criteria have been followed to guarantee a double containment system and cooling to maintain the fuel below an acceptable temperature. The double containment is provided by two static barriers: At the reactor, fuels are placed in containers playing the role of the first barrier; the storage wells constitute the second barrier. Spent fuel placed in wells is cooledmore » by natural convection: a boundary layer is created along the outer side of the well. The heated air rises along the well leading to a thermosiphon flow that extracts the heat released. For heat transfer, studies, computations, and experimental tests have been carried out to calculate and determine the temperature of the containers and the fuel rod temperatures in various situations. The CASCAD vault storage can be applied to light water reactor (LWR) fuels without any difficulties if two requirements are satisfied: (1) Spend fuels have to be inserted in tight canisters. (2) Spent fuels have to be received only after a minimum decay time of 5 yr.« less

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

2015-09-01

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirementsmore » for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.« less

LWRS ATR Irradiation Testing Readiness Status

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kristine Barrett

2012-09-01

The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R&D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Testmore » Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics« less

Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N.

2012-07-01

The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existingmore » facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)« less

Burnup calculations and chemical analysis of irradiated fuel samples studied in LWR-PROTEUS phase II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grimm, P.; Guenther-Leopold, I.; Berger, H. D.

2006-07-01

The isotopic compositions of 5 UO{sub 2} samples irradiated in a Swiss PWR power plant, which were investigated in the LWR-PROTEUS Phase II programme, were calculated using the CASMO-4 and BOXER assembly codes. The burnups of the samples range from 50 to 90 MWd/kg. The results for a large number of actinide and fission product nuclides were compared to those of chemical analyses performed using a combination of chromatographic separation and mass spectrometry. A good agreement of calculated and measured concentrations is found for many of the nuclides investigated with both codes. The concentrations of the Pu isotopes are mostlymore » predicted within {+-}10%, the two codes giving quite different results, except for {sup 242}Pu. Relatively significant deviations are found for some isotopes of Cs and Sm, and large discrepancies are observed for Eu and Gd. The overall quality of the predictions by the two codes is comparable, and the deviations from the experimental data do not generally increase with burnup. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yan, Pei-Yang; Zhang, Guojing; Gullickson, Eric M.

Extreme ultraviolet lithography (EUVL) mask multi-layer (ML) blank surface roughness specification historically comes from blank defect inspection tool requirement. Later, new concerns on ML surface roughness induced wafer pattern line width roughness (LWR) arise. In this paper, we have studied wafer level pattern LWR as a function of EUVL mask surface roughness via High-NA Actinic Reticle Review Tool. We found that the blank surface roughness induced LWR at current blank roughness level is in the order of 0.5nm 3σ for NA=0.42 at the best focus. At defocus of ±40nm, the corresponding LWR will be 0.2nm higher. Further reducing EUVL maskmore » blank surface roughness will increase the blank cost with limited benefit in improving the pattern LWR, provided that the intrinsic resist LWR is in the order of 1nm and above.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Carmack; L. Braase; F. Goldner

The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors, enhance proliferation resistance of nuclear fuel, effectively utilize nuclear energy resources, and address the longer-term waste management challenges. This includes development of a state of the art Research and Development (R&D) infrastructure to support the use of a “goal oriented science based approach.” AFC uses a “goal oriented, science based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performancemore » under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. One of the most challenging aspects of AFC is the management, integration, and coordination of major R&D activities across multiple organizations. AFC interfaces and collaborates with Fuel Cycle Technologies (FCT) campaigns, universities, industry, various DOE programs and laboratories, federal agencies (e.g., Nuclear Regulatory Commission [NRC]), and international organizations. Key challenges are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Challenged with the research and development of fuels for two different reactor technology platforms, AFC targeted transmutation fuel development and focused ceramic fuel development for Advanced LWR Fuels.« less

Default operational intervention levels (OILs) for severe nuclear power plant or spent fuel pool emergencies.

PubMed

McKenna, T; Kutkov, V; Vilar Welter, P; Dodd, B; Buglova, E

2013-05-01

Experience and studies show that for an emergency at a nuclear power plant involving severe core damage or damage to the fuel in spent fuel pools, the following actions may need to be taken in order to prevent severe deterministic health effects and reduce stochastic health effects: (1) precautionary protective actions and other response actions for those near the facility (i.e., within the zones identified by the International Atomic Energy Agency) taken immediately upon detection of facility conditions indicating possible severe damage to the fuel in the core or in the spent fuel pool; and (2) protective actions and other response actions taken based on environmental monitoring and sampling results following a release. This paper addresses the second item by providing default operational intervention levels [OILs, which are similar to the U.S. derived response levels (DRLs)] for promptly assessing radioactive material deposition, as well as skin, food, milk and drinking water contamination, following a major release of fission products from the core or spent fuel pool of a light water reactor (LWR) or a high power channel reactor (RBMK), based on the International Atomic Energy Agency's guidance.

PIE of nuclear grade SiC/SiC flexural coupons irradiated to 10 dpa at LWR temperature

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koyanagi, Takaaki; Katoh, Yutai

Silicon carbide fiber-reinforced SiC matrix (SiC/SiC) composites are being actively investigated for accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this study examined SiC/SiC composites following neutron irradiation at 230–340°C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials are chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC)-coated Hi-NicalonTM Type-S (HNS), TyrannoTM SA3 (SA3), and SCS-Ultra TM (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexuralmore » behavior, dynamic Young’s modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young’s moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. This study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.« less

Durability test on irradiated rock-like oxide fuels

NASA Astrophysics Data System (ADS)

Kuramoto, K.; Nitani, N.; Yamashita, T.

2003-06-01

For a profitable use of Pu, Japan Atomic Energy Research Institute has been promoting researches for once-through type fuels. The strategy consists of stable rock-like oxide fuel fabrication in conventional fuel facilities followed by almost complete Pu burning in LWR and disposal of chemically stable spent fuel without further processing. Because leach rates of hazardous nuclides, such as TRU and β-emitters, that have long half-lives, are very important for the evaluation of geological safety, leaching tests in deionized water at 363 K were performed with reference to the MCC-1 method. Five irradiated fuel pellets, a single phase fuel of a yttria-stabilized zirconia (YSZ) containing UO 2 (U-YSZ), two fuels of U-YSZ particle dispersed in MgAl 2O 4 (SPI) or Al 2O 3 (COR) matrix, two homogeneous-blended fuels of U-YSZ and SPI or COR powders, were submitted to the tests. Stainless steel containers with Au coating and ethylene propylene diene monomer were used as leaching vessels and packing, respectively. The evaluated normalized leach rates of Zr, U and Pu were obviously lower than those of the other important elements and nuclides. Americium, Np and especially Y showed unexpectedly high evaluated normalized leach rates. The volatile elements, Cs and I, showed enhanced leaching within particle-dispersed type fuels because of crack formation around the particle.

Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 5. Appendices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1976-05-01

Volume V of the five-volume report consists of appendices, which provide supplementary information, with emphasis on characteristics of geologic formations that might be used for final storage or disposal. Appendix titles are: selected glossary; conversion factors; geologic isolation, including, (a) site selection factors for repositories of wastes in geologic media, (b) rock types--geologic occurrence, (c) glossary of geohydrologic terms, and (d) 217 references; the ocean floor; and, government regulations pertaining to the management of radioactive materials. (JGB)

Method and apparatus for controlled size distribution of gel microspheres formed from aqueous dispersions. [Patent application

DOEpatents

Ryon, A.D.; Haas, P.A.; Vavruska, J.S.

1982-01-19

The present invention is directed to a method and apparatus for making a population of dense, closely size-controlled microspheres by sol-gel procedures wherein said microspheres are characterized by a significant percentage of said population being within a predetermined, relatively narrow size range. This is accomplished by subjecting aqueous dispersions of a sol, within a water-immiscible organic liquid to a turbulent flow. Microsphere populations thus provided are useful in vibratory-packed processes for nuclear fuels to be irradiated in LWR- and FBR-type nuclear reactors.

Continuum model for hydrogen pickup in zirconium alloys of LWR fuel cladding

NASA Astrophysics Data System (ADS)

Wang, Xing; Zheng, Ming-Jie; Szlufarska, Izabela; Morgan, Dane

2017-04-01

A continuum model for calculating the time-dependent hydrogen pickup fractions in various Zirconium alloys under steam and pressured water oxidation has been developed in this study. Using only one fitting parameter, the effective hydrogen gas partial pressure at the oxide surface, a qualitative agreement is obtained between the predicted and previously measured hydrogen pickup fractions. The calculation results therefore demonstrate that H diffusion through the dense oxide layer plays an important role in the hydrogen pickup process. The limitations and possible improvement of the model are also discussed.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

DOE PAGES

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; ...

2015-03-19

FeCrAl is an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In our study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. Also, the total tritium inventory insidemore » the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Konyashov, Vadim V.; Krasnov, Alexander M.

Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. Anmore » approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)« less

Summary of BISON Development and Validation Activities - NEAMS FY16 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, R. L.; Pastore, G.; Gamble, K. A.

This summary report contains an overview of work performed under the work package en- titled “FY2016 NEAMS INL-Engineering Scale Fuel Performance (BISON)” A first chapter identifies the specific FY-16 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes major additional accomplishments, which in- clude: 1) publication of a journal article on solution verification and validation of BISON for LWR fuel, 2) publication of a journal article on 3D Missing Pellet Surface (MPS) analysis of BWR fuel, 3) use of BISON to designmore » a unique 3D MPS validation experiment for future in- stallation in the Halden research reactor, 4) participation in an OECD benchmark on Pellet Clad Mechanical Interaction (PCMI), 5) participation in an OECD benchmark on Reactivity Insertion Accident (RIA) analysis, 6) participation in an OECD activity on uncertainity quantification and sensitivity analysis in nuclear fuel modeling and 7) major improvements to BISON’s fission gas behavior models. A final chapter outlines FY-17 future work.« less

Ion irradiation testing and characterization of FeCrAl candidate alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anderoglu, Osman; Aydogan, Eda; Maloy, Stuart Andrew

2014-10-29

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commerciallymore » available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.« less

Performance and economics analysis of several laser fusion breeder fueled electricity generation systems

NASA Astrophysics Data System (ADS)

Berwald, D. H.; Maniscalco, J. A.

1981-01-01

The paper evaluates the potential of several future electricity generating systems composed of laser fusion-driven breeder reactors that provide fissile fuel for current technology light water fission power reactors (LWRs). The performance and economic feasibility of four fusion breeder blanket technologies for laser fusion drivers, namely uranium fast fission (UFF) blankets, uranium-thorium fast fission (UTFF) blankets, thorium fast fission (TFF) blankets and thorium-suppressed fission (TSF) blankets, are considered, including design and costs of two kinds, fixed (indirect) costs associated with plant capital and variable (direct) costs associated with fuel processing and operation and maintenance. Results indicate that the UTFF and TFF systems produce electricity most inexpensively and that any of the four breeder blanket concepts, including the TSF and UFF systems, can produce electricity for about 25 to 33% above the cost of electricity produced by a new LWR operating on the current once-through cycle. It is suggested that fusion breeders could supply most or all of our fissile fuel makeup requirements within about 20 years after commercial introduction.

Effects of the shape anisotropy and biasing field on the magnetization reversal process of the diamond-shaped NiFe nano films

NASA Astrophysics Data System (ADS)

Xu, Sichen; Yin, Jianfeng; Tang, Rujun; Zhang, Wenxu; Peng, Bin; Zhang, Wanli

2017-11-01

The effects of the planar shape anisotropy and biasing field on the magnetization reversal process (MRP) of the diamond-shaped NiFe nano films have been investigated by micromagnetic simulations. Results show that when the length to width ratio (LWR) of the diamond-shaped film is small, the MRP of the diamond-shaped films are sensitive to LWR. But when LWR is larger than 2, a stable domain switching mode is observed which nucleates from the center of the diamond and then expands to the edges. At a fixed LWR, the magnitude of the switching fields decrease with the increase of the biasing field, but the domain switching mode is not affected by the biasing field. Further analysis shows that demagnetization energy dominates over the MRP of the diamond-shaped films. The above LWR dependence of MRP can be well explained by a variation of the shape anisotropic factor with LWR.

Development of an integrated, unattended assay system for LWR-MOX fuel pellet trays

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stewart, J.E.; Hatcher, C.R.; Pollat, L.L.

1994-08-01

Four identical unattended plutonium assay systems have been developed for use at the new light-water-reactor mixed oxide (LWR-MOX) fuel fabrication facility at Hanau, Germany. The systems provide quantitative plutonium verification for all MOX pellet trays entering or leaving a large, intermediate store. Pellet-tray transport and storage systems are highly automated. Data from the ``I-Point`` (information point) assay systems will be shared by the Euratom and International Atomic Energy Agency (IAEA) Inspectorates. The I-Point system integrates, for the first time, passive neutron coincidence counting (NCC) with electro-mechanical sensing (EMS) in unattended mode. Also, provisions have been made for adding high-resolution gammamore » spectroscopy. The system accumulates data for every tray entering or leaving the store between inspector visits. During an inspection, data are analyzed and compared with operator declarations for the previous inspection period, nominally one month. Specification of the I-point system resulted from a collaboration between the IAEA, Euratom, Siemens, and Los Alamos. Hardware was developed by Siemens and Los Alamos through a bilateral agreement between the German Federal Ministry of Research and Technology (BMFT) and the US DOE. Siemens also provided the EMS subsystem, including software. Through the USSupport Program to the IAEA, Los Alamos developed the NCC software (NCC COLLECT) and also the software for merging and reviewing the EMS and NCC data (MERGE/REVIEW). This paper describes the overall I-Point system, but emphasizes the NCC subsystem, along with the NCC COLLECT and MERGE/REVIEW codes. We also summarize comprehensive testing results that define the quality of assay performance.« less

Critical processes and parameters in the development of accident tolerant fuels drop-in capsule irradiation tests

DOE PAGES

Barrett, K. E.; Ellis, K. D.; Glass, C. R.; ...

2015-12-01

The goal of the Accident Tolerant Fuel (ATF) program is to develop the next generation of Light Water Reactor (LWR) fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. An irradiation test series has been defined to assess the performance of proposed ATF concepts under normal LWR operating conditions. The Phase I ATF irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL). Design, analysis, and fabrication processes formore » ATR drop-in capsule experiment preparation are presented in this paper to demonstrate the importance of special design considerations, parameter sensitivity analysis, and precise fabrication and inspection techniques for figure innovative materials used in ATF experiment assemblies. A Taylor Series Method sensitivity analysis approach was used to identify the most critical variables in cladding and rodlet stress, temperature, and pressure calculations for design analyses. The results showed that internal rodlet pressure calculations are most sensitive to the fission gas release rate uncertainty while temperature calculations are most sensitive to cladding I.D. and O.D. dimensional uncertainty. The analysis showed that stress calculations are most sensitive to rodlet internal pressure uncertainties, however the results also indicated that the inside radius, outside radius, and internal pressure were all magnified as they propagate through the stress equation. This study demonstrates the importance for ATF concept development teams to provide the fabricators as much information as possible about the material properties and behavior observed in prototype testing, mock-up fabrication and assembly, and chemical and mechanical testing of the materials that may have been performed in the concept development phase. Special handling, machining, welding, and inspection of materials, if known, should also be communicated to the experiment fabrication and inspection team.« less

Fuel Cycle System Analysis Handbook

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven J. Piet; Brent W. Dixon; Dirk Gombert

2009-06-01

This Handbook aims to improve understanding and communication regarding nuclear fuel cycle options. It is intended to assist DOE, Campaign Managers, and other presenters prepare presentations and reports. When looking for information, check here. The Handbook generally includes few details of how calculations were performed, which can be found by consulting references provided to the reader. The Handbook emphasizes results in the form of graphics and diagrams, with only enough text to explain the graphic, to ensure that the messages associated with the graphic is clear, and to explain key assumptions and methods that cause the graphed results. Some ofmore » the material is new and is not found in previous reports, for example: (1) Section 3 has system-level mass flow diagrams for 0-tier (once-through), 1-tier (UOX to CR=0.50 fast reactor), and 2-tier (UOX to MOX-Pu to CR=0.50 fast reactor) scenarios - at both static and dynamic equilibrium. (2) To help inform fast reactor transuranic (TRU) conversion ratio and uranium supply behavior, section 5 provides the sustainable fast reactor growth rate as a function of TRU conversion ratio. (3) To help clarify the difference in recycling Pu, NpPu, NpPuAm, and all-TRU, section 5 provides mass fraction, gamma, and neutron emission for those four cases for MOX, heterogeneous LWR IMF (assemblies mixing IMF and UOX pins), and a CR=0.50 fast reactor. There are data for the first 10 LWR recycle passes and equilibrium. (4) Section 6 provides information on the cycle length, planned and unplanned outages, and TRU enrichment as a function of fast reactor TRU conversion ratio, as well as the dilution of TRU feedstock by uranium in making fast reactor fuel. (The recovered uranium is considered to be more pure than recovered TRU.) The latter parameter impacts the required TRU impurity limits specified by the Fuels Campaign. (5) Section 7 provides flows for an 800-tonne UOX separation plant. (6) To complement 'tornado' economic uncertainty diagrams, which show at a glance combined uncertainty information, section 9.2 has a new set of simpler graphs that show the impact on fuel cycle costs for once through, 1-tier, and 2-tier scenarios as a function of key input parameters.« less

Transuranic Waste Burning Potential of Thorium Fuel in a Fast Reactor - 12423

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wenner, Michael; Franceschini, Fausto; Ferroni, Paolo

Westinghouse Electric Company (referred to as 'Westinghouse' in the rest of this paper) is proposing a 'back-to-front' approach to overcome the stalemate on nuclear waste management in the US. In this approach, requirements to further the societal acceptance of nuclear waste are such that the ultimate health hazard resulting from the waste package is 'as low as reasonably achievable'. Societal acceptability of nuclear waste can be enhanced by reducing the long-term radiotoxicity of the waste, which is currently driven primarily by the protracted radiotoxicity of the transuranic (TRU) isotopes. Therefore, a transition to a more benign radioactive waste can bemore » accomplished by a fuel cycle capable of consuming the stockpile of TRU 'legacy' waste contained in the LWR Used Nuclear Fuel (UNF) while generating waste which is significantly less radio-toxic than that produced by the current open U-based fuel cycle (once through and variations thereof). Investigation of a fast reactor (FR) operating on a thorium-based fuel cycle, as opposed to the traditional uranium-based is performed. Due to a combination between its neutronic properties and its low position in the actinide chain, thorium not only burns the legacy TRU waste, but it does so with a minimal production of 'new' TRUs. The effectiveness of a thorium-based fast reactor to burn legacy TRU and its flexibility to incorporate various fuels and recycle schemes according to the evolving needs of the transmutation scenario have been investigated. Specifically, the potential for a high TRU burning rate, high U-233 generation rate if so desired and low concurrent production of TRU have been used as metrics for the examined cycles. Core physics simulations of a fast reactor core running on thorium-based fuels and burning an external TRU feed supply have been carried out over multiple cycles of irradiation, separation and reprocessing. The TRU burning capability as well as the core isotopic content have been characterized. Results will be presented showing the potential for thorium to reach a high TRU transmutation rate over a wide variety of fuel types (oxide, metal, nitride and carbide) and transmutation schemes (recycle or partition of in-bred U-233). In addition, a sustainable scheme has been devised to burn the TRU accumulated in the core inventory once the legacy TRU supply has been exhausted, thereby achieving long-term virtually TRU-free. A comprehensive 'back-to-front' approach to the fuel cycle has recently been proposed by Westinghouse which emphasizes achieving 'acceptable', low-radiotoxicity, high-level waste, with the intent not only to satisfy all technical constraints but also to improve public acceptance of nuclear energy. Following this approach, the thorium fuel cycle, due to its low radiotoxicity and high potential for TRU transmutation has been selected as a promising solution. Additional studies not shown here have shown significant reduction of decay heat. The TRU burning potential of the Th-based fuel cycle has been illustrated with a variety of fuel types, using the Toshiba ARR to perform the analysis, including scenarios with continued LWR operation of either uranium fueled or thorium fueled LWRs. These scenarios will afford overall reduction in actinide radiotoxicity, however when the TRU supply is exhausted, a continued U- 235 LWR operation must be assumed to provide TRU makeup feed. This scenario will never reach the characteristically low TRU content of a closed thorium fuel cycle with its associated potential benefits on waste radiotoxicity, as exemplified by the transition scenario studied. At present, the cases studied indicate ThC as a potential fuel for maximizing TRU burning, while ThN with nitrogen enriched to 95% N-15 shows the highest breeding potential. As a result, a transition scenario with ThN was developed to show that a sustainable, closed Th-cycle can be achieved starting from burning the legacy TRU stock and completing the transmutation of the residual TRU remaining in the core inventory after the legacy TRU external supply has been exhausted. The radiotoxicity of the actinide waste during the various phases has been characterized, showing the beneficial effect of the decreasing content of TRU in the recycled fuel as the transition to a closed Th-based fuel cycle is undertaken. Due to the back-to-front nature of the proposed methodology, detailed designs are not the first step taken when assessing a fuel cycle scenario potential. As a result, design refinement is still required and should be expected in future studies. Moreover, significant safety assessment, including determination of associated reactivity coefficients, fuel and reprocessing feasibility studies and economic assessments will still be needed for a more comprehensive and meaningful comparison against other potential solutions. With the above considerations in mind, the potential advantages of thorium fuelled reactors on HLW management optimization lead us to believe that thorium fuelled reactor systems can play a significant role in the future and deserve further consideration. (authors)« less

A high burnup model developed for the DIONISIO code

NASA Astrophysics Data System (ADS)

Soba, A.; Denis, A.; Romero, L.; Villarino, E.; Sardella, F.

2013-02-01

A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO2 fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in 235U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

Fabrication of light water reactor tritium targets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pilger, J.P.

1991-11-01

The mission of the Fabrication Development Task of the Tritium Target Development Project is: to produce a documented technology basis, including specifications and procedures for target rod fabrication; to demonstrate that light water tritium targets can be manufactured at a rate consistent with tritium production requirements; and to develop quality control methods to evaluate target rod components and assemblies, and establish correlations between evaluated characteristics and target rod performance. Many of the target rod components: cladding tubes, end caps, plenum springs, etc., have similar counterparts in LWR fuel rods. High production rate manufacture and inspection of these components has beenmore » adequately demonstrated by nuclear fuel rod manufacturers. This summary describes the more non-conventional manufacturing processes and inspection techniques developed to fabricate target rod components whose manufacturability at required production rates had not been previously demonstrated.« less

Ultra-High Temperature Steam Corrosion of Complex Silicates for Nuclear Applications: A Computational Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rashkeev, Sergey N.; Glazoff, Michael V.; Tokuhiro, Akira

2014-01-01

Stability of materials under extreme conditions is an important issue for safety of nuclear reactors. Presently, silicon carbide (SiC) is being studied as a cladding material candidate for fuel rods in boiling-water and pressurized water-cooled reactors (BWRs and PWRs) that would substitute or modify traditional zircaloy materials. The rate of corrosion of the SiC ceramics in hot vapor environment (up to 2200 degrees C) simulating emergency conditions of light water reactor (LWR) depends on many environmental factors such as pressure, temperature, viscosity, and surface quality. Using the paralinear oxidation theory developed for ceramics in the combustion reactor environment, we estimatedmore » the corrosion rate of SiC ceramics under the conditions representing a significant power excursion in a LWR. It was established that a significant time – at least 100 h – is required for a typical SiC braiding to significantly degrade even in the most aggressive vapor environment (with temperatures up to 2200 °C) which is possible in a LWR at emergency condition. This provides evidence in favor of using the SiC coatings/braidings for additional protection of nuclear reactor rods against off-normal material degradation during power excursions or LOCA incidents. Additionally, we discuss possibilities of using other silica based ceramics in order to find materials with even higher corrosion resistance than SiC. In particular, we found that zircon (ZrSiO4) is also a very promising material for nuclear applications. Thermodynamic and first-principles atomic-scale calculations provide evidence of zircon thermodynamic stability in aggressive environments at least up to 1535 degrees C.« less

Data summary report for fission product release test VI-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, M.F.; Collins, J.L.; Lorenz, R.A.

The first in a series of high-temperature fission product release test in a new vertical test apparatus was conducted in flowing steam. The test specimen was a 15.2-cm-long section of a fuel rod from the Oconee 1 PWR; it had been irradiated to a burnup of /approximately/42 MWd/kg. Using an induction furnace, it was heated under simulated LWR accident conditions -- 20 min at 2000 K and 20 min at 2300 K -- in a hot cell-mounted test apparatus. Posttest inspection showed severe oxidation but only minimal fragmentation of the fuel specimen; cladding melting was apparent only near the topmore » end. Based on fission product measured in the fuel and/or calculated by ORIGEN, analyses of test components showed total releases from the fuel of 47% for /sup 85/Kr, 33% for /sup 125/Sb, 37% for /sup 129/I, 84% for /sup 110m/Ag, and 63% for /sup 137/Cs. Large fractions (36% and 30%, respectively) of the released /sup 110m/Ag and /sup 125/Sb were retained in the furnace above the fuel. Pretest and posttest analysis of the fuel specimen indicated a /sup 134/Cs release of 65%, which is very good agreement with the /sup 137/Cs value. 21 refs., 24 figs., 16 tabs.« less

Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrani, Kurt A; Kiggans Jr, James O; McMurray, Jake W

2016-01-01

Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhancemore » heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasiblemore » to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)« less

Novel EUV photoresist for sub-7nm node (Conference Presentation)

NASA Astrophysics Data System (ADS)

Furukawa, Tsuyoshi; Naruoka, Takehiko; Nakagawa, Hisashi; Miyata, Hiromu; Shiratani, Motohiro; Hori, Masafumi; Dei, Satoshi; Ayothi, Ramakrishnan; Hishiro, Yoshi; Nagai, Tomoki

2017-04-01

Extreme ultraviolet (EUV) lithography has been recognized as a promising candidate for the manufacturing of semiconductor devices as LS and CH pattern for 7nm node and beyond. EUV lithography is ready for high volume manufacturing stage. For the high volume manufacturing of semiconductor devices, significant improvement of sensitivity and line edge roughness (LWR) and Local CD Uniformity (LCDU) is required for EUV resist. It is well-known that the key challenge for EUV resist is the simultaneous requirement of ultrahigh resolution (R), low line edge roughness (L) and high sensitivity (S). Especially high sensitivity and good roughness is important for EUV lithography high volume manufacturing. We are trying to improve sensitivity and LWR/LCDU from many directions. From material side, we found that both sensitivity and LWR/LCDU are simultaneously improved by controlling acid diffusion length and efficiency of acid generation using novel resin and PAG. And optimizing EUV integration is one of the good solution to improve sensitivity and LWR/LCDU. We are challenging to develop new multi-layer materials to improve sensitivity and LWR/LCDU. Our new multi-layer materials are designed for best performance in EUV lithography system. From process side, we found that sensitivity was substantially improved maintaining LWR applying novel type of chemical amplified resist (CAR) and process. EUV lithography evaluation results obtained for new CAR EUV interference lithography. And also metal containing resist is one possibility to break through sensitivity and LWR trade off. In this paper, we will report the recent progress of sensitivity and LWR/LCDU improvement of JSR novel EUV resist and process.

Recent Progress in Using Advanced Characterization and Modeling Approaches to Study Radiation Effects in Oxide Ceramics

DOE PAGES

Bai, Xian-Ming

2014-10-23

I serve as a Guest Editor for the Nuclear Materials Committee of the TMS Structural Materials Division, and coordinated the topic ‘‘Radiation Effects in Oxide Ceramics and Novel LWR Fuels" for JOM in the December 2014 issue. I selected five articles related this topic. These articles talk about some recent progress of using advanced experimental and modeling tools to study radiation effects in oxide ceramics at atomistic scale and mesoscale. In this guest editor commentary article, I summarize the novel aspects of these papers and also provide some suggestions for future research directions.

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pope, M. A.; Sen, R. S.; Ougouag, A. M.

2012-07-01

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space availablemore » for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)« less

Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag

2012-04-01

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel,more » the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.« less

Line-width roughness of advanced semiconductor features by using FIB and planar-TEM as reference metrology

NASA Astrophysics Data System (ADS)

Takamasu, Kiyoshi; Takahashi, Satoru; Kawada, Hiroki; Ikota, Masami

2018-03-01

LER (Line Edge Roughness) and LWR (Line Width Roughness) of the semiconductor device are an important evaluation scale of the performance of the device. Conventionally, LER and LWR is evaluated from CD-SEM (Critical Dimension Scanning Electron Microscope) images. However, CD-SEM measurement has a problem that high frequency random noise is large, and resolution is not sufficiently high. For random noise of CD-SEM measurement, some techniques are proposed. In these methods, it is necessary to set parameters for model and processing, and it is necessary to verify the correctness of these parameters using reference metrology. We have already proposed a novel reference metrology using FIB (Focused Ion Beam) process and planar-TEM (Transmission Electron Microscope) method. In this study, we applied the proposed method to three new samples such as SAQP (Self-Aligned Quadruple Patterning) FinFET device, EUV (Extreme Ultraviolet Lithography) conventional resist, and EUV new material resist. LWR and PSD (Power Spectral Density) of LWR are calculated from the edge positions on planar-TEM images. We confirmed that LWR and PSD of LWR can be measured with high accuracy and evaluated the difference by the proposed method. Furthermore, from comparisons with PSD of the same sample by CD-SEM, the validity of measurement of PSD and LWR by CD-SEM can be verified.

Nuclear fuel requirements for the American economy - A model

NASA Astrophysics Data System (ADS)

Curtis, Thomas Dexter

A model is provided to determine the amounts of various fuel streams required to supply energy from planned and projected nuclear plant operations, including new builds. Flexible, user-defined scenarios can be constructed with respect to energy requirements, choices of reactors and choices of fuels. The model includes interactive effects and extends through 2099. Outputs include energy provided by reactors, the number of reactors, and masses of natural Uranium and other fuels used. Energy demand, including electricity and hydrogen, is obtained from US DOE historical data and projections, along with other studies of potential hydrogen demand. An option to include other energy demand to nuclear power is included. Reactor types modeled include (thermal reactors) PWRs, BWRs and MHRs and (fast reactors) GFRs and SFRs. The MHRs (VHTRs), GFRs and SFRs are similar to those described in the 2002 DOE "Roadmap for Generation IV Nuclear Energy Systems." Fuel source choices include natural Uranium, self-recycled spent fuel, Plutonium from breeder reactors and existing stockpiles of surplus HEU, military Plutonium, LWR spent fuel and depleted Uranium. Other reactors and fuel sources can be added to the model. Fidelity checks of the model's results indicate good agreement with historical Uranium use and number of reactors, and with DOE projections. The model supports conclusions that substantial use of natural Uranium will likely continue to the end of the 21st century, though legacy spent fuel and depleted uranium could easily supply all nuclear energy demand by shifting to predominant use of fast reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, Jon; Hayes, Steven; Walters, L. C.

This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO 2 and UO 2-PuO 2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availabilitymore » are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.« less

Difference in EUV photoresist design towards reduction of LWR and LCDU

NASA Astrophysics Data System (ADS)

Jiang, Jing; De Simone, Danilo; Vandenberghe, Geert

2017-03-01

Pattern fidelity of EUV lithography is crucial for high resolution features, since small variation can affect device performance and even cause short or open circuit. For 1D features, dense lines and contact holes are the most common features for active, metal and contact layer, therefore line width roughness (LWR) and local critical dimension uniformity (LCDU) are important indexes to monitor. Both LWR and LCDU are greatly influenced by photon and acid shot noise. In addition, LWR is also affected by resist mechanical properties, like pattern collapse. In this study, we studied the influence of different chemically amplified resist components, such as polymer, PAG and quencher for both types and concentrations in order to understand the relative extent of influences of deprotection, acid diffusion, and base neutralization on pattern fidelity. However, conventional methods to approach higher resolution or low LWR/LCDU by sacrificing the dose are not sustainable. In order to continue to improve resist performance, a new component, metal salt sensitizer, is introduced into the resist system. This metal salt is able to achieve 30% dose reduction by increasing EUV absorption, maintaining LWR. We believe metal sensitizer might give us a new way to challenge the RLS trade-off.

Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miao, Yinbin; Gamble, Kyle A.; Andersson, David

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety ofmore » LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.« less

Self-Sustaining Thorium Boiling Water Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra

The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare themore » RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.« less

Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beyer, Brian David; Beddingfield, David H; Durst, Philip

2010-01-01

The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguardsmore » criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.« less

A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

NASA Astrophysics Data System (ADS)

Schneider, E. A.; Deinert, M. R.; Cady, K. B.

2006-10-01

The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

The need for LWR metrology standardization: the imec roughness protocol

NASA Astrophysics Data System (ADS)

Lorusso, Gian Francesco; Sutani, Takumichi; Rutigliani, Vito; van Roey, Frieda; Moussa, Alain; Charley, Anne-Laure; Mack, Chris; Naulleau, Patrick; Constantoudis, Vassilios; Ikota, Masami; Ishimoto, Toru; Koshihara, Shunsuke

2018-03-01

As semiconductor technology keeps moving forward, undeterred by the many challenges ahead, one specific deliverable is capturing the attention of many experts in the field: Line Width Roughness (LWR) specifications are expected to be less than 2nm in the near term, and to drop below 1nm in just a few years. This is a daunting challenge and engineers throughout the industry are trying to meet these targets using every means at their disposal. However, although current efforts are surely admirable, we believe they are not enough. The fact is that a specification has a meaning only if there is an agreed methodology to verify if the criterion is met or not. Such a standardization is critical in any field of science and technology and the question that we need to ask ourselves today is whether we have a standardized LWR metrology or not. In other words, if a single reference sample were provided, would everyone measuring it get reasonably comparable results? We came to realize that this is not the case and that the observed spread in the results throughout the industry is quite large. In our opinion, this makes the comparison of LWR data among institutions, or to a specification, very difficult. In this paper, we report the spread of measured LWR data across the semiconductor industry. We investigate the impact of image acquisition, measurement algorithm, and frequency analysis parameters on LWR metrology. We review critically some of the International Technology Roadmap for Semiconductors (ITRS) metrology guidelines (such as measurement box length larger than 2μm and the need to correct for SEM noise). We compare the SEM roughness results to AFM measurements. Finally, we propose a standardized LWR measurement protocol - the imec Roughness Protocol (iRP) - intended to ensure that every time LWR measurements are compared (from various sources or to specifications), the comparison is sensible and sound. We deeply believe that the industry is at a point where it is imperative to guarantee that when talking about a critical parameter such like LWR, everyone speaks the same language, which is not currently the case.

Relationship between sensitizer concentration and resist performance of chemically amplified extreme ultraviolet resists in sub-10 nm half-pitch resolution region

NASA Astrophysics Data System (ADS)

Kozawa, Takahiro; Santillan, Julius Joseph; Itani, Toshiro

2017-01-01

The development of lithography processes with sub-10 nm resolution is challenging. Stochastic phenomena such as line width roughness (LWR) are significant problems. In this study, the feasibility of sub-10 nm fabrication using chemically amplified extreme ultraviolet resists with photodecomposable quenchers was investigated from the viewpoint of the suppression of LWR. The relationship between sensitizer concentration (the sum of acid generator and photodecomposable quencher concentrations) and resist performance was clarified, using the simulation based on the sensitization and reaction mechanisms of chemically amplified resists. For the total sensitizer concentration of 0.5 nm-3 and the effective reaction radius for the deprotection of 0.1 nm, the reachable half-pitch while maintaining 10% critical dimension (CD) LWR was 11 nm. The reachable half-pitch was 7 nm for 20% CD LWR. The increase in the effective reaction radius is required to realize the sub-10 nm fabrication with 10% CD LWR.

Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

NASA Astrophysics Data System (ADS)

Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

2018-02-01

SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

Mechanistic prediction of fission-gas behavior during in-cell transient heating tests on LWR fuel using the GRASS-SST and FASTGRASS computer codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J; Gehl, S M

1979-01-01

GRASS-SST and FASTGRASS are mechanistic computer codes for predicting fission-gas behavior in UO/sub 2/-base fuels during steady-state and transient conditions. FASTGRASS was developed in order to satisfy the need for a fast-running alternative to GRASS-SST. Althrough based on GRASS-SST, FASTGRASS is approximately an order of magnitude quicker in execution. The GRASS-SST transient analysis has evolved through comparisons of code predictions with the fission-gas release and physical phenomena that occur during reactor operation and transient direct-electrical-heating (DEH) testing of irradiated light-water reactor fuel. The FASTGRASS calculational procedure is described in this paper, along with models of key physical processes included inmore » both FASTGRASS and GRASS-SST. Predictions of fission-gas release obtained from GRASS-SST and FASTGRASS analyses are compared with experimental observations from a series of DEH tests. The major conclusions is that the computer codes should include an improved model for the evolution of the grain-edge porosity.« less

Synthesis and sintering of UN-UO2 fuel composites

NASA Astrophysics Data System (ADS)

Jaques, Brian J.; Watkins, Jennifer; Croteau, Joseph R.; Alanko, Gordon A.; Tyburska-Püschel, Beata; Meyer, Mitch; Xu, Peng; Lahoda, Edward J.; Butt, Darryl P.

2015-11-01

The design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO2, which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO2 to UN has been suggested. In order to avoid carbon impurities, UN was synthesized from elemental uranium using a hydride-dehydride-nitride thermal synthesis route prior to mixing with up to 10 wt% UO2 in a planetary ball mill. UN and UN - UO2 composite pellets were sintered in Ar - (0-1 at%) N2 to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO2 composite pellets were also sintered in Ar - 100 ppm N2 to assess the effects of temperature (1700-2000 °C) on the final grain morphology and phase concentration.

Production of FR Tubing from Advanced ODS Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maloy, Stuart Andrew; Lavender, Curt; Omberg, Ron

2016-10-25

Significant research is underway to develop LWR nuclear fuels with improved accident tolerance. One of the leading candidate materials for cladding are the FeCrAl alloys. New alloys produced at ORNL called Gen I and Gen II FeCrAl alloys possess excellent oxidation resistance in steam up to 1400°C and in parallel methods are being developed to produce tubing from these alloys. Century tubing continues to produce excellent tubing from FeCrAl alloys. This memo reports receipt of ~21 feet of Gen I FeCrAl alloy tubing. This tubing will be used for future tests including burst testing, mechanical testing and irradiation testing.

Metrics for the technical performance evaluation of light water reactor accident-tolerant fuel

DOE PAGES

Bragg-Sitton, Shannon M.; Todosow, Michael; Montgomery, Robert; ...

2017-03-26

The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident-tolerant fuel (ATF) for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, andmore » economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+), fuels with enhanced accident tolerance would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. Research and development of ATF in the United States is conducted under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Advanced Fuels Campaign. The DOE is sponsoring multiple teams to develop ATF concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This study summarizes the technical evaluation methodology proposed in the United States to aid in the optimization and prioritization of candidate ATF designs.« less

Advanced Instrumentation for Transient Reactor Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corradini, Michael L.; Anderson, Mark; Imel, George

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and designmore » increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).« less

The design and implementation of photoacoustic based laser warning receiver for harsh environments

NASA Astrophysics Data System (ADS)

El-Sherif, Ashraf F.; Ayoub, H. S.; El-Sharkawy, Yasser H.; Gomaa, Walid; Hassan, H. H.

2018-01-01

This paper discusses the implementation of new type of laser warning receiver (LWR) system, based on the detection of photoacoustic signals, induced by high power infrared laser designators pulses on target's surfaces. This system appends conventional optoelectronic based LWR to decrease the false alarm rate (FAR) in harsh environments, where ambient conditions are expected to obstruct optical LWR. To improve the sensitivity of the photoacoustic based LWR system, some metallic and polymeric target shielding materials were studied, in order to cover a friendly civil structure, vehicle or a maritime entity with a low cost large area acoustic detector array shield. A thermographic investigation of target surface material- laser reaction, signal processing and system configuration and functional analysis are also presented.

Stationary Liquid Fuel Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, Won Sik; Grandy, Andrew; Boroski, Andrew

For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excessmore » reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel container is penetrated by twelve hexagonal control assembly (CA) guide tubes, each of which has 3.0 mm thickness and 69.4 mm flat-to-flat outer distance. The distance between two neighboring CA guide tube is selected to be 26 cm to provide an adequate space for CA driving systems. The fuel container has 18181 penetrating coolant tubes of 6.0 mm inner diameter and 2.0 mm thickness. The coolant tubes are arranged in a triangular lattice with a lattice pitch of 1.21 cm. The fuel, structure, and coolant volume fractions inside the fuel container are 0.386, 0.383, and 0.231, respectively. Separate steel reflectors and B4C shields are used outside of the fuel container. Six gas expansion modules (GEMs) of 5.0 cm thickness are introduced in the radial reflector region. Between the radial reflector and the fuel container is a 2.5 cm sodium gap. The TRU inventory at the beginning of equilibrium cycle (BOEC) is 5081 kg, whereas the TRU inventory at the beginning of life (BOL) was 3541 kg. This is because the equilibrium cycle fuel contains a significantly smaller fissile fraction than the LWR TRU feed. The fuel inventory at BOEC is composed of 34.0 a/o TRU, 41.4 a/o Ce, 23.6 a/o Co, and 1.03 a/o solid fission products. Since uranium-free fuel is used, a theoretical maximum TRU consumption rate of 1.011 kg/day is achieved. The semi-continuous fuel cycle based on the 300-batch, 1- day cycle approximation yields a burnup reactivity loss of 26 pcm/day, and requires a daily reprocessing of 32.5 kg of SLFFR fuel. This yields a daily TRU charge rate of 17.45 kg, including a makeup TRU feed of 1.011 kg recovered from the LWR used fuel. The charged TRU-Ce-Co fuel is composed of 34.4 a/o TRU, 40.6 a/o Ce, and 25.0 a/o Co.« less

Stochastic Optimization for Nuclear Facility Deployment Scenarios

NASA Astrophysics Data System (ADS)

Hays, Ross Daniel

Single-use, low-enriched uranium oxide fuel, consumed through several cycles in a light-water reactor (LWR) before being disposed, has become the dominant source of commercial-scale nuclear electric generation in the United States and throughout the world. However, it is not without its drawbacks and is not the only potential nuclear fuel cycle available. Numerous alternative fuel cycles have been proposed at various times which, through the use of different reactor and recycling technologies, offer to counteract many of the perceived shortcomings with regards to waste management, resource utilization, and proliferation resistance. However, due to the varying maturity levels of these technologies, the complicated material flow feedback interactions their use would require, and the large capital investments in the current technology, one should not deploy these advanced designs without first investigating the potential costs and benefits of so doing. As the interactions among these systems can be complicated, and the ways in which they may be deployed are many, the application of automated numerical optimization to the simulation of the fuel cycle could potentially be of great benefit to researchers and interested policy planners. To investigate the potential of these methods, a computational program has been developed that applies a parallel, multi-objective simulated annealing algorithm to a computational optimization problem defined by a library of relevant objective functions applied to the Ver ifiable Fuel Cycle Simulati on Model (VISION, developed at the Idaho National Laboratory). The VISION model, when given a specified fuel cycle deployment scenario, computes the numbers and types of, and construction, operation, and utilization schedules for, the nuclear facilities required to meet a predetermined electric power demand function. Additionally, it calculates the location and composition of the nuclear fuels within the fuel cycle, from initial mining through to eventual disposal. By varying the specifications of the deployment scenario, the simulated annealing algorithm will seek to either minimize the value of a single objective function, or enumerate the trade-off surface between multiple competing objective functions. The available objective functions represent key stakeholder values, minimizing such important factors as high-level waste disposal burden, required uranium ore supply, relative proliferation potential, and economic cost and uncertainty. The optimization program itself is designed to be modular, allowing for continued expansion and exploration as research needs and curiosity indicate. The utility and functionality of this optimization program are demonstrated through its application to one potential fuel cycle scenario of interest. In this scenario, an existing legacy LWR fleet is assumed at the year 2000. The electric power demand grows exponentially at a rate of 1.8% per year through the year 2100. Initially, new demand is met by the construction of 1-GW(e) LWRs. However, beginning in the year 2040, 600-MW(e) sodium-cooled, fast-spectrum reactors operating in a transuranic burning regime with full recycling of spent fuel become available to meet demand. By varying the fraction of new capacity allocated to each reactor type, the optimization program is able to explicitly show the relationships that exist between uranium utilization, long-term heat for geologic disposal, and cost-of-electricity objective functions. The trends associated with these trade-off surfaces tend to confirm many common expectations about the use of nuclear power, namely that while overall it is quite insensitive to variations in the cost of uranium ore, it is quite sensitive to changes in the capital costs of facilities. The optimization algorithm has shown itself to be robust and extensible, with possible extensions to many further fuel cycle optimization problems of interest.

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

NASA Astrophysics Data System (ADS)

Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

2016-11-01

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

DOE PAGES

Williamson, R. L.; Capps, N. A.; Liu, W.; ...

2016-09-27

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) ormore » plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used in this paper to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. Finally, in comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.« less

Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bhatti, Zaki; Hyland, B.; Edwards, G.W.R.

2013-07-01

The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in themore » Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)« less

The Use of Thorium within the Nuclear Power Industry - 13472

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Keith

2013-07-01

Thorium is 3 to 4 times more abundant than uranium and is widely distributed in nature as an easily exploitable resource in many countries. Unlike natural uranium, which contains ∼0.7% fissile {sup 235}U isotope, natural thorium does not contain any fissile material and is made up of the fertile {sup 232}Th isotope only. Therefore thorium and thorium-based fuel as metal, oxide or carbide, has been utilized in combination with fissile {sup 235}U or {sup 239}Pu in nuclear research and power reactors for conversion to fissile {sup 233}U, thereby enlarging fissile material resources. During the pioneering years of nuclear energy, frommore » the mid 1950's to mid 1970's, there was considerable interest worldwide to develop thorium fuels and fuel cycles in order to supplement uranium reserves. Thorium fuels and fuel cycles are particularly relevant to countries having large thorium deposits but very limited uranium reserves for their long term nuclear power programme. The feasibility of thorium utilization in high temperature gas cooled reactors (HTGR), light water reactors (LWR), pressurized heavy water reactors (PHWRs), liquid metal cooled fast breeder reactors (LMFBR) and molten salt breeder reactors (MSBR) were demonstrated. The initial enthusiasm for thorium fuels and fuel cycles was not sustained among the developing countries later, due to new discovery of uranium deposits and their improved availability. However, in recent times, the need for proliferation-resistance, longer fuel cycles, higher burnup, and improved waste form characteristics, reduction of plutonium inventories and in situ use of bred-in fissile material has led to renewed interest in thorium-based fuels and fuel cycles. (authors)« less

Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferrandis, J.Y.; Leveque, G.; Rosenkrantz, E.

2015-07-01

First of all, we will present the main principle of the method. A piezoelectric transducer, driven by a pulse generator, generates the acoustic waves in a cavity that may be the fuel rod or a chamber connected to an instrumented rod. The composition determination consists in measuring the time of flight of the acoustic signal emitted. The pressure can be estimated by a calibration process, above the measurement of the amplitude of the signal. Two projects will then be detailed. The first project consists in the development of advanced instrumentation for in-pile experiments in Material Testing Reactor. It constitutes amore » main goal for the improvement of the nuclear fuel behavior knowledge. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). As a first step of the development program, we performed in-pile tests on the most sensitive component, i.e., the piezoelectric transducer. For this purpose, the active part of this sensor has been qualified on gamma and neutron radiations and at high temperature. Various industrial piezo-ceramics were exposed to a high activity Cobalt source for few days. The cumulated dose was ranged from 50 kGy up to 2 MGy. Next, these devices were placed inside a Material Test Reactor to investigate their reliability towards neutron fluence. The final fluence after 150 days of irradiation was up to 1.6.10{sup 21}n/cm{sup 2} (for thermal neutron). Irreversible variations have been measured. Next, a specific sensor has been implemented on an instrumented fuel rod and tested in the frame of a REMORA 3 Irradiation test. It was the first experiment under high mixed, temperature neutron and gamma flux. A first irradiation phase took place in March 2010 in the OSIRIS reactor and in November 2010 for the second step of the irradiation. The instrumented fuel rod incorporating the ultrasonic gas composition sensor was finally irradiated during 2 weeks in nominal conditions. Neutronics calculation will be performed in order to calculate the thermal and fast neutron fluence and the gamma dose absorbed by acoustic sensor. A first evaluation gives a thermal fluence about 4,5.10{sup 19} n/cm{sup 2}, a fast neutrons fluence about 4,5.1018 n/cm{sup 2} and a total gamma dose up to 0,25 MGy The maximal temperature during the irradiation test was about 150 C. Although the ultrasonic sensor appears to be damaged, the optimization of the electrical attack parameters and the development of a new signal processing maintain the measurement feasibility up the end of the irradiation campaign. It was the first time that the composition of fission gas has been monitored all along an irradiation experiment in a MTR, giving access to the gas release kinetics. New researches involve thick film transducers produced by screen-printing process in order to propose piezoelectric structures for harsh temperature and irradiation measurements. The second project consists in the development of a non-destructive device that can be directly applied on a LWR fuel rod. The problem to be solved relates to the measurement of the fission gas pressure and composition in a fuel rod using a non-destructive method. Fuel rod internal pressure is one of the safety criteria applied in nuclear power analyses. This criterion must be verified in order to avoid any fuel-cladding gap reopening risk and therefore any local clad ballooning. Apart from the safety implications, this parameter is also a fuel behaviour indicator and reflects the overall fuel performance in operation, but also during shipping and long-term storage. Rod internal pressure is one criterion amongst others, like cladding corrosion, against which the acceptable fuel burn-up limit is set. A sensor has been achieved in 2007. A full-scale hot cell test of the internal gas pressure and composition measurement by an acoustic sensor was conducted successfully between 2008 and 2010 on 5 high burn-up MOX fuel rods and 2 very high burn-up UO{sub 2} fuel rods in LECA Facility at Cadarache Centre. An improvement of this sensor has been proposed, allowing us to divide by two the uncertainty on the pressure measurement. In the case of hot-cell measurements, viscous liquid can be used to couple the sensor with the rod. For gas content with a pressure exceeding 15 bars and a 10% Xe/Kr ratio, such coupling may reduce relative acoustic method accuracy by ±7% for pressure measurement result and ±0.25 % for the assessment of gas composition. These results make it possible to demonstrate the feasibility of the technique on LWR fuel rods. The transducer and the associated methodology are now operational for non-destructive measurements in hot lab facilities and allow characterising the fission gas without puncturing the fuel rods. Up to now, any other non-destructive method can be proposed. A next step will be the development of an industrial application in a fuel storage pool in order to perform a large number of measurements on a fuel assembly in nuclear plants.« less

Uncertainty in the delayed neutron fraction in fuel assembly depletion calculations

NASA Astrophysics Data System (ADS)

Aures, Alexander; Bostelmann, Friederike; Kodeli, Ivan A.; Velkov, Kiril; Zwermann, Winfried

2017-09-01

This study presents uncertainty and sensitivity analyses of the delayed neutron fraction of light water reactor and sodium-cooled fast reactor fuel assemblies. For these analyses, the sampling-based XSUSA methodology is used to propagate cross section uncertainties in neutron transport and depletion calculations. Cross section data is varied according to the SCALE 6.1 covariance library. Since this library includes nu-bar uncertainties only for the total values, it has been supplemented by delayed nu-bar uncertainties from the covariance data of the JENDL-4.0 nuclear data library. The neutron transport and depletion calculations are performed with the TRITON/NEWT sequence of the SCALE 6.1 package. The evolution of the delayed neutron fraction uncertainty over burn-up is analysed without and with the consideration of delayed nu-bar uncertainties. Moreover, the main contributors to the result uncertainty are determined. In all cases, the delayed nu-bar uncertainties increase the delayed neutron fraction uncertainty. Depending on the fuel composition, the delayed nu-bar values of uranium and plutonium in fact give the main contributions to the delayed neutron fraction uncertainty for the LWR fuel assemblies. For the SFR case, the uncertainty of the scattering cross section of U-238 is the main contributor.

Neutronics Investigations for the Lower Part of a Westinghouse SVEA-96+ Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, M.F.; Luethi, A.; Seiler, R.

2002-05-15

Accurate critical experiments have been performed for the validation of total fission (F{sub tot}) and {sup 238}U-capture (C{sub 8}) reaction rate distributions obtained with CASMO-4, HELIOS, BOXER, and MCNP4B for the lower axial region of a real Westinghouse SVEA-96+ fuel assembly. The assembly comprised fresh fuel with an average {sup 235}U enrichment of 4.02 wt%, a maximum enrichment of 4.74 wt%, 14 burnable-absorber fuel pins, and full-density water moderation. The experimental configuration investigated was core 1A of the LWR-PROTEUS Phase I project, where 61 different fuel pins, representing {approx}64% of the assembly, were gamma-scanned individually. Calculated (C) and measured (E)more » values have been compared in terms of C/E distributions. For F{sub tot}, the standard deviations are 1.2% for HELIOS, 0.9% for CASMO-4, 0.8% for MCNP4B, and 1.7% for BOXER. Standard deviations of 1.1% for HELIOS, CASMO-4, and MCNP4B and 1.2% for BOXER were obtained in the case of C{sub 8}. Despite the high degree of accuracy observed on the average, it was found that the five burnable-absorber fuel pins investigated showed a noticeable underprediction of F{sub tot}, quite systematically, for the deterministic codes evaluated (average C/E for the burnable-absorber fuel pins in the range 0.974 to 0.988, depending on the code)« less

Corrigendum to "Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions" [J. Nucl. Mater. 448 (2014) 520-533

NASA Astrophysics Data System (ADS)

Ott, L. J.; Robb, K. R.; Wang, D.

2015-06-01

In Section 5.2, certain material properties for "FeCrAl oxide" were not modeled based on "stainless steel oxide" as indicated in the text. Instead, the "FeCrAl oxide" material properties were modeled using the default properties in MELCOR for "zirconium oxide". The properties affected are the FeCrAl oxide density, specific heat, enthalpy, thermal conductivity, melting point, and latent heat of fusion. Table 5.1 and Figs. 5.1a-d from Section 5.2 have been corrected below. As discussed below, the overall conclusions of the paper remain unchanged.

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sridharan, Kumar; Mariani, Robert; Bai, Xianming

Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO 2 layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins inmore » accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO 4) and silica (SiO 2) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi 2 coating) during clad quenching experiments is discussed in detail.« less

Graphite Waste Tank Cleanup and Decontamination under the Marcoule UP1 D and D Program - 13166

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomasset, Philippe; Chabeuf, Jean-Michel; Thiebaut, Valerie

2013-07-01

The UP1 plant in Marcoule reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. During more than 40 years, the decladding operations produced thousands of tons of processed waste, mainly magnesium and graphite fragments. In the absence of a French repository for the graphite waste, the graphite sludge content of the storage pits had to be retrieved and transferred into a newer and safer pit. After an extensive R and D program, the equipment and process necessary for retrieval operations were designed, built and tested. Themore » innovative process is mainly based on the use of two pumps (one to capture and the other one to transfer the sludge) working one after the other and a robotic arm mounted on a telescopic mast. A dedicated process was also set up for the removal of the biggest fragments. The retrieval of the most irradiating fragments was a challenge. Today, the first pit is totally empty and its stainless steel walls have been decontaminated using gels. In the second pit, the sludge retrieval and transfer operations have been almost completed. Most of the non-pumpable graphite fragments has been removed and transferred to a new storage pit. After more than 6 years of operations in sludge retrieval, a lot of experience was acquired from which important 'lessons learned' could be shared. (authors)« less

Data summary report for fission product release test VI-5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, M.F.; Lorenz, R.A.; Travis, J.R.

Test VI-5, the fifth in a series of high-temperature fission product release tests in a vertical test apparatus, was conducted in a flowing mixture of hydrogen and helium. The test specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium which had been irradiated to a burnup of {approximately}42 MWd/kg. Using a hot cell-mounted test apparatus, the fuel rod was heated in an induction furnace under simulated LWR accident conditions to two test temperatures, 2000 K for 20 min and then 2700 K for an additional 20 min. The released fission products were collected inmore » three sequentially operated collection trains on components designed to measure fission product transport characteristics and facilitate sampling and analysis. The results from this test were compared with those obtained in previous tests in this series and with the CORSOR-M and ORNL diffusion release models for fission product release. 21 refs., 19 figs., 12 tabs.« less

Steam Oxidation of FeCrAl and SiC in the Severe Accident Test Station (SATS)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pint, Bruce A.; Unocic, Kinga A.; Terrani, Kurt A.

2015-08-01

Numerous research projects are directed towards developing accident tolerant fuel (ATF) concepts that will enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the U.S. program, the high temperature steam oxidation performance of ATF solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012 [1-3] and this facility continues to support those efforts in the ATF community. Compared to the current UO2/Zr-based alloy fuel system, alternative cladding materials can offer slower oxidation kinetics and a smaller enthalpy of oxidation that can significantly reduce the rate of heatmore » and hydrogen generation in the core during a coolant-limited severe accident [4-5]. Thus, steam oxidation behavior is a key aspect of the evaluation of ATF concepts. This report summarizes recent work to measure steam oxidation kinetics of FeCrAl and SiC specimens in the SATS.« less

Isotopic signature of atmospheric xenon released from light water reactors.

PubMed

Kalinowski, Martin B; Pistner, Christoph

2006-01-01

A global monitoring system for atmospheric xenon radioactivity is being established as part of the International Monitoring System to verify compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The isotopic activity ratios of (135)Xe, (133m)Xe, (133)Xe and (131m)Xe are of interest for distinguishing nuclear explosion sources from civilian releases. Simulations of light water reactor (LWR) fuel burn-up through three operational reactor power cycles are conducted to explore the possible xenon isotopic signature of nuclear reactor releases under different operational conditions. It is studied how ratio changes are related to various parameters including the neutron flux, uranium enrichment and fuel burn-up. Further, the impact of diffusion and mixing on the isotopic activity ratio variability are explored. The simulations are validated with reported reactor emissions. In addition, activity ratios are calculated for xenon isotopes released from nuclear explosions and these are compared to the reactor ratios in order to determine whether the discrimination of explosion releases from reactor effluents is possible based on isotopic activity ratios.

Improvements and applications of COBRA-TF for stand-alone and coupled LWR safety analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Avramova, M.; Cuervo, D.; Ivanov, K.

2006-07-01

The advanced thermal-hydraulic subchannel code COBRA-TF has been recently improved and applied for stand-alone and coupled LWR core calculations at the Pennsylvania State Univ. in cooperation with AREVA NP GmbH (Germany)) and the Technical Univ. of Madrid. To enable COBRA-TF for academic and industrial applications including safety margins evaluations and LWR core design analyses, the code programming, numerics, and basic models were revised and substantially improved. The code has undergone through an extensive validation, verification, and qualification program. (authors)

Flowsheet Analysis of U-Pu Co-Crystallization Process as a New Reprocessing System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shunji Homma; Jun-ichi Ishii; Jiro Koga

2006-07-01

A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined bymore » flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced. (authors)« less

Synthesis and sintering of UN-UO 2 fuel composites

DOE PAGES

Jaques, Brian J.; Watkins, Jennifer; Croteau, Joseph R.; ...

2015-06-17

In this study, the design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO 2, which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO 2 to UN has been suggested. In order to avoid carbon impurities, UN was synthesized frommore » elemental uranium using a hydride-dehydride-nitride thermal synthesis route prior to mixing with up to 10 wt% UO 2 in a planetary ball mill. UN and UN – UO 2 composite pellets were sintered in Ar – (0–1 at%) N 2 to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO 2 composite pellets were also sintered in Ar – 100 ppm N 2 to assess the effects of temperature (1700–2000 °C) on the final grain morphology and phase concentration.« less

Effect of the oxidation front penetration on in-clad hydrogen migration

NASA Astrophysics Data System (ADS)

Feria, F.; Herranz, L. E.

2018-03-01

In LWR fuel claddings the embrittlement due to hydrogen precipitates (i.e., hydrides) is a degrading mechanism that concerns in nuclear safety, particularly in dry storage. A relevant factor is the radial distribution of the hydrogen absorbed, especially the hydride rim formed. Thus, a reliable assessment of fuel performance should account for hydrogen migration. Based on the current state of modelling of hydrogen dynamics in the cladding, a 1D radial model has been derived and coupled with the FRAPCON code. The model includes the effect of the oxidation front progression on in-clad hydrogen migration, based on experimental observations found (i.e., dissolution/diffusion/re-precipitation of the hydrogen in the matrix ahead of the oxidation front). A remarkable quantitative impact of this new contribution has been shown by analyzing the hydrogen profile across the cladding of several high burnup fuel scenarios (>60 GW d/tU); other potential contributions like thermodiffusion and diffusion in the hydride phase hardly make any difference. Comparisons against PIE measurements allow concluding that the model accuracy notably increases when the effect of the oxidation front is accounted for in the hydride rim formation. In spite of the promising results, further validation would be needed.

FASTGRASS implementation in BISON and Fission gas behavior characterization in UO 2 and connection to validating MARMOT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yun, Di; Mo, Kun; Ye, Bei

2015-09-30

This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL). Two major accomplishments in FY 15 are summarized in this report: (1) implementation of the FASTGRASS module in the BISON code; and (2) a Xe implantation experiment for large-grained UO 2. Both BISON AND MARMOT codes have been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. To contribute to the development of the Moose-Bison-Marmot (MBM) code suite, we have implemented the FASTGRASS fission gas model as a module inmore » the BISON code. Based on rate theory formulations, the coupled FASTGRASS module in BISON is capable of modeling LWR oxide fuel fission gas behavior and fission gas release. In addition, we conducted a Xe implantation experiment at the Argonne Tandem Linac Accelerator System (ATLAS) in order to produce the needed UO 2 samples with desired bubble morphology. With these samples, further experiments to study the fission gas diffusivity are planned to provide validation data for the Fission Gas Release Model in MARMOT codes.« less

Influence of post exposure bake time on EUV photoresist RLS trade-off

NASA Astrophysics Data System (ADS)

Vesters, Yannick; De Simone, Danilo; De Gendt, Stefan

2017-03-01

To achieve high volume manufacturing, EUV photoresists need to push back the "RLS trade-off" by simultaneously improving Resolution, Line-Width Roughness and Sensitivity (exposure dose). Acid diffusion in chemically amplified resist is known to impact these performances. This work studies the diffusion of acid in chemically amplified resist by varying the post exposure bake duration while monitoring the evolution of CD and LWR for 6 chemically amplified EUV photoresists (CAR). We observed a first regime where both CD and LWR quickly decrease during the first 30s of post exposure bake (PEB). This can be related to the deprotection reaction taking place in the exposed part of the resist. After 60s the decrease in CD and LWR slows down significantly, likely related to a regime of acid diffusion from exposed to unexposed region, and acid-quencher neutralization at the interface of these two regions. We tested two resists with different protecting group and the one having lower activation energy shows a faster CD change in the second regime, resulting in a worsening of LWR for longer PEB time. On the contrary, a resist with a high quencher loading shows reduced net diffusion of acid towards the unexposed region and controls the resist edge profile. In other words longer PEB does not degrade LWR, but as it reduces the line CD, sensitivity is impacted. With an appropriate ratio selection of quencher to PAG, an EUV dose reduction of up to 12% can be achieved with a change from a standard 60 second to a 240 second PEB time, while keeping LWR and resolution constant and therefore pushing the RLS performances. Finally, we confirmed that the observations on positive tone development (PTD) resist could be applied to negative tone development (NTD) resist: with a high quencher NTD resist we observed a dose reduction of 8% for longer PEB time, keeping LWR and resolution constant.

Interdiffusion behavior of U3Si2 with FeCrAl via diffusion couple studies

NASA Astrophysics Data System (ADS)

Hoggan, Rita E.; He, Lingfeng; Harp, Jason M.

2018-04-01

Uranium silicide (U3Si2) is a candidate to replace uranium oxide (UO2) as light water reactor (LWR) fuel because of its higher thermal conductivity and higher fissile density relative to the current standard, UO2. A class of Fe, Cr, Al alloys collectively known as FeCrAl alloys that have superior mechanical and oxidation resistance are being considered as an alternative to the standard Zirconium based LWR cladding. The interdiffusion behavior between FeCrAl and U3Si2 is investigated in this study. Commercially available FeCrAl, along with U3Si2 pellets were placed in diffusion couples. Individual tests were ran at temperatures ranging from 500 °C to 1000 °C for 30 h and 100 h. The interdiffusion was analyzed with an optical microscope, scanning electron microscope, and transmission electron microscope. Uniform and planar interdiffusion layers along the material interface were illustrated with backscatter electron micrographs and energy-dispersive X-ray spectroscopy. Electron diffraction was used to validate phases present in the system, including distinct U2Fe3Si/UFe2 and UFeSi layers at the material interface. U and Fe diffused far into the FeCrAl and U3Si2 matrix, respectively, in the higher temperature tests. No interaction was observed at 500 °C for 30 h.

CTF Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Avramova, Maria N.; Salko, Robert K.

Coolant-Boiling in Rod Arrays|Two Fluids (COBRA-TF) is a thermal/ hydraulic (T/H) simulation code designed for light water reactor (LWR) vessel analysis. It uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and 3D Cartesian forms of 9 conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and had been used and modified by several institutions over the last few decades. COBRA-TF also found use at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Management Group (RDFMG) and has been improved, updated, andmore » subsequently re-branded as CTF. As part of the improvement process, it was necessary to generate sufficient documentation for the open-source code which had lacked such material upon being adopted by RDFMG. This document serves mainly as a theory manual for CTF, detailing the many two-phase heat transfer, drag, and important accident scenario models contained in the code as well as the numerical solution process utilized. Coding of the models is also discussed, all with consideration for updates that have been made when transitioning from COBRA-TF to CTF. Further documentation outside of this manual is also available at RDFMG which focus on code input deck generation and source code global variable and module listings.« less

EUV lithography for 22nm half pitch and beyond: exploring resolution, LWR, and sensitivity tradeoffs

NASA Astrophysics Data System (ADS)

Putna, E. Steve; Younkin, Todd R.; Leeson, Michael; Caudillo, Roman; Bacuita, Terence; Shah, Uday; Chandhok, Manish

2011-04-01

The International Technology Roadmap for Semiconductors (ITRS) denotes Extreme Ultraviolet (EUV) lithography as a leading technology option for realizing the 22nm half pitch node and beyond. According to recent assessments made at the 2010 EUVL Symposium, the readiness of EUV materials remains one of the top risk items for EUV adoption. The main development issue regarding EUV resists has been how to simultaneously achieve high resolution, high sensitivity, and low line width roughness (LWR). This paper describes our strategy, the current status of EUV materials, and the integrated post-development LWR reduction efforts made at Intel Corporation. Data collected utilizing Intel's Micro- Exposure Tool (MET) is presented in order to examine the feasibility of establishing a resist process that simultaneously exhibits <=22nm half-pitch (HP) L/S resolution at <=11.3mJ/cm2 with <=3nm LWR.

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Loflin, Leonard; McRimmon, Beth

2014-12-18

This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

The impact of integrated water management on the Space Station propulsion system

NASA Technical Reports Server (NTRS)

Schmidt, George R.

1987-01-01

The water usage of elements in the Space Station integrated water system (IWS) is discussed, and the parameters affecting the overall water balance and the water-electrolysis propulsion-system requirements are considered. With nominal IWS operating characteristics, extra logistic water resupply (LWR) is found to be unnecessary in the satisfaction of the nominal propulsion requirements. With the consideration of all possible operating characteristics, LWR will not be required in 65.5 percent of the cases, and for 17.9 percent of the cases LWR can be eliminated by controlling the stay time of theShuttle Orbiter orbiter.

UFD Storage and Transportation - Transportation Working Group Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maheras, Steven J.; Ross, Steven B.

2011-08-01

The Used Fuel Disposition (UFD) Transportation Task commenced in October 2010. As its first task, Pacific Northwest National Laboratory (PNNL) compiled a list of structures, systems, and components (SSCs) of transportation systems and their possible degradation mechanisms during extended storage. The list of SSCs and the associated degradation mechanisms [known as features, events, and processes (FEPs)] were based on the list of used nuclear fuel (UNF) storage system SSCs and degradation mechanisms developed by the UFD Storage Task (Hanson et al. 2011). Other sources of information surveyed to develop the list of SSCs and their degradation mechanisms included references suchmore » as Evaluation of the Technical Basis for Extended Dry Storage and Transportation of Used Nuclear Fuel (NWTRB 2010), Transportation, Aging and Disposal Canister System Performance Specification, Revision 1 (OCRWM 2008), Data Needs for Long-Term Storage of LWR Fuel (EPRI 1998), Technical Bases for Extended Dry Storage of Spent Nuclear Fuel (EPRI 2002), Used Fuel and High-Level Radioactive Waste Extended Storage Collaboration Program (EPRI 2010a), Industry Spent Fuel Storage Handbook (EPRI 2010b), and Transportation of Commercial Spent Nuclear Fuel, Issues Resolution (EPRI 2010c). SSCs include items such as the fuel, cladding, fuel baskets, neutron poisons, metal canisters, etc. Potential degradation mechanisms (FEPs) included mechanical, thermal, radiation and chemical stressors, such as fuel fragmentation, embrittlement of cladding by hydrogen, oxidation of cladding, metal fatigue, corrosion, etc. These degradation mechanisms are discussed in Section 2 of this report. The degradation mechanisms have been evaluated to determine if they would be influenced by extended storage or high burnup, the need for additional data, and their importance to transportation. These categories were used to identify the most significant transportation degradation mechanisms. As expected, for the most part, the transportation importance was mirrored by the importance assigned by the UFD Storage Task. A few of the more significant differences are described in Section 3 of this report« less

Heterogeneous sodium fast reactor designed for transmuting minor actinide waste isotopes into plutonium fuel

NASA Astrophysics Data System (ADS)

Bays, Samuel Eugene

2008-10-01

In the past several years there has been a renewed interest in sodium fast reactor (SFR) technology for the purpose of destroying transuranic waste (TRU) produced by light water reactors (LWR). The utility of SFRs as waste burners is due to the fact that higher neutron energies allow all of the actinides, including the minor actinides (MA), to contribute to fission. It is well understood that many of the design issues of LWR spent nuclear fuel (SNF) disposal in a geologic repository are linked to MAs. Because the probability of fission for essentially all the "non-fissile" MAs is nearly zero at low neutron energies, these isotopes act as a neutron capture sink in most thermal reactor systems. Furthermore, because most of the isotopes produced by these capture reactions are also non-fissile, they too are neutron sinks in most thermal reactor systems. Conversely, with high neutron energies, the MAs can produce neutrons by fast fission. Additionally, capture reactions transmute the MAs into mostly plutonium isotopes, which can fission more readily at any energy. The transmutation of non-fissile into fissile atoms is the premise of the plutonium breeder reactor. In a breeder reactor, not only does the non-fissile "fertile" U-238 atom contribute fast fission neutrons, but also transmutes into fissile Pu-239. The fissile value of the plutonium produced by MA transmutation can only be realized in fast neutron spectra. This is due to the fact that the predominate isotope produced by MA transmutation, Pu-238, is itself not fissile. However, the Pu-238 fission cross section is significantly larger than the original transmutation parent, predominately: Np-237 and Am-241, in the fast energy range. Also, Pu-238's fission cross section and fission-to-capture ratio is almost as high as that of fissile Pu-239 in the fast neutron spectrum. It is also important to note that a neutron absorption in Pu-238, that does not cause fission, will instead produce fissile Pu-239. Given this fast fissile quality and also the fact that Pu-238 is transmuted from Np-237 and Am-241, these MAs are regarded as fertile material in the SFR design proposed by this dissertation. This dissertation demonstrates a SFR design which is dedicated to plutonium breeding by targeting Am-241 transmutation. This SFR design uses a moderated axial transmutation target that functions primarily as a pseudo-blanket fuel, which is reprocessed with the active driver fuel in an integrated recycling strategy. This work demonstrates the cost and feasibility advantages of plutonium breeding via MA transmutation by adopting reactor, reprocessing and fuel technologies previously demonstrated for traditional breeder reactors. The fuel cycle proposed seeks to find a harmony between the waste management advantages of transuranic burning SFRs and the resource sustainability of traditional plutonium breeder SFRs. As a result, the enhanced plutonium conversion from MAs decreases the burner SFR's fuel costs, by extracting more fissile value from the initial TRU purchased through SNF reprocessing.

Development/Modernization of an Advanced Non-Light Water Reactor Probabilistic Risk Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henneke, Dennis W.; Robinson, James

In 2015, GE Hitachi Nuclear Energy (GEH) teamed with Argonne National Laboratory (Argonne) to perform Research and Development (R&D) of next-generation Probabilistic Risk Assessment (PRA) methodologies for the modernization of an advanced non-Light Water Reactor (non-LWR) PRA. This effort built upon a PRA developed in the early 1990s for GEH’s Power Reactor Inherently Safe Module (PRISM) Sodium Fast Reactor (SFR). The work had four main tasks: internal events development modeling the risk from the reactor for hazards occurring at-power internal to the plant; an all hazards scoping review to analyze the risk at a high level from external hazards suchmore » as earthquakes and high winds; an all modes scoping review to understand the risk at a high level from operating modes other than at-power; and risk insights to integrate the results from each of the three phases above. To achieve these objectives, GEH and Argonne used and adapted proven PRA methodologies and techniques to build a modern non-LWR all hazards/all modes PRA. The teams also advanced non-LWR PRA methodologies, which is an important outcome from this work. This report summarizes the project outcomes in two major phases. The first phase presents the methodologies developed for non-LWR PRAs. The methodologies are grouped by scope, from Internal Events At-Power (IEAP) to hazards analysis to modes analysis. The second phase presents details of the PRISM PRA model which was developed as a validation of the non-LWR methodologies. The PRISM PRA was performed in detail for IEAP, and at a broader level for hazards and modes. In addition to contributing methodologies, this project developed risk insights applicable to non-LWR PRA, including focus-areas for future R&D, and conclusions about the PRISM design.« less

Surface Shortwave and Longe Wave Solar Radiation Atmospheric Aerosols Radiative Forcing Using Sunphotometer , Modis Satellite and Cnr -1 Measurements Over Western Indian Tropical Site or Udaipur ( 24.57N, 73. 69E, 588M Asl)

NASA Astrophysics Data System (ADS)

Vyas, B. M.

2017-12-01

The analysis of investigation describes the experimental results of monthly surafcae short wave radiative(SWR) and longwave radaitive(LWR) atmospheric aerosols radaitive forcing derived from daily mesaured values of AOD at 550 nm from MODIS Terra and Acqau satellite as well as hourly measurement of AOD at 500nm from MICROTOPS _II sunsphotometer ( M/S Solar Light Co. USA) with round the clock of 24 hourly measurement of CNR-1 ( M/s KIP & ZONN, Netherland) during the clear sky days over Udaipur. For the present investigation, such above simulatneous daily data sets of period from Oct.,2011 to June 2017 were used to study the monthly and sesaonal ground level SWR and LWR over a semi- urban and semi-arid western Indian tropical site for pre- monsoon, post-monsoon and winter months. In this study, a well known method of computing surface SWR and LWR has been employed as Method -1 as suggested by Shrivastava et al., 2011. A stong and distinct different sesaonal surface SWR and LWR due to atmospheric aerosols has observed that the well defined seasonal neagtive SWR is observed maximum in pre- monsoon and minimum in winter and post-monsoon months. But in contary to the above, higher positive monthly LWR values are noticed in pre-monsoon as compared to in winter months. The The inter- annual sesaonal trend of the SWR and LWR are also noticed in the present work. The reslts of present study will be compared with other availlable simillar study using SBDART at other other Indian stations.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Besmann, Theodore M; Shin, Dongwon

TRISO coated particle fuel is envisioned as a next generation replacement for current urania pellet fuel in LWR applications. To obtain adequate fissile loading the kernel of the TRISO particle will need to be UN. In support of the fuel development effort, an assessment of phase regions of interest in the U-C-N system was undertaken as the fuel will be prepared by the carbothermic reduction of the oxide and it will be in equilibrium with carbon within the TRISO particle. The phase equilibria and thermochemistry of the U-C-N system is reviewed, including nitrogen pressure measurements above various phase fields. Selectedmore » measurements were used to fit a first order model of the UC1-xNx phase, represented by the inter-solution of UN and UC. Fit to the data was significantly improved by also adjusting the heat of formation for UN by ~12 kJ/mol and the phase equilbria was best reproduced by also adjusting the heat for U2N3 by +XXX. The determined interaction parameters yielded a slightly positive deviation from ideality, which agrees with lattice parameter measurements which show positive deviation from Vegard s law. The resultant model together with reported values for other phases in the system were used to generate isothermal sections of the U-C-N phase diagram. Nitrogen partial pressures were also computed for regions of interest.« less

Void reactivity feedback analysis for U-based and Th-based LWR incineration cycles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindley, B.A.; Parks, G.T.; Franceschini, F.

2013-07-01

In reduced-moderation LWRs, an external supply of transuranic (TRU) can be incinerated by mixing it with a fertile isotope ({sup 238}U or {sup 232}Th) and recycling all the actinides after each cycle. Performance is limited by coolant reactivity feedback - the moderator density coefficient (MDC) must be kept negative. The MDC is worse when more TRU is loaded, but TRU feed is also needed to maintain criticality. To assess the performance of this fuel cycle in different neutron spectra, three LWRs are considered: 'reference' PWRs and reduced-moderation PWRs and BWRs. The MDC of the equilibrium cycle is analysed by reactivitymore » decomposition with perturbed coolant density by isotope and neutron energy. The results show that using {sup 232}Th as a fertile isotope yields superior performance to {sup 238}U. This is due essentially to the high resonance η of U bred from Th (U3), which increases the fissility of the U3-TRU isotope vector in the Th-fueled system relative to the U-fueled system, and also improves the MDC in a sufficiently hard spectrum. Spatial separation of TRU and U3 in the Th-fueled system renders further improvement by hardening the neutron spectrum in the TRU and softening it in the U3. This improves the TRU η and increases the negative MDC contribution from reduced thermal fission in U3. (authors)« less

Uranium chloride extraction of transuranium elements from LWR fuel

DOEpatents

Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

1992-08-25

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure.

Uranium chloride extraction of transuranium elements from LWR fuel

DOEpatents

Miller, William E.; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Pierce, R. Dean

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

Overview of the multifaceted activities towards development and deployment of nuclear-grade FeCrAl Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G; Yamamoto, Yukinori; Pint, Bruce A

2016-01-01

A large effort is underway under the leadership of US DOE Fuel Cycle R&D program to develop advanced FeCrAl alloys as accident tolerant fuel (ATF) cladding to replace Zr-based alloys in light water reactors. The primary motivation is the excellent oxidation resistance of these alloys in high-temperature steam environments right up to their melting point (roughly three orders of magnitude slower oxidation kinetics than zirconium). A multifaceted effort is ongoing to rapidly advance FeCrAl alloys as a mature ATF concept. The activities span the broad spectrum of alloy development, environmental testing (high-temperature high-pressure water and elevated temperature steam), detailed mechanicalmore » characterization, material property database development, neutron irradiation, thin tube production, and multiple integral fuel test campaigns. Instead of off-the-shelf commercial alloys that might not prove optimal for the LWR fuel cladding application, a large amount of effort has been placed on the alloy development to identify the most optimum composition and microstructure for this application. The development program is targeting a cladding that offers performance comparable to or better than modern Zr-based alloys under normal operating and off-normal conditions. This paper provides a comprehensive overview of the systematic effort to advance nuclear-grade FeCrAl alloys as an ATF cladding in commercial LWRs.« less

Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

NASA Astrophysics Data System (ADS)

Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

2017-07-01

Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (D

Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, A.L.; Wilson, J.H.; Arwood, P.C.

The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) andmore » ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models.« less

78 FR 55118 - Seismic Instrumentation for Nuclear Power Plants

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-09

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0202] Seismic Instrumentation for Nuclear Power Plants... Reports for Nuclear Power Plants: LWR Edition,'' Section 3.7.4, ``Seismic Instrumentation.'' DATES: Submit... Nuclear Power Plants: LWR Edition'' (SRP, from the current Revision 2 to a new Revision 3). The proposed...

Comparative testing of nondestructive examination techniques for concrete structures

NASA Astrophysics Data System (ADS)

Clayton, Dwight A.; Smith, Cyrus M.

2014-03-01

A multitude of concrete-based structures are typically part of a light water reactor (LWR) plant to provide foundation, support, shielding, and containment functions. Concrete has been used in the construction of nuclear power plants (NPPs) because of three primary properties, its inexpensiveness, its structural strength, and its ability to shield radiation. Examples of concrete structures important to the safety of LWR plants include containment building, spent fuel pool, and cooling towers. Comparative testing of the various NDE concrete measurement techniques requires concrete samples with known material properties, voids, internal microstructure flaws, and reinforcement locations. These samples can be artificially created under laboratory conditions where the various properties can be controlled. Other than NPPs, there are not many applications where critical concrete structures are as thick and reinforced. Therefore, there are not many industries other than the nuclear power plant or power plant industry that are interested in performing NDE on thick and reinforced concrete structures. This leads to the lack of readily available samples of thick and heavily reinforced concrete for performing NDE evaluations, research, and training. The industry that typically performs the most NDE on concrete structures is the bridge and roadway industry. While bridge and roadway structures are thinner and less reinforced, they have a good base of NDE research to support their field NDE programs to detect, identify, and repair concrete failures. This paper will summarize the initial comparative testing of two concrete samples with an emphasis on how these techniques could perform on NPP concrete structures.

Radiation-induced grain subdivision and bubble formation in U3Si2 at LWR temperature

NASA Astrophysics Data System (ADS)

Yao, Tiankai; Gong, Bowen; He, Lingfeng; Harp, Jason; Tonks, Michael; Lian, Jie

2018-01-01

U3Si2, an advanced fuel form proposed for light water reactors (LWRs), has excellent thermal conductivity and a high fissile element density. However, limited understanding of the radiation performance and fission gas behavior of U3Si2 is available at LWR conditions. This study explores the irradiation behavior of U3Si2 by 300 keV Xe+ ion beam bombardment combining with in-situ transmission electron microscopy (TEM) observation. The crystal structure of U3Si2 is stable against radiation-induced amorphization at 350 °C even up to a very high dose of 64 displacements per atom (dpa). Grain subdivision of U3Si2 occurs at a relatively low dose of 0.8 dpa and continues to above 48 dpa, leading to the formation of high-density nanoparticles. Nano-sized Xe gas bubbles prevail at a dose of 24 dpa, and Xe bubble coalescence was identified with the increase of irradiation dose. The volumetric swelling resulting from Xe gas bubble formation and coalescence was estimated with respect to radiation dose, and a 2.2% volumetric swelling was observed for U3Si2 irradiated at 64 dpa. Due to extremely high susceptibility to oxidation, the nano-sized U3Si2 grains upon radiation-induced grain subdivision were oxidized to nanocrystalline UO2 in a high vacuum chamber for TEM observation, eventually leading to the formation of UO2 nanocrystallites stable up to 80 dpa.

Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data

NASA Astrophysics Data System (ADS)

Rochman, Dimitri A.; Vasiliev, Alexander; Dokhane, Abdelhamid; Ferroukhi, Hakim

2018-05-01

This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal.

Analysis of key safety metrics of thorium utilization in LWRs

DOE PAGES

Ade, Brian J.; Bowman, Stephen M.; Worrall, Andrew; ...

2016-04-08

Here, thorium has great potential to stretch nuclear fuel reserves because of its natural abundance and because it is possible to breed the 232Th isotope into a fissile fuel ( 233U). Various scenarios exist for utilization of thorium in the nuclear fuel cycle, including use in different nuclear reactor types (e.g., light water, high-temperature gas-cooled, fast spectrum sodium, and molten salt reactors), along with use in advanced accelerator-driven systems and even in fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based onmore » concepts that mix thorium with uranium (UO 2 + ThO 2) or that add fertile thorium (ThO 2) fuel pins to typical LWR fuel assemblies. Utilization of mixed fuel assemblies (PuO 2 + ThO 2) is also possible. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts to the nuclear fuel. Thorium and its irradiation products have different nuclear characteristics from those of uranium and its irradiation products. ThO 2, alone or mixed with UO 2 fuel, leads to different chemical and physical properties of the fuel. These key reactor safety–related issues have been studied at Oak Ridge National Laboratory and documented in “Safety and Regulatory Issues of the Thorium Fuel Cycle” (NUREG/CR-7176, U.S. Nuclear Regulatory Commission, 2014). Various reactor analyses were performed using the SCALE code system for comparison of key performance parameters of both ThO 2 + UO 2 and ThO 2 + PuO 2 against those of UO 2 and typical UO 2 + PuO 2 mixed oxide fuels, including reactivity coefficients and power sharing between surrounding UO 2 assemblies and the assembly of interest. The decay heat and radiological source terms for spent fuel after its discharge from the reactor are also presented. Based on this evaluation, potential impacts on safety requirements and identification of knowledge gaps that require additional analysis or research to develop a technical basis for the licensing of thorium fuel are identified.« less

Microstructure investigations of U3Si2 implanted by high-energy Xe ions at 600 °C

NASA Astrophysics Data System (ADS)

Miao, Yinbin; Harp, Jason; Mo, Kun; Kim, Yeon Soo; Zhu, Shaofei; Yacout, Abdellatif M.

2018-05-01

The microstructure investigations on a high-energy Xe-implanted U3Si2 pellet were performed. The promising accident tolerant fuel (ATF) candidate, U3Si2, was irradiated by 84 MeV Xe ions at 600 °C at Argonne Tandem Linac Accelerator System (ATLAS). The characterizations of the Xe implanted sample were conducted using advanced transmission electron microscopy (TEM) techniques. An oxidation layer was observed on the sample surface after irradiation under the ∼10-5 Pa vacuum. The study on the oxidation layer not only unveils the readily oxidation behavior of U3Si2 under high-temperature irradiation conditions, but also develops an understanding of its oxidation mechanism. Intragranular Xe bubbles with bimodal size distribution were observed within the Xe deposition region of the sample induced by 84 MeV Xe ion implantation. At the irradiation temperature of 600 °C, the gaseous swelling strain contributed by intragranular bubbles was found to be insignificant, indicating an acceptable fission gas behavior of U3Si2 as a light water reactor (LWR) fuel operating at such a temperature.

On the effect of irradiation-induced resolution in modelling fission gas release in UO2 LWR fuel

NASA Astrophysics Data System (ADS)

Lösönen, Pekka

2017-12-01

Irradiation resolution of gas atoms and vacancies from intra- and intergranular bubbles in sintered UO2 fuel was studied by comparing macroscopic models with a more mechanistic approach. The applied macroscopic models imply the resolution rate of gas atoms to be proportional to gas concentration in intragranular bubbles and at grain boundary (including intergranular bubbles). A relation was established between the macroscopic models and a single encounter of an energetic fission fragment with a bubble. The effect of bubble size on resolution was quantified. The number of resoluted gas atoms per encounter of a fission fragment per bubble was of the same order of magnitude for intra- and intergranular bubbles. However, the resulting macroscopic resolution rate of gas atoms was about two orders of magnitude larger from intragranular bubbles. The number of vacancies resoluted from a grain face bubble by a passing fission fragment was calculated. The obtained correlations for resolution of gas atoms from intragranular bubbles and grain boundaries and for resolution of vacancies from grain face bubbles were used to demonstrate the effect of irradiation resolution on fission gas release.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jaques, Brian J.; Watkins, Jennifer; Croteau, Joseph R.

In this study, the design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO 2, which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO 2 to UN has been suggested. In order to avoid carbon impurities, UN was synthesized frommore » elemental uranium using a hydride-dehydride-nitride thermal synthesis route prior to mixing with up to 10 wt% UO 2 in a planetary ball mill. UN and UN – UO 2 composite pellets were sintered in Ar – (0–1 at%) N 2 to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO 2 composite pellets were also sintered in Ar – 100 ppm N 2 to assess the effects of temperature (1700–2000 °C) on the final grain morphology and phase concentration.« less

Short communication on " In-situ TEM ion irradiation investigations on U 3Si 2 at LWR temperatures"

DOE PAGES

Miao, Yinbin; Harp, Jason; Mo, Kun; ...

2016-11-21

Here, the radiation-induced amorphization of U 3Si 2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 10 15 ions/cm 2 to examine their amorphization behavior under light water reactor (LWR) conditions. U 3Si 2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

Short Communication on "In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures"

NASA Astrophysics Data System (ADS)

Miao, Yinbin; Harp, Jason; Mo, Kun; Bhattacharya, Sumit; Baldo, Peter; Yacout, Abdellatif M.

2017-02-01

The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Buongiorno, Jacopo

2010-01-01

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors. FEAST-METAL was benchmarked against the open-literature EBR-II database for steady state and furnace tests (transients). The results show that the code is able to predict important phenomena such as clad strain, fission gas release, clad wastage, clad failure time, axial fuel slug deformation and fuel constituent redistribution, satisfactorily.

COL Application Content Guide for HTGRs: Revision to RG 1.206, Part 1 - Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wayne Moe

2012-08-01

A combined license (COL) application is required by the Nuclear Regulatory Commission (NRC) for all proposed nuclear plants. The information requirements for a COL application are set forth in 10 CFR 52.79, “Contents of Applications; Technical Information in Final Safety Analysis Report.” An applicant for a modular high temperature gas-cooled reactor (HTGR) must develop and submit for NRC review and approval a COL application which conforms to these requirements. The technical information necessary to allow NRC staff to evaluate a COL application and resolve all safety issues related to a proposed nuclear plant is detailed and comprehensive. To this, Regulatorymore » Guide (RG) 1.206, “Combined License Applications for Nuclear Power Plants” (LWR Edition), was developed to assist light water reactor (LWR) applicants in incorporating and effectively formatting required information for COL application review (Ref. 1). However, the guidance prescribed in RG 1.206 presumes a LWR design proposal consistent with the systems and functions associated with large LWR power plants currently operating under NRC license.« less

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

NASA Astrophysics Data System (ADS)

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng; Alankar, Alankar; Subramanian, Gopinath; Stanek, Christopher

2017-01-01

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

Waste Estimates for a Future Recycling Plant in the US Based Upon AREVA Operating Experience - 13206

DOE Office of Scientific and Technical Information (OSTI.GOV)

Foare, Genevieve; Meze, Florian; Bader, Sven

2013-07-01

Estimates of process and secondary wastes produced by a recycling plant built in the U.S., which is composed of a used nuclear fuel (UNF) reprocessing facility and a mixed oxide (MOX) fuel fabrication facility, are performed as part of a U.S. Department of Energy (DOE) sponsored study [1]. In this study, a set of common inputs, assumptions, and constraints were identified to allow for comparison of these wastes between different industrial teams. AREVA produced a model of a reprocessing facility, an associated fuel fabrication facility, and waste treatment facilities to develop the results for this study. These facilities were dividedmore » into a number of discrete functional areas for which inlet and outlet flow streams were clearly identified to allow for an accurate determination of the radionuclide balance throughout the facility and the waste streams. AREVA relied primarily on its decades of experience and feedback from its La Hague (reprocessing) and MELOX (MOX fuel fabrication) commercial operating facilities in France to support this assessment. However, to perform these estimates for a U.S. facility with different regulatory requirements and to take advantage of some technological advancements, such as in the potential treatment of off-gases, some deviations from this experience were necessary. A summary of AREVA's approach and results for the recycling of 800 metric tonnes of initial heavy metal (MTIHM) of LWR UNF per year into MOX fuel under the assumptions and constraints identified for this DOE study are presented. (authors)« less

Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Davis, W. Jr.; West, G.A.; Stacy, R.G.

1979-03-22

Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO/sub 2/ or 96 to 97% ThO/sub 2/--3 to 4% UO/sub 2/. Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO/sub 2/ or ThO/sub 2/--UO/sub 2/ sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared intomore » lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO/sub 2/ from BWRs and of Zircaloy-4-clad UO/sub 2/ from PWRs. Median particle sizes of UO/sub 2/ from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 ..mu..m; particle sizes of ThO/sub 2/--UO/sub 2/, under these same conditions, ranged from 137 to 202 ..mu..m. Similarly, median particle sizes of UO/sub 2/ from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 ..mu..m. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels; however, unirradiated fuel from vendors was not available for performing comparative shearing experiments. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution estimates can be made of fractions of dislodged fuel having dimensions less than specified values.« less

Magnesium transport extraction of transuranium elements from LWR fuel

DOEpatents

Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.; Pierce, R. Dean

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.

Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unal, Cetin; Galloway, Jack D.

2014-09-12

In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermalmore » swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.« less

Technologies that affect the weaning rate in beef cattle production systems.

PubMed

Dill, Matheus Dhein; Pereira, Gabriel Ribas; Costa, João Batista Gonçalves; Canellas, Leonardo Canali; Peripolli, Vanessa; Neto, José Braccini; Sant'Anna, Danilo Menezes; McManus, Concepta; Barcellos, Júlio Otávio Jardim

2015-10-01

We investigated the differences between weaning rates and technologies adopted by farmers in cow-calf production systems in Rio Grande do Sul State, Brazil. Interviews were carried out with 73 farmers about 48 technologies that could affect reproductive performance. Data were analyzed by multivariate analysis using a non-hierarchical cluster method. The level of significance was set at P < 0.05. Three distinct clusters of farmers were created (R (2) = 0.90), named as low (LWR), intermediate (IWR), and high (HWR) weaning rate, with 100, 91, and 96 % of the farmers identified within their respective groups and average weaning rates of 59, 72, and 83 %, respectively. IWR and HWR farmers used more improved natural pasture, fixed-time artificial insemination, selection for birth weight, and proteinated salt compared to LWR. HWR farmers used more stocking rate control, and IWR farmers used more ultrasound to evaluate reproductive performance compared to the LWR group. IWR and HWR adopted more technologies related to nutrition and reproductive aspects of the herd in comparison to LWR. We concluded that farmers with higher technology use on farm had higher weaning rates which could be used to benefit less efficient farmers.

EUV lithography for 30nm half pitch and beyond: exploring resolution, sensitivity, and LWR tradeoffs

NASA Astrophysics Data System (ADS)

Putna, E. Steve; Younkin, Todd R.; Chandhok, Manish; Frasure, Kent

2009-03-01

The International Technology Roadmap for Semiconductors (ITRS) denotes Extreme Ultraviolet (EUV) lithography as a leading technology option for realizing the 32nm half-pitch node and beyond. Readiness of EUV materials is currently one high risk area according to assessments made at the 2008 EUVL Symposium. The main development issue regarding EUV resist has been how to simultaneously achieve high sensitivity, high resolution, and low line width roughness (LWR). This paper describes the strategy and current status of EUV resist development at Intel Corporation. Data is presented utilizing Intel's Micro-Exposure Tool (MET) examining the feasibility of establishing a resist process that simultaneously exhibits <=30nm half-pitch (HP) L/S resolution at <=10mJ/cm2 with <=4nm LWR.

EUV lithography for 22nm half pitch and beyond: exploring resolution, LWR, and sensitivity tradeoffs

NASA Astrophysics Data System (ADS)

Putna, E. Steve; Younkin, Todd R.; Caudillo, Roman; Chandhok, Manish

2010-04-01

The International Technology Roadmap for Semiconductors (ITRS) denotes Extreme Ultraviolet (EUV) lithography as a leading technology option for realizing the 22nm half pitch node and beyond. Readiness of EUV materials is currently one high risk area according to recent assessments made at the 2009 EUVL Symposium. The main development issue regarding EUV resist has been how to simultaneously achieve high sensitivity, high resolution, and low line width roughness (LWR). This paper describes the strategy and current status of EUV resist development at Intel Corporation. Data collected utilizing Intel's Micro-Exposure Tool (MET) is presented in order to examine the feasibility of establishing a resist process that simultaneously exhibits <=22nm half-pitch (HP) L/S resolution at <= 12.5mJ/cm2 with <= 4nm LWR.

Environmental Effect on Evolutionary Cyclic Plasticity Material Parameters of 316 Stainless Steel: An Experimental & Material Modeling Approach

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin

2014-09-20

This report provides an update on an earlier assessment of environmentally assisted fatigue for light water reactor (LWR) materials under extended service conditions. This report is a deliverable under the work package for environmentally assisted fatigue in the Light Water Reactor Sustainability (LWRS) program. The overall objective of this LWRS project is to assess the degradation by environmentally assisted cracking/fatigue of LWR materials such as various alloy base metals and their welds used in reactor coolant system piping. This effort is to support the Department of Energy LWRS program for developing tools to understand the aging/failure mechanism and to predictmore » the remaining life of LWR components for anticipated 60-80 year operation.« less

Severe accident modeling of a PWR core with different cladding materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, S. C.; Henry, R. E.; Paik, C. Y.

2012-07-01

The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCSmore » rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)« less

Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

DOE PAGES

Koyanagi, Takaaki; Katoh, Yutai

2017-07-04

Silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites are being actively investigated for use in accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this paper examined SiC/SiC composites following neutron irradiation at 230–340 °C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials were chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC) -coated Hi-Nicalon™ Type-S (HNS), Tyranno™ SA3 (SA3), and SCS-Ultra™ (SCS) SiC fibers. The irradiation resistance of these composites wasmore » investigated based on flexural behavior, dynamic Young's modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young's moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. Finally, this study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.« less

Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koyanagi, Takaaki; Katoh, Yutai

Silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites are being actively investigated for use in accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this paper examined SiC/SiC composites following neutron irradiation at 230–340 °C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials were chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC) -coated Hi-Nicalon™ Type-S (HNS), Tyranno™ SA3 (SA3), and SCS-Ultra™ (SCS) SiC fibers. The irradiation resistance of these composites wasmore » investigated based on flexural behavior, dynamic Young's modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young's moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. Finally, this study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.« less

Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

NASA Astrophysics Data System (ADS)

Koyanagi, Takaaki; Katoh, Yutai

2017-10-01

Silicon carbide (SiC) fiber-reinforced SiC matrix (SiC/SiC) composites are being actively investigated for use in accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this study examined SiC/SiC composites following neutron irradiation at 230-340 °C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials were chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC) -coated Hi-Nicalon™ Type-S (HNS), Tyranno™ SA3 (SA3), and SCS-Ultra™ (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young's modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young's moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. This study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

Qualification of CASMO5 / SIMULATE-3K against the SPERT-III E-core cold start-up experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grandi, G.; Moberg, L.

SIMULATE-3K is a three-dimensional kinetic code applicable to LWR Reactivity Initiated Accidents. S3K has been used to calculate several international recognized benchmarks. However, the feedback models in the benchmark exercises are different from the feedback models that SIMULATE-3K uses for LWR reactors. For this reason, it is worth comparing the SIMULATE-3K capabilities for Reactivity Initiated Accidents against kinetic experiments. The Special Power Excursion Reactor Test III was a pressurized-water, nuclear-research facility constructed to analyze the reactor kinetic behavior under initial conditions similar to those of commercial LWRs. The SPERT III E-core resembles a PWR in terms of fuel type, moderator,more » coolant flow rate, and system pressure. The initial test conditions (power, core flow, system pressure, core inlet temperature) are representative of cold start-up, hot start-up, hot standby, and hot full power. The qualification of S3K against the SPERT III E-core measurements is an ongoing work at Studsvik. In this paper, the results for the 30 cold start-up tests are presented. The results show good agreement with the experiments for the reactivity initiated accident main parameters: peak power, energy release and compensated reactivity. Predicted and measured peak powers differ at most by 13%. Measured and predicted reactivity compensations at the time of the peak power differ less than 0.01 $. Predicted and measured energy release differ at most by 13%. All differences are within the experimental uncertainty. (authors)« less

Enhancing BWR proliferation resistance fuel with minor actinides

NASA Astrophysics Data System (ADS)

Chang, Gray S.

2009-03-01

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate-term of nuclear energy reconnaissance.

Development of the Mathematics of Learning Curve Models for Evaluating Small Modular Reactor Economics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, T. J.

2014-02-01

The cost of nuclear power is a straightforward yet complicated topic. It is straightforward in that the cost of nuclear power is a function of the cost to build the nuclear power plant, the cost to operate and maintain it, and the cost to provide fuel for it. It is complicated in that some of those costs are not necessarily known, introducing uncertainty into the analysis. For large light water reactor (LWR)-based nuclear power plants, the uncertainty is mainly contained within the cost of construction. The typical costs of operations and maintenance (O&M), as well as fuel, are well knownmore » based on the current fleet of LWRs. However, the last currently operating reactor to come online was Watts Bar 1 in May 1996; thus, the expected construction costs for gigawatt (GW)-class reactors in the United States are based on information nearly two decades old. Extrapolating construction, O&M, and fuel costs from GW-class LWRs to LWR-based small modular reactors (SMRs) introduces even more complication. The per-installed-kilowatt construction costs for SMRs are likely to be higher than those for the GW-class reactors based on the property of the economy of scale. Generally speaking, the economy of scale is the tendency for overall costs to increase slower than the overall production capacity. For power plants, this means that doubling the power production capacity would be expected to cost less than twice as much. Applying this property in the opposite direction, halving the power production capacity would be expected to cost more than half as much. This can potentially make the SMRs less competitive in the electricity market against the GW-class reactors, as well as against other power sources such as natural gas and subsidized renewables. One factor that can potentially aid the SMRs in achieving economic competitiveness is an economy of numbers, as opposed to the economy of scale, associated with learning curves. The basic concept of the learning curve is that the more a new process is repeated, the more efficient the process can be made. Assuming that efficiency directly relates to cost means that the more a new process is repeated successfully and efficiently, the less costly the process can be made. This factor ties directly into the factory fabrication and modularization aspect of the SMR paradigm—manufacturing serial, standardized, identical components for use in nuclear power plants can allow the SMR industry to use the learning curves to predict and optimize deployment costs.« less

Nuclear data uncertainty propagation by the XSUSA method in the HELIOS2 lattice code

NASA Astrophysics Data System (ADS)

Wemple, Charles; Zwermann, Winfried

2017-09-01

Uncertainty quantification has been extensively applied to nuclear criticality analyses for many years and has recently begun to be applied to depletion calculations. However, regulatory bodies worldwide are trending toward requiring such analyses for reactor fuel cycle calculations, which also requires uncertainty propagation for isotopics and nuclear reaction rates. XSUSA is a proven methodology for cross section uncertainty propagation based on random sampling of the nuclear data according to covariance data in multi-group representation; HELIOS2 is a lattice code widely used for commercial and research reactor fuel cycle calculations. This work describes a technique to automatically propagate the nuclear data uncertainties via the XSUSA approach through fuel lattice calculations in HELIOS2. Application of the XSUSA methodology in HELIOS2 presented some unusual challenges because of the highly-processed multi-group cross section data used in commercial lattice codes. Currently, uncertainties based on the SCALE 6.1 covariance data file are being used, but the implementation can be adapted to other covariance data in multi-group structure. Pin-cell and assembly depletion calculations, based on models described in the UAM-LWR Phase I and II benchmarks, are performed and uncertainties in multiplication factor, reaction rates, isotope concentrations, and delayed-neutron data are calculated. With this extension, it will be possible for HELIOS2 users to propagate nuclear data uncertainties directly from the microscopic cross sections to subsequent core simulations.

LWR First Recycle of TRU with Thorium Oxide for Transmutation and Cross Sections

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrea Alfonsi; Gilles Youinou; Sonat Sen

2013-02-01

Thorium has been considered as an option to uranium-based fuel, based on considerations of resource utilization (thorium is approximately three times more plentiful than uranium) and as a result of concerns about proliferation and waste management (e.g. reduced production of plutonium, etc.). Since the average composition of natural Thorium is dominated (100%) by the fertile isotope Th-232, Thorium is only useful as a resource for breeding new fissile materials, in this case U-233. Consequently a certain amount of fissile material must be present at the start-up of the reactor in order to guarantee its operation. The thorium fuel can bemore » used in both once-through and recycle options, and in both fast and thermal spectrum systems. The present study has been aimed by the necessity of investigating the option of using reprocessed plutonium/TRU, from a once-through reference LEU scenario (50 GWd/ tIHM), mixed with natural thorium and the need of collect data (mass fractions, cross-sections etc.) for this particular fuel cycle scenario. As previously pointed out, the fissile plutonium is needed to guarantee the operation of the reactor. Four different scenarios have been considered: • Thorium – recycled Plutonium; • Thorium – recycled Plutonium/Neptunium; • Thorium – recycled Plutonium/Neptunium/Americium; • Thorium – recycled Transuranic. The calculations have been performed with SCALE6.1-TRITON.« less

LWR First Recycle of TRU with Thorium Oxide for Transmutation and Cross Sections

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrea Alfonsi; Gilles Youinou

2012-07-01

Thorium has been considered as an option to uranium-based fuel, based on considerations of resource utilization (thorium is approximately three times more plentiful than uranium) and as a result of concerns about proliferation and waste management (e.g. reduced production of plutonium, etc.). Since the average composition of natural Thorium is dominated (100%) by the fertile isotope Th-232, Thorium is only useful as a resource for breeding new fissile materials, in this case U-233. Consequently a certain amount of fissile material must be present at the start-up of the reactor in order to guarantee its operation. The thorium fuel can bemore » used in both once-through and recycle options, and in both fast and thermal spectrum systems. The present study has been aimed by the necessity of investigating the option of using reprocessed plutonium/TRU, from a once-through reference LEU scenario (50 GWd/ tIHM), mixed with natural thorium and the need of collect data (mass fractions, cross-sections etc.) for this particular fuel cycle scenario. As previously pointed out, the fissile plutonium is needed to guarantee the operation of the reactor. Four different scenarios have been considered: • Thorium – recycled Plutonium; • Thorium – recycled Plutonium/Neptunium; • Thorium – recycled Plutonium/Neptunium/Americium; • Thorium – recycled Transuranic. The calculations have been performed with SCALE6.1-TRITON.« less

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert, E-mail: robert.montgomery@pnnl.gov; Tomé, Carlos, E-mail: tome@lanl.gov; Liu, Wenfeng, E-mail: wenfeng.liu@anatech.com

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC)more » polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.« less

Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

DOE PAGES

Yamamoto, Yukinori; Pint, Bruce A.; Terrani, Kurt A.; ...

2015-10-19

Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitivemore » to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741°C.« less

Propagation of neutron-reaction uncertainties through multi-physics models of novel LWR's

NASA Astrophysics Data System (ADS)

Hernandez-Solis, Augusto; Sjöstrand, Henrik; Helgesson, Petter

2017-09-01

The novel design of the renewable boiling water reactor (RBWR) allows a breeding ratio greater than unity and thus, it aims at providing for a self-sustained fuel cycle. The neutron reactions that compose the different microscopic cross-sections and angular distributions are uncertain, so when they are employed in the determination of the spatial distribution of the neutron flux in a nuclear reactor, a methodology should be employed to account for these associated uncertainties. In this work, the Total Monte Carlo (TMC) method is used to propagate the different neutron-reactions (as well as angular distributions) covariances that are part of the TENDL-2014 nuclear data (ND) library. The main objective is to propagate them through coupled neutronic and thermal-hydraulic models in order to assess the uncertainty of important safety parameters related to multi-physics, such as peak cladding temperature along the axial direction of an RBWR fuel assembly. The objective of this study is to quantify the impact that ND covariances of important nuclides such as U-235, U-238, Pu-239 and the thermal scattering of hydrogen in H2O have in the deterministic safety analysis of novel nuclear reactors designs.

In vivo quantification of response to treatment in patients with multiple myeloma by 1H magnetic resonance spectroscopy of bone marrow.

PubMed

Oriol, Albert; Valverde, Daniel; Capellades, Jaume; Cabañas, Miquel E; Ribera, Josep-Maria; Arús, Carles

2007-04-01

Magnetic resonance imaging (MRI) is the gold standard non-invasive technique to detect malignant disease in the bone marrow. Proton magnetic resonance spectroscopy (MRS) can be performed as a quick adjunct to routine spinal MRI. We performed proton MRS to patients with multiple myeloma (MM) at diagnosis and after treatment to investigate the possible correlation of MRS data with response to therapy. Twenty-one patients with newly diagnosed MM underwent combined MRI/MRS explorations of a transverse center section in the fifth lumbar vertebral body. MRS was acquired with STEAM and 40 ms TE. Areas of unsuppressed water and lipid resonances were used to calculate the lipid-to-water ratio (LWR). No association was detected between initial LWRs and the clinical characteristics of patients. Post treatment MRS was available in 16 patients of whom 11 (69%) presented an LWR increase, this included all complete responders (8/8, 100%, P = 0.012). A post-treatment LWR value equal to or larger than one is proposed as a non-invasive marker of complete response to treatment. Only patients responding to treatment presented a significant increase in bone marrow LWR after therapy. MRS may provide an adequate quantification of response to chemotherapy in patients with MM.

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

EUV process establishment through litho and etch for N7 node

NASA Astrophysics Data System (ADS)

Kuwahara, Yuhei; Kawakami, Shinichiro; Kubota, Minoru; Matsunaga, Koichi; Nafus, Kathleen; Foubert, Philippe; Mao, Ming

2016-03-01

Extreme ultraviolet lithography (EUVL) technology is steadily reaching high volume manufacturing for 16nm half pitch node and beyond. However, some challenges, for example scanner availability and resist performance (resolution, CD uniformity (CDU), LWR, etch behavior and so on) are remaining. Advance EUV patterning on the ASML NXE:3300/ CLEAN TRACK LITHIUS Pro Z- EUV litho cluster is launched at imec, allowing for finer pitch patterns for L/S and CH. Tokyo Electron Ltd. and imec are continuously collabo rating to develop manufacturing quality POR processes for NXE:3300. TEL's technologies to enhance CDU, defectivity and LWR/LER can improve patterning performance. The patterning is characterized and optimized in both litho and etch for a more complete understanding of the final patterning performance. This paper reports on post-litho CDU improvement by litho process optimization and also post-etch LWR reduction by litho and etch process optimization.

Theoretical study on effects of photodecomposable quenchers in line-and-space pattern fabrication with 7 nm quarter-pitch using chemically amplified electron beam resist process

NASA Astrophysics Data System (ADS)

Kozawa, Takahiro

2017-04-01

The line width roughness (LWR) is a significant issue in the development of chemically amplified resists. The increase in sensitizer concentration is inevitable for the suppression of LWR in the sub-10 nm fabrication. In this study, we investigated the effects of photodecomposable quenchers from the viewpoint of the excluded volume effect, assuming line-and-space patterns with 7 nm quarter-pitch (7 nm space width and 28 nm pitch). The pattern formation of chemically amplified electron beam resists with photodecomposable quenchers was calculated and compared with those with conventional quenchers. It was found that the sum of the concentrations of acid generators and quenchers (photodecomposable or conventional quenchers) can be reduced without decreasing the chemical gradient (an indicator of LWR) by using the photodecomposable quenchers. The photodecomposable quenchers are considered essential in the high-resolution fabrication.

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

Experimental Recreation of Large-Scale Coastal Bedforms and Hummocky Cross-Stratification in Sheet Flow Conditions

NASA Astrophysics Data System (ADS)

Vermaas, T.; Kleinhans, M. G.; Huisman, C.; Schretlen, J. L.; van der Werf, J. J.; Ribberink, J. S.; Ruessink, G.

2010-12-01

In shallow marine environments various types of large bed forms emerge under waves and currents. There is no consensus on whether and how these bedforms can be classified in a genetically meaningful sense. Hypotheses for their genesis vary from a large variety of causal mechanisms for a number of different ripples to a single growing instability mechanism, reflecting a limited understanding. Our objective is to understand the formative mechanism of a family of large bedforms referred to as Large Wave Ripples in coastal literature and Hummocks in sedimentological literature, which also describes the hummocky cross stratification (HCS) found in the sedimentary rock record. The formative conditions for hummocks have been debated extensively, particularly whether currents or specific particle sizes were required. We collected and compared existing field and laboratory data and we conducted a full scale experiment in the Hannover Grosse Welle wave flume (300 m long, 5 m wide and 7 m deep). Experiments were done for several conditions, including a storm sequence, with 0.7-1.7 m regular trochoidal waves or irregular waves with periods of 5-7.5 s over sand with mean particle sizes of 0.256 (in 2007) or 0.137 mm (in 2008). Bed profiles were collected mechanically and acoustically. A conductivity probe (CCM) was used to measure sheet flow thickness or absence and near-bed flow and suspended sand concentrations were measured in detail with acoustical profilers. From the data collection, we found that there is no distinction empirically between LWR and Hummocks. Both are found around the inception of sheet flow and have the same dimensions. In the experiments we produced short wave ripples superimposed on large wave ripples below and in the transition to sheet flow conditions. The SWR were well predicted by a recent particle-size dependent ripple length predictor. No available predictor matched the LWR dimensions. The LWR remained present in strong sheet flow conditions and migrated slowly in the direction of wave advance due to wave asymmetry. LWR height was less than 0.07 m whilst lengths were about 13 m. Despite the sheet flow conditions and fine sediment, the LWR scaled as orbital ripples though a factor of 2 longer (i.e. with the orbital diameter d = uT/pi with u the orbital velocity amplitude and T the wave period). Laquer peels of the 2007 experiment demonstrated that the LWR formed Hummocky Cross-Stratification. We conclude that hummocks were experimentally created in a full-scale facility during sheet flow conditions without currents. Furthermore, LWR and hummocks are the same features.

BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wirth, Brian; Terrani, Kurt A.; Sweet, Ryan T.

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling.more » As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations of the IFA-796 rod 4 experiment, but show that varying the creep model (within the range investigated here) only provide a minimal increase in the fuel radius and maximum cladding hoop stress. Continued investigation of fuel behavioral models will include benchmarking the modified fuel creep model against available experimental data, as well as an investigation of the role that fuel cracking will play in the compliance of the fuel. Correctly calculating stress evolution in the fuel is key to assessing fuel behavior up to gap closure and the subsequent deformation of the cladding due to PCMI. The inclusion of frictional contact should also be investigated to determine the axial elongation of the fuel rods for comparison with data from this experiment.« less

Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sickafus, Kurt E.; Wirth, Brian; Miller, Larry

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectivesmore » of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as the possibilities for enhanced fuel/clad system performance and longevity.« less

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sweet, Ryan; George, Nathan M.; Terrani, Kurt A.

2016-08-30

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling themore » integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.« less

Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durkee, Jr., Joe W.

A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20,more » 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/ 137Cs 134Cs/ 154Eu, and 154Eu/ 137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the FA models serve as source materials for the pre- and postelectrorefining models to be reported in Parts 2 and 3.« less

Measurements and sensitivities of LWR in poly spacers

NASA Astrophysics Data System (ADS)

Ayal, Guy; Shauly, Eitan; Levi, Shimon; Siany, Amit; Adan, Ofer; Shacham-Diamand, Yosi

2010-03-01

LER and LWR have long been considered a primary issue in process development and monitoring. Development of a low power process flavors emphasizes the effect of LER, LWR on different aspects of the device. Gate level performance, particularly leakage current at the front end of line, resistance and reliability in the back-end layers. Traditionally as can be seen in many publications, for the front end of line the focus is mainly on Poly and Active area layers. Poly spacers contribution to the gate leakage, for example, is rarely discussed. Following our research done on sources of gate leakage, we found leakage current (Ioff) in some processes to be highly sensitive to changes in the width of the Poly spacers - even more strongly to the actual Poly gate CDs. Therefore we decided to measure Poly spacers LWR, its correlation to the LWR in the poly, and its sensitivity to changes in layout and OPC. In our last year publication, we defined the terms LLER (Local Line Edge Roughness) and LLWR (Local Line Width Roughness). The local roughness is measured as the 3-sigma value of the line edge/width in a 5-nm segment around the measurement point. We will use these terms in this paper to evaluate the Poly roughness impact on Poly spacer's roughness. A dedicated test chip was designed for the experiments, having various transistors layout configurations with different densities to cover the all range of process design rules. Applied Materials LER and LWR innovative algorithms were used to measure and characterize the spacer roughness relative to the distance from the active edges and from other spaces. To accurately measure all structures in a reasonable time, the recipes were automatically generated from CAD. On silicon, after poly spacers generation, the transistors no longer resemble the Poly layer CAD layout, their morphology is different compared with Photo/Etch traditional structures , and dimensions vary significantly. In this paper we present metrology and characterization of poly spacer LLWR and LLER compared to that of the poly gate in various transistor shapes, showing that the relation between them depends on the transistor architecture (final layout, including OPC). We will show how the spacer deposition may reduce, keep or even enlarge the roughness measured on Poly, depending on transistor layout , but surprisingly, not dependent on proximity effects.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sen, Ramazan Sonat; Hummel, Andrew John; Hiruta, Hikaru

The deterministic full core simulators require homogenized group constants covering the operating and transient conditions over the entire lifetime. Traditionally, the homogenized group constants are generated using lattice physics code over an assembly or block in the case of prismatic high temperature reactors (HTR). For the case of strong absorbers that causes strong local depressions on the flux profile require special techniques during homogenization over a large volume. Fuel blocks with burnable poisons or control rod blocks are example of such cases. Over past several decades, there have been a tremendous number of studies performed for improving the accuracy ofmore » full-core calculations through the homogenization procedure. However, those studies were mostly performed for light water reactor (LWR) analyses, thus, may not be directly applicable to advanced thermal reactors such as HTRs. This report presents the application of SuPer-Homogenization correction method to a hypothetical HTR core.« less

Lower Length Scale Characterization and Validation of Formation and Stability of Helium Bubbles in Nano-structured Ferritic Alloys under Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Huijuan; Yun, Di; Hoelzer, David

In order to extend the operating license of current light water reactors (LWRs) in the United States and other countries to as many as 80 years or longer, it is demanding to identify potential materials for many of the internal structural components and fasteners. We proposed that 14YWT iron alloy can be adopted in such applications with its excellent material properties, such as high-temperature strength, low creep rate, and high irradiation resistance. Application with 14YWT would improve the void/helium swelling characteristics of the LWR fuels, extend the burn-up limits with the tolerant temperature up to 800oC and reduce the hydrogenmore » production. One key feature of 14YWT material property enhancement is the ultrafine high density of 2-4nm Y-Ti-O enriched nanoclusters (NCs) within the 14YWT iron matrix. The NCs can effectively pin the ultra-fine grain boundaries and dislocations, which significantly enhance mechanical properties of the alloy. Moreover, these nanoclusters remain stable with no coarsening after a large dose of ion irradiation. After ion irradiation, the helium bubbles are observed extremely uniform in size (1nm) and quite homogeneously distributed within the 14YWT matrix, which indicates that the microstructure of 14YWT remains remarkably tolerance to radiation damage. However, there is a lack of understanding of 14YWT both theoretically and experimentally in order to understand the mechanism behind the material property enhancement and to further develop and design a new generation of advanced structural material for current LWR applications and future fusion applications.« less

Advanced Fuels Campaign FY 2014 Accomplishments Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Braase, Lori; May, W. Edgar

The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This includes development of a state-of-the art Research and Development (R&D) infrastructure to support the use of a “goal-oriented science-based approach.” In support of the Fuel Cycle Research and Development (FCRD) program, AFC is responsible for developing advanced fuels technologies to support the various fuel cyclemore » options defined in the Department of Energy (DOE) Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. AFC uses a “goal-oriented, science-based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. The modeling and simulation activities for fuel performance are carried out under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which is closely coordinated with AFC. In this report, the word “fuel” is used generically to include fuels, targets, and their associated cladding materials. R&D of light water reactor (LWR) fuels with enhanced accident tolerance is also conducted by AFC. These fuel systems are designed to achieve significantly higher fuel and plant performance to allow operation to significantly higher burnup, and to provide enhanced safety during design basis and beyond design basis accident conditions. The overarching goal is to develop advanced nuclear fuels and materials that are robust, have high performance capability, and are more tolerant to accident conditions than traditional fuel systems. AFC management and integration activities included continued support for international collaborations, primarily with France, Japan, the European Union, Republic of Korea, and China, as well as various working group and expert group activities in the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD-NEA) and the International Atomic Energy Agency (IAEA). Three industry-led Funding Opportunity Announcements (FOAs) and two university-led Integrated Research Projects (IRPs), funded in 2013, made significant progress in fuels and materials development. All are closely integrated with AFC and Accident Tolerant Fuels (ATF) research. Accomplishments made during fiscal year (FY) 2014 are highlighted in this report, which focuses on completed work and results. The process details leading up to the results are not included; however, the lead technical contact is provided for each section.« less

Salt transport extraction of transuranium elements from LWR fuel

DOEpatents

Pierce, R.D.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Miller, W.E.

1992-11-03

A process is described for separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl[sub 2] and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750 C to about 850 C to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl[sub 2] having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO[sub 2]. The Ca metal and CaCl[sub 2] is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including MgCl[sub 2] to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy. 2 figs.

Salt transport extraction of transuranium elements from lwr fuel

DOEpatents

Pierce, R. Dean; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750.degree. C. to about 850.degree. C. to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including Mg Cl.sub.2 to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy.

Evanescent wave absorbance based fiber optic biosensor for label-free detection of E. coli at 280 nm wavelength.

PubMed

Bharadwaj, Reshma; Sai, V V R; Thakare, Kamini; Dhawangale, Arvind; Kundu, Tapanendu; Titus, Susan; Verma, Pradeep Kumar; Mukherji, Soumyo

2011-03-15

A novel label-free technique for the detection of pathogens based on evanescent wave absorbance (EWA) changes at 280 nm from a U-bent optical fiber sensor is demonstrated. Bending a decladded fiber into a U-shaped structure enhances the penetration depth of evanescent waves and hence sensitivity of the probe. We show that the enhanced EWA response from such U-bent probes, caused by the inherent optical absorbance properties of bacterial cells or biomolecules specifically bound to the sensor surface, can be exploited for the detection of pathogens. A portable optical set-up with a UV light emitting diode, a spectrometer and U-bent fiber optic probe of 200 μm core diameter, 0.75 mm bend radius and effective probe length of 1cm demonstrated an ability to detect less than 1000 cfu/ml. Copyright © 2011. Published by Elsevier B.V.

Severe Accident Test Station Activity Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pint, Bruce A.; Terrani, Kurt A.

2015-06-01

Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accidentmore » Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.« less

Progress In Developing Laser Based Post Irradiation Examination Infrastructure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, James A.; Scott, Clark L.; Benefiel, Brad C.

To be able to understand the performance of reactor fuels and materials, irradiated materials must be characterized effectively and efficiently in a high rad environment. The characterization work must be performed remotely and in an environment hostile to instrumentation. Laser based characterization techniques provide the ability to be remote and robust in a hot-cell environment. Laser based instrumentation also can provide high spatial resolution suitable for scanning and imaging large areas. The INL is currently developing three laser based Post Irradiation Examination (PIE) stations for the Hot Fuel Examination Facility at the INL. These laser based systems will characterize irradiatedmore » materials and fuels. The characterization systems are the following: Laser Shock Laser based ultrasonic C-scan system Gas Assay, Sample, and Recharge system (GASR, up-grade to an existing system). The laser shock technique will characterize material properties and failure loads/mechanisms in various materials such as LWR fuel, plate fuel, and next generation fuel forms, for PIE in high radiation areas. The laser shock-technique induces large amplitude shock waves to mechanically characterize interfaces such as the fuel-clad bond. The shock wave travels as a compression wave through the material to the free (unconfined) back surface and reflects back through the material under test as a rarefaction (tensile) wave. This rarefaction wave is the physical mechanism that produces internal de-lamination failure. As part of the laser shock system, a laser-based ultrasonic C-scan system will be used to detect and characterize debonding caused by the laser shock technique. The laser ultrasonic system will be fully capable of performing classical non-destructive evaluation testing and imaging functions such as microstructure characterization, flaw detection and dimensional metrology in complex components. The purpose of the GASR is to measure the pressure/volume of the plenum of an irradiated fuel element and obtain fission gas samples for analysis. The study of pressure and volume in the plenum of an irradiated fuel element and the analysis of fission gases released from the fuel is important to understanding the performance of reactor fuels and materials. This system may also be used to measure the pressure/volume of other components (such as control blades) and obtain gas samples from these components for analysis. The main function of the laser in this application is to puncture the fuel element to allow the fission gas to escape and if necessary to weld the spot close. The GASR station will have the inherent capability to perform cutting welding and joining functions within a hot-cell.« less

Implications of Zircaloy creep and growth to light water reactor performance

NASA Astrophysics Data System (ADS)

Franklin, David G.; Adamson, Ronald B.

1988-10-01

Deformation of zirconium alloy components in nuclear reactors has been a concern since the decision of Admiral Rickover to use them in the US Navy submarine reactors. With the exception of the first few light water reactors (LWRs) most of the core structural materials have been fabricated from either Zircaloy-2 or Zircaloy-4. Performance of these alloys has been extremely good, even though the effects of irradiation on deformation magnitudes and mechanisms were not fully appreciated until extensive service and in-reactor tests were accomplished. Since the reactor components are designed to operate at stress levels well below yield for normal conditions, the only significant deformation is time dependent. Although creep was anticipated, the enhancement by neutron irradiation and the stress-free, nearly constant-volume shape change known as irradiation growth were not known prior to materials testing in reactors under controlled conditions. Both of these phenomena have significant impact on performance and must be accounted for properly in design. Although irradiation creep and growth have resulted in only one significant performance problem (creep collapse of fuel cladding, which has been eliminated), deformation magnitudes and, particularly, differentials in strain magnitudes, are a continuing source of interest. Factors that affect dimensional stability due to both creep and growth include temperature, fluence, residual stress, texture, and microstructure. The first two are reactor variables and the others are related to component fabrication history. This paper includes a review of the applications of Zircaloy creep and growth to LWR fuel designs, a review of the impact of in-reactor creep and growth on fuel rod and fuel assembly performance, and comments on potential improvements. Since the reactor design, fuel design and the core environment in BWRs and PWRs are quite different, appropriate separation of the application of effects are made; of course, the basic phenomena are the same in both systems.

Measures of the environmental footprint of the front end of the nuclear fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

E. Schneider; B. Carlsen; E. Tavrides

2013-11-01

Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle (FEFC) have focused primarily on energy consumption and CO2 emissions. Results have varied widely. This work builds upon reports from operating facilities and other primary data sources to build a database of front end environmental impacts. This work also addresses land transformation and water withdrawals associated with the processes of the FEFC. These processes include uranium extraction, conversion, enrichment, fuel fabrication, depleted uranium disposition, and transportation. To allow summing the impacts across processes, all impacts were normalized per tonne of natural uranium mined as wellmore » as per MWh(e) of electricity produced, a more conventional unit for measuring environmental impacts that facilitates comparison with other studies. This conversion was based on mass balances and process efficiencies associated with the current once-through LWR fuel cycle. Total energy input is calculated at 8.7 x 10- 3 GJ(e)/MWh(e) of electricity and 5.9 x 10- 3 GJ(t)/MWh(e) of thermal energy. It is dominated by the energy required for uranium extraction, conversion to fluoride compound for subsequent enrichment, and enrichment. An estimate of the carbon footprint is made from the direct energy consumption at 1.7 kg CO2/MWh(e). Water use is likewise dominated by requirements of uranium extraction, totaling 154 L/MWh(e). Land use is calculated at 8 x 10- 3 m2/MWh(e), over 90% of which is due to uranium extraction. Quantified impacts are limited to those resulting from activities performed within the FEFC process facilities (i.e. within the plant gates). Energy embodied in material inputs such as process chemicals and fuel cladding is identified but not explicitly quantified in this study. Inclusion of indirect energy associated with embodied energy as well as construction and decommissioning of facilities could increase the FEFC energy intensity estimate by a factor of up to 2.« less

Sensitivity Analysis and Optimization of the Nuclear Fuel Cycle: A Systematic Approach

NASA Astrophysics Data System (ADS)

Passerini, Stefano

For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon as technically feasible in order to extend the nuclear fuel resources. More recently, arguments have been made for deployment of fast reactors in order to reduce the amount of higher actinides, hence the longevity of radioactivity, in the materials destined to a geologic repository. The cost of the fast reactors, together with concerns about the proliferation of the technology of extraction of plutonium from used LWR fuel as well as the large investments in construction of reprocessing facilities have been the basis for arguments to defer the introduction of recycling technologies in many countries including the US. In this thesis, the impacts of alternative reactor technologies on the fuel cycle are assessed. Additionally, metrics to characterize the fuel cycles and systematic approaches to using them to optimize the fuel cycle are presented. The fuel cycle options of the 2010 MIT fuel cycle study are re-examined in light of the expected slower rate of growth in nuclear energy today, using the CAFCA (Code for Advanced Fuel Cycle Analysis). The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycle options available in the future. The options include limited recycling in L WRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. Additional fuel cycle scenarios presented for the first time in this work assume the deployment of innovative recycling reactor technologies such as the Reduced Moderation Boiling Water Reactors and Uranium-235 initiated Fast Reactors. A sensitivity study focused on system and technology parameters of interest has been conducted to test the robustness of the conclusions presented in the MIT Fuel Cycle Study. These conclusions are found to still hold, even when considering alternative technologies and different sets of simulation assumptions. Additionally, a first of a kind optimization scheme for the nuclear fuel cycle analysis is proposed and the applications of such an optimization are discussed. Optimization metrics of interest for different stakeholders in the fuel cycle (economics, fuel resource utilization, high level waste, transuranics/proliferation management, and environmental impact) are utilized for two different optimization techniques: a linear one and a stochastic one. Stakeholder elicitation provided sets of relative weights for the identified metrics appropriate to each stakeholder group, which were then successfully used to arrive at optimum fuel cycle configurations for recycling technologies. The stochastic optimization tool, based on a genetic algorithm, was used to identify non-inferior solutions according to Pareto's dominance approach to optimization. The main tradeoff for fuel cycle optimization was found to be between economics and most of the other identified metrics. (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

Tools for placing the radiological health hazard in perspective following a severe emergency at a light water reactor (LWR) or its spent fuel pool.

PubMed

McKenna, Thomas; Welter, Phillip Vilar; Callen, Jessica; Martincic, Rafael; Dodd, Brian; Kutkov, Vladimir

2015-01-01

Experience from past nuclear and radiological emergencies shows that placing the radiological health hazard in perspective and having a definition of "safe" are required in order to prevent members of the public, those responsible for protecting the public (i.e., decision makers), and others from taking inappropriate and damaging actions that are not justified based on the radiological health hazard. The principle concerns of the public during a severe nuclear power plant or spent fuel pool emergency are "Am I safe?" and "What should I do to be safe?" However, these questions have not been answered to the satisfaction of the public, despite various protective actions being implemented to ensure their safety. Instead, calculated doses or various measured quantities (e.g., ambient dose rate or radionuclide concentrations) are used to describe the situation to the public without placing them into perspective in terms of the possible radiological health hazard, or if they have, it has been done incorrectly. This has contributed to members of the public taking actions that do more harm than good in the belief that they are protecting themselves. Based on established international guidance, this paper provides a definition of "safe" for the radiological health hazard for use in nuclear or radiological emergencies and a system for putting the radiological health hazard in perspective for quantities most commonly measured after a release resulting from a severe emergency at a light water reactor or its spent fuel pool.

Capillary Rise: Validity of the Dynamic Contact Angle Models.

PubMed

Wu, Pingkeng; Nikolov, Alex D; Wasan, Darsh T

2017-08-15

The classical Lucas-Washburn-Rideal (LWR) equation, using the equilibrium contact angle, predicts a faster capillary rise process than experiments in many cases. The major contributor to the faster prediction is believed to be the velocity dependent dynamic contact angle. In this work, we investigated the dynamic contact angle models for their ability to correct the dynamic contact angle effect in the capillary rise process. We conducted capillary rise experiments of various wetting liquids in borosilicate glass capillaries and compared the model predictions with our experimental data. The results show that the LWR equations modified by the molecular kinetic theory and hydrodynamic model provide good predictions on the capillary rise of all the testing liquids with fitting parameters, while the one modified by Joos' empirical equation works for specific liquids, such as silicone oils. The LWR equation modified by molecular self-layering model predicts well the capillary rise of carbon tetrachloride, octamethylcyclotetrasiloxane, and n-alkanes with the molecular diameter or measured solvation force data. The molecular self-layering model modified LWR equation also has good predictions on the capillary rise of silicone oils covering a wide range of bulk viscosities with the same key parameter W(0), which results from the molecular self-layering. The advantage of the molecular self-layering model over the other models reveals the importance of the layered molecularly thin wetting film ahead of the main meniscus in the energy dissipation associated with dynamic contact angle. The analysis of the capillary rise of silicone oils with a wide range of bulk viscosities provides new insights into the capillary dynamics of polymer melts.

Pretest predictions for the response of a 1:8-scale steel LWR containment building model to static overpressurization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clauss, D.B.

The analyses used to predict the behavior of a 1:8-scale model of a steel LWR containment building to static overpressurization are described and results are presented. Finite strain, large displacement, and nonlinear material properties were accounted for using finite element methods. Three-dimensional models were needed to analyze the penetrations, which included operable equipment hatches, personnel lock representations, and a constrained pipe. It was concluded that the scale model would fail due to leakage caused by large deformations of the equipment hatch sleeves. 13 refs., 34 figs., 1 tab.

Fabrication Control Plan for ORNL RH-LOCA ATF Test Specimens to be Irradiated in the ATR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Howard, Richard; Teague, Michael

2014-06-01

The purpose of this fabrication plan is (1) to summarize the design of a set of rodlets that will be fabricated and then irradiated in the Advanced Test Reactor (ATR) and (2) provide requirements for fabrication and acceptance criteria for inspections of the Light Water Reactor (LWR) – Accident Tolerant Fuels (ATF) rodlet components. The functional and operational (F&OR) requirements for the ATF program are identified in the ATF Test Plan. The scope of this document only covers fabrication and inspections of rodlet components detailed in drawings 604496 and 604497. It does not cover the assembly of these items tomore » form a completed test irradiation assembly or the inspection of the final assembly, which will be included in a separate INL final test assembly specification/inspection document. The controls support the requirements that the test irradiations must be performed safely and that subsequent examinations must provide valid results.« less

Experience gained with the Synroc demonstration plant at ANSTO and its relevance to plutonium immobilization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jostsons, A.; Ridal, A.; Mercer, D.J.

1996-05-01

The Synroc Demonstration Plant (SDP) was designed and constructed at Lucas Heights to demonstrate the feasibility of Synroc production on a commercial scale (10 kg/hr) with simulated Purex liquid HLW. Since commissioning of the SDP in 1987, over 6000 kg of Synroc has been fabricated with a range of feeds and waste loadings. The SDP utilises uniaxial hot-pressing to consolidate Synroc. Pressureless sintering and hot-isostatic pressing have also been studied at smaller scales. The results of this extensive process development have been incorporated in a conceptual design for a radioactive plant to condition HLW from a reprocessing plant with amore » capacity to treat 800 tpa of spent LWR fuel. Synroic containing TRU, including Pu, and fission products has been fabricated and characterised in a glove-box facility and hot cells, respectively. The extensive experience in processing of Synroc over the past 15 years is summarised and its relevance to immobilization of surplus plutonium is discussed.« less

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope

2011-10-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh

2010-09-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

Radioactive Waste Conditioning, Immobilisation, And Encapsulation Processes And Technologies: Overview And Advances (Chapter 7)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jantzen, Carol M.; Lee, William E.; Ojovan, Michael I.

The main immobilization technologies that are available commercially and have been demonstrated to be viable are cementation, bituminization, and vitrification. Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either alkali borosilicate glass or alkali aluminophosphate glass. The exact compositions of nuclear waste glasses are tailored for easy preparation and melting, avoidance of glass-in-glass phase separation, avoidance of uncontrolled crystallization, and acceptable chemical durability, e.g., leach resistance. Glass has also been used to stabilize a variety of lowmore » level wastes (LLW) and mixed (radioactive and hazardous) low level wastes (MLLW) from other sources such as fuel rod cladding/decladding processes, chemical separations, radioactive sources, radioactive mill tailings, contaminated soils, medical research applications, and other commercial processes. The sources of radioactive waste generation are captured in other chapters in this book regarding the individual practices in various countries (legacy wastes, currently generated wastes, and future waste generation). Future waste generation is primarily driven by interest in sources of clean energy and this has led to an increased interest in advanced nuclear power production. The development of advanced wasteforms is a necessary component of the new nuclear power plant (NPP) flowsheets. Therefore, advanced nuclear wasteforms are being designed for robust disposal strategies. A brief summary is given of existing and advanced wasteforms: glass, glass-ceramics, glass composite materials (GCM’s), and crystalline ceramic (mineral) wasteforms that chemically incorporate radionuclides and hazardous species atomically in their structure. Cementitious, geopolymer, bitumen, and other encapsulant wasteforms and composites that atomically bond and encapsulate wastes are also discussed. The various processing technologies are cross-referenced to the various types of wasteforms since often a particular type of wasteform can be made by a variety of different processing technologies.« less

Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

NASA Astrophysics Data System (ADS)

Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

2017-01-01

The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

High Burnup Effects Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barner, J.O.; Cunningham, M.E.; Freshley, M.D.

1990-04-01

This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water reactor (LWR) fuel irradiated to high burnup levels. The HBEP was organized into three tasks as follows: Task 1 -- high burnup effects evaluations; Task 2 -- fission gas sampling; and Task 3 -- parameter effects study. During the coursemore » of the HBEP, a program that extended over 10 years, 82 fuel rods from a variety of sources were characterized, irradiated, and then examined in detail after irradiation. The study of fission gas release at high burnup levels was the principal objective of the program and it may be concluded that no significant enhancement of fission gas release at high burnup levels was observed for the examined rods. The rim effect, an as yet unquantified contributor to athermal fission gas release, was concluded to be the one truly high-burnup effect. Though burnup enhancement of fission gas release was observed to be low, a full understanding of the rim region and rim effect has not yet emerged and this may be a potential area of further research. 25 refs., 23 figs., 4 tabs.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Epiney, A.; Canepa, S.; Zerkak, O.

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

Double-deprotected chemically amplified photoresists (DD-CAMP): higher-order lithography

NASA Astrophysics Data System (ADS)

Earley, William; Soucie, Deanna; Hosoi, Kenji; Takahashi, Arata; Aoki, Takashi; Cardineau, Brian; Miyauchi, Koichi; Chun, Jay; O'Sullivan, Michael; Brainard, Robert

2017-03-01

The synthesis and lithographic evaluation of 193-nm and EUV photoresists that utilize a higher-order reaction mechanism of deprotection is presented. Unique polymers utilize novel blocking groups that require two acid-catalyzed steps to be removed. When these steps occur with comparable reaction rates, the overall reaction can be higher order (<= 1.85). The LWR of these resists is plotted against PEB time for a variety of compounds to acquire insight into the effectiveness of the proposed higher-order mechanisms. Evidence acquired during testing of these novel photoresist materials supports the conclusion that higher-order reaction kinetics leads to improved LWR vs. control resists.

The line roughness improvement with plasma coating and cure treatment for 193nm lithography and beyond

NASA Astrophysics Data System (ADS)

Zheng, Erhu; Huang, Yi; Zhang, Haiyang

2017-03-01

As CMOS technology reaches 14nm node and beyond, one of the key challenges of the extension of 193nm immersion lithography is how to control the line edge and width roughness (LER/LWR). For Self-aligned Multiple Patterning (SaMP), LER becomes larger while LWR becomes smaller as the process proceeds[1]. It means plasma etch process becomes more and more dominant for LER reduction. In this work, we mainly focus on the core etch solution including an extra plasma coating process introduced before the bottom anti reflective coating (BARC) open step, and an extra plasma cure process applied right after BARC-open step. Firstly, we leveraged the optimal design experiment (ODE) to investigate the impact of plasma coating step on LER and identified the optimal condition. ODE is an appropriate method for the screening experiments of non-linear parameters in dynamic process models, especially for high-cost-intensive industry [2]. Finally, we obtained the proper plasma coating treatment condition that has been proven to achieve 32% LER improvement compared with standard process. Furthermore, the plasma cure scheme has been also optimized with ODE method to cover the LWR degradation induced by plasma coating treatment.

Ab-initio study of C and O impurities in uranium nitride

NASA Astrophysics Data System (ADS)

Lopes, Denise Adorno; Claisse, Antoine; Olsson, Pär

2016-09-01

Uranium nitride (UN) has been considered a potential fuel for Generation IV (GEN-IV) nuclear reactors as well as a possible new fuel for Light Water Reactors (LWR), which would permit an extension of the fuel residence time in the reactor. Carbon and oxygen impurities play a key role in the UN microstructure, influencing important parameters such as creep, swelling, gas release under irradiation, compatibility with structural steel and coolants, and thermal stability. In this work, a systematic study of the electronic structure of UN containing C and O impurities using first-principles calculations by the Density Functional Theory (DFT) method is presented. In order to describe accurately the localized U 5f electrons, the DFT + U formalism was adopted. Moreover, to avoid convergence toward metastable states, the Occupation Matrix Control (OMC) methodology was applied. The incorporation of C and O in the N-vacancy is found to be energetically favorable. In addition, only for O, the incorporation in the interstitial position is energetically possible, showing some degree of solubility for this element in this site. The binding energies show that the pairs (Csbnd Nvac) and (Osbnd Nvac) interact much further than the other defects, which indicate the possible occurrence of vacancy drag phenomena and clustering of these impurities in grain boundaries, dislocations and free surfaces. The migration energy of an impurity by single N-vacancy show that C and O employ different paths during diffusion. Oxygen migration requires significantly lower energy than carbon. This fact is due to flexibility in the Usbnd O chemical bonds, which bend during the diffusion forming a pseudo UO2 coordination. On the other hand, C and N have a directional and inflexible chemical bond with uranium; always requiring the octahedral coordination. These findings provide detailed insight into how these impurities behave in the UN matrix, and can be of great interest for assisting the development of this new nuclear fuel for next-generation reactors.

LARGE SCALE METHOD FOR THE PRODUCTION AND PURIFICATION OF CURIUM

DOEpatents

Higgins, G.H.; Crane, W.W.T.

1959-05-19

A large-scale process for production and purification of Cm/sup 242/ is described. Aluminum slugs containing Am are irradiated and declad in a NaOH-- NaHO/sub 3/ solution at 85 to 100 deg C. The resulting slurry filtered and washed with NaOH, NH/sub 4/OH, and H/sub 2/O. Recovery of Cm from filtrate and washings is effected by an Fe(OH)/sub 3/ precipitation. The precipitates are then combined and dissolved ln HCl and refractory oxides centrifuged out. These oxides are then fused with Na/sub 2/CO/sub 3/ and dissolved in HCl. The solution is evaporated and LiCl solution added. The Cm, rare earths, and anionic impurities are adsorbed on a strong-base anfon exchange resin. Impurities are eluted with LiCl--HCl solution, rare earths and Cm are eluted by HCl. Other ion exchange steps further purify the Cm. The Cm is then precipitated as fluoride and used in this form or further purified and processed. (T.R.H.)

User Guide for VISION 3.4.7 (Verifiable Fuel Cycle Simulation) Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

2011-07-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters and options; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating 'what if' scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level. The model is not intended as amore » tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., 'reactor types' not individual reactors and 'separation types' not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation or disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. You must use Powersim Studio 8 or better. We have tested VISION with the Studio 8 Expert, Executive, and Education versions. The Expert and Education versions work with the number of reactor types of 3 or less. For more reactor types, the Executive version is currently required. The input files are Excel2003 format (xls). The output files are macro-enabled Excel2007 format (xlsm). VISION 3.4 was designed with more flexibility than previous versions, which were structured for only three reactor types - LWRs that can use only uranium oxide (UOX) fuel, LWRs that can use multiple fuel types (LWR MF), and fast reactors. One could not have, for example, two types of fast reactors concurrently. The new version allows 10 reactor types and any user-defined uranium-plutonium fuel is allowed. (Thorium-based fuels can be input but several features of the model would not work.) The user identifies (by year) the primary fuel to be used for each reactor type. The user can identify for each primary fuel a contingent fuel to use if the primary fuel is not available, e.g., a reactor designated as using mixed oxide fuel (MOX) would have UOX as the contingent fuel. Another example is that a fast reactor using recycled transuranic (TRU) material can be designated as either having or not having appropriately enriched uranium oxide as a contingent fuel. Because of the need to study evolution in recycling and separation strategies, the user can now select the recycling strategy and separation technology, by year.« less

Implications of Fast Reactor Transuranic Conversion Ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven J. Piet; Edward A. Hoffman; Samuel E. Bays

2010-11-01

Theoretically, the transuranic conversion ratio (CR), i.e. the transuranic production divided by transuranic destruction, in a fast reactor can range from near zero to about 1.9, which is the average neutron yield from Pu239 minus 1. In practice, the possible range will be somewhat less. We have studied the implications of transuranic conversion ratio of 0.0 to 1.7 using the fresh and discharge fuel compositions calculated elsewhere. The corresponding fissile breeding ratio ranges from 0.2 to 1.6. The cases below CR=1 (“burners”) do not have blankets; the cases above CR=1 (“breeders”) have breeding blankets. The burnup was allowed to floatmore » while holding the maximum fluence to the cladding constant. We graph the fuel burnup and composition change. As a function of transuranic conversion ratio, we calculate and graph the heat, gamma, and neutron emission of fresh fuel; whether the material is “attractive” for direct weapon use using published criteria; the uranium utilization and rate of consumption of natural uranium; and the long-term radiotoxicity after fuel discharge. For context, other cases and analyses are included, primarily once-through light water reactor (LWR) uranium oxide fuel at 51 MWth-day/kg-iHM burnup (UOX-51). For CR<1, the heat, gamma, and neutron emission increase as material is recycled. The uranium utilization is at or below 1%, just as it is in thermal reactors as both types of reactors require continuing fissile support. For CR>1, heat, gamma, and neutron emission decrease with recycling. The uranium utilization exceeds 1%, especially as all the transuranic elements are recycled. exceeds 1%, especially as all the transuranic elements are recycled. At the system equilibrium, heat and gamma vary by somewhat over an order of magnitude as a function of CR. Isotopes that dominate heat and gamma emission are scattered throughout the actinide chain, so the modest impact of CR is unsurprising. Neutron emitters are preferentially found among the higher actinides, so the neutron emission varies much stronger with CR, about three orders of magnitude.« less

Summary of the Advanced Reactor Design Criteria (ARDC) Phase 2 Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holbrook, Mark Raymond

This report provides an end-of-year summary reflecting the progress and status of proposed regulatory design criteria for advanced non-LWR designs in accordance with the Level 3 milestone in M3AT-15IN2001017 in work package AT-15IN200101. These criteria have been designated as ARDC, and they provide guidance to future applicants for addressing the GDC that are currently applied specifically to LWR designs. The report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of example adaptations of ARDC for Sodium Fast Reactor (SFR) and modular High Temperature Gas-cooled Reactor (HTGR) designs.

Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

NASA Astrophysics Data System (ADS)

Fratoni, Massimiliano

This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be transmuted; this burnup is slightly superior to that attainable in helium-cooled reactors. A preliminary analysis of the modular variant for the PB-AHTR investigated the triple heterogeneity of this design and determined its performance characteristics.

Long Wavelength Ripples in the Nearshore

NASA Astrophysics Data System (ADS)

Alcinov, T.; Hay, A. E.

2008-12-01

Sediment bedforms are ubiquitous in the nearshore environment, and their characteristics and evolution have a direct effect on the hydrodynamics and the rate of sediment transport. The focus of this study is long wavelength ripples (LWR) observed at two locations in the nearshore at roughly 3m water depth under combined current and wave conditions in Duck, North Carolina. LWR are straight-crested bedforms with wavelengths in the range of 20-200cm, and steepness of about 0.1. They occur in the build up and decay of storms, in a broader range of values of the flow parameters compared to other ripple types. The main goal of the study is to test the maximum gross bedform-normal transport (mGBNT) hypothesis, which states that the orientation of ripples in directionally varying flows is such that the gross sediment transport normal to the ripple crest is maximized. Ripple wavelengths and orientation are measured from rotary fanbeam images and current and wave conditions are obtained from electromagnetic (EM) flowmeters and an offshore pressure gauge array. Preliminary tests in which transport direction is estimated from the combined flow velocity vectors indicate that the mGBNT is not a good predictor of LWR orientation. Results from tests of the mGBNT hypothesis using a sediment transport model will be presented.

Regulatory Risk Reduction for Advanced Reactor Technologies – FY2016 Status and Work Plan Summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moe, Wayne Leland

2016-08-01

Millions of public and private sector dollars have been invested over recent decades to realize greater efficiency, reliability, and the inherent and passive safety offered by advanced nuclear reactor technologies. However, a major challenge in experiencing those benefits resides in the existing U.S. regulatory framework. This framework governs all commercial nuclear plant construction, operations, and safety issues and is highly large light water reactor (LWR) technology centric. The framework must be modernized to effectively deal with non-LWR advanced designs if those designs are to become part of the U.S energy supply. The U.S. Department of Energy’s (DOE) Advanced Reactor Technologiesmore » (ART) Regulatory Risk Reduction (RRR) initiative, managed by the Regulatory Affairs Department at the Idaho National Laboratory (INL), is establishing a capability that can systematically retire extraneous licensing risks associated with regulatory framework incompatibilities. This capability proposes to rely heavily on the perspectives of the affected regulated community (i.e., commercial advanced reactor designers/vendors and prospective owner/operators) yet remain tuned to assuring public safety and acceptability by regulators responsible for license issuance. The extent to which broad industry perspectives are being incorporated into the proposed framework makes this initiative unique and of potential benefit to all future domestic non-LWR applicants« less

Impact of implicit effects on uncertainties and sensitivities of the Doppler coefficient of a LWR pin cell

NASA Astrophysics Data System (ADS)

Hursin, Mathieu; Leray, Olivier; Perret, Gregory; Pautz, Andreas; Bostelmann, Friederike; Aures, Alexander; Zwermann, Winfried

2017-09-01

In the present work, PSI and GRS sensitivity analysis (SA) and uncertainty quantification (UQ) methods, SHARK-X and XSUSA respectively, are compared for reactivity coefficient calculation; for reference the results of the TSUNAMI and SAMPLER modules of the SCALE code package are also provided. The main objective of paper is to assess the impact of the implicit effect, e.g., considering the effect of cross section perturbation on the self-shielding calculation, on the Doppler coefficient SA and UQ. Analyses are done for a Light Water Reactor (LWR) pin cell based on Phase I of the UAM LWR benchmark. The negligence of implicit effects in XSUSA and TSUNAMI leads to deviations of a few percent between the sensitivity profiles compared to SAMPLER and TSUNAMI (incl. implicit effects) except for 238U elastic scattering. The implicit effect is much larger for the SHARK-X calculations because of its coarser energy group structure between 10 eV and 10 keV compared to the applied SCALE libraries. It is concluded that the influence of the implicit effect strongly depends on the energy mesh of the nuclear data library of the neutron transport solver involved in the UQ calculations and may be magnified by the response considered.

A fracture mechanics approach for estimating fatigue crack initiation in carbon and low-alloy steels in LWR coolant environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Park, H. B.; Chopra, O. K.

2000-04-10

A fracture mechanics approach for elastic-plastic materials has been used to evaluate the effects of light water reactor (LWR) coolant environments on the fatigue lives of carbon and low-alloy steels. The fatigue life of such steel, defined as the number of cycles required to form an engineering-size crack, i.e., 3-mm deep, is considered to be composed of the growth of (a) microstructurally small cracks and (b) mechanically small cracks. The growth of the latter was characterized in terms of {Delta}J and crack growth rate (da/dN) data in air and LWR environments; in water, the growth rates from long crack testsmore » had to be decreased to match the rates from fatigue S-N data. The growth of microstructurally small cracks was expressed by a modified Hobson relationship in air and by a slip dissolution/oxidation model in water. The crack length for transition from a microstructurally small crack to a mechanically small crack was based on studies on small crack growth. The estimated fatigue S-N curves show good agreement with the experimental data for these steels in air and water environments. At low strain amplitudes, the predicted lives in water can be significantly lower than the experimental values.« less

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

DOE PAGES

Chopra, O. K.; Stevens, G. L.; Tregoning, R.; ...

2017-10-06

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewheremore » since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en, which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.« less

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chopra, O. K.; Stevens, G. L.; Tregoning, R.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewheremore » since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en, which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.« less

Research Program of a Super Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

Development of a new lattice physics code robin for PWR application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, S.; Chen, G.

2013-07-01

This paper presents a description of methodologies and preliminary verification results of a new lattice physics code ROBIN, being developed for PWR application at Shanghai NuStar Nuclear Power Technology Co., Ltd. The methods used in ROBIN to fulfill various tasks of lattice physics analysis are an integration of historical methods and new methods that came into being very recently. Not only these methods like equivalence theory for resonance treatment and method of characteristics for neutron transport calculation are adopted, as they are applied in many of today's production-level LWR lattice codes, but also very useful new methods like the enhancedmore » neutron current method for Dancoff correction in large and complicated geometry and the log linear rate constant power depletion method for Gd-bearing fuel are implemented in the code. A small sample of verification results are provided to illustrate the type of accuracy achievable using ROBIN. It is demonstrated that ROBIN is capable of satisfying most of the needs for PWR lattice analysis and has the potential to become a production quality code in the future. (authors)« less

SiC/SiC Cladding Materials Properties Handbook

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snead, Mary A.; Katoh, Yutai; Koyanagi, Takaaki

When a new class of material is considered for a nuclear core structure, the in-pile performance is usually assessed based on multi-physics modeling in coordination with experiments. This report aims to provide data for the mechanical and physical properties and environmental resistance of silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites for use in modeling for their application as accidenttolerant fuel cladding for light water reactors (LWRs). The properties are specific for tube geometry, although many properties can be predicted from planar specimen data. This report presents various properties, including mechanical properties, thermal properties, chemical stability under normal and offnormalmore » operation conditions, hermeticity, and irradiation resistance. Table S.1 summarizes those properties mainly for nuclear-grade SiC/SiC composites fabricated via chemical vapor infiltration (CVI). While most of the important properties are available, this work found that data for the in-pile hydrothermal corrosion resistance of SiC materials and for thermal properties of tube materials are lacking for evaluation of SiC-based cladding for LWR applications.« less

Multi-Group Formulation of the Temperature-Dependent Resonance Scattering Model and its Impact on Reactor Core Parameters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ghrayeb, Shadi Z.; Ougouag, Abderrafi M.; Ouisloumen, Mohamed

2014-01-01

A multi-group formulation for the exact neutron elastic scattering kernel is developed. It incorporates the neutron up-scattering effects, stemming from lattice atoms thermal motion and accounts for it within the resulting effective nuclear cross-section data. The effects pertain essentially to resonant scattering off of heavy nuclei. The formulation, implemented into a standalone code, produces effective nuclear scattering data that are then supplied directly into the DRAGON lattice physics code where the effects on Doppler Reactivity and neutron flux are demonstrated. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering,more » which in turn affect the estimation of core reactivity and burnup characteristics. The results show an increase in values of Doppler temperature feedback coefficients up to -10% for UOX and MOX LWR fuels compared to the corresponding values derived using the traditional asymptotic elastic scattering kernel. This paper also summarizes the results done on this topic to date.« less

Core characterization of the new CABRI Water Loop Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ritter, G.; Rodiac, F.; Beretz, D.

2011-07-01

The CABRI experimental reactor is located at the Cadarache nuclear research center, southern France. It is operated by the Atomic Energy Commission (CEA) and devoted to IRSN (Institut de Radioprotection et de Surete Nucleaire) safety programmes. It has been successfully operated during the last 30 years, enlightening the knowledge of FBR and LWR fuel behaviour during Reactivity Insertion Accident (RIA) and Loss Of Coolant Accident (LOCA) transients in the frame of IPSN (Institut de Protection et de Surete Nucleaire) and now IRSN programmes devoted to reactor safety. This operation was interrupted in 2003 to allow for a whole facility renewalmore » programme for the need of the CABRI International Programme (CIP) carried out by IRSN under the OECD umbrella. The principle of operation of the facility is based on the control of {sup 3}He, a major gaseous neutron absorber, in the core geometry. The purpose of this paper is to illustrate how several dosimetric devices have been set up to better characterize the core during the upcoming commissioning campaign. It presents the schemes and tools dedicated to core characterization. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kohnert, Aaron A.; Dasgupta, Dwaipayan; Wirth, Brian

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensuremore » that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior.« less

Corrosion and Corrosion Control in Light Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry M.

2013-08-01

Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

Down-selection of candidate alloys for further testing of advanced replacement materials for LWR core internals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Was, Gary; Leonard, Keith J.; Tan, Lizhen

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) Light Water Reactor Sustainability Program to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to identify and develop advanced alloys with superiormore » degradation resistance in light water reactor (LWR)-relevant environments by 2024.« less

Prototype Demonstration of Gamma- Blind Tensioned Metastable Fluid Neutron/Multiplicity/Alpha Detector – Real Time Methods for Advanced Fuel Cycle Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDeavitt, Sean M.

The content of this report summarizes a multi-year effort to develop prototype detection equipment using the Tensioned Metastable Fluid Detector (TMFD) technology developed by Taleyarkhan [1]. The context of this development effort was to create new methods for evaluating and developing advanced methods for safeguarding nuclear materials along with instrumentation in various stages of the fuel cycle, especially in material balance areas (MBAs) and during reprocessing of used nuclear fuel. One of the challenges related to the implementation of any type of MBA and/or reprocessing technology (e.g., PUREX or UREX) is the real-time quantification and control of the transuranic (TRU)more » isotopes as they move through the process. Monitoring of higher actinides from their neutron emission (including multiplicity) and alpha signatures during transit in MBAs and in aqueous separations is a critical research area. By providing on-line real-time materials accountability, diversion of the materials becomes much more difficult. The Tensioned Metastable Fluid Detector (TMFD) is a transformational technology that is uniquely capable of both alpha and neutron spectroscopy while being “blind” to the intense gamma field that typically accompanies used fuel – simultaneously with the ability to provide multiplicity information as well [1-3]. The TMFD technology was proven (lab-scale) as part of a 2008 NERI-C program [1-7]. The bulk of this report describes the advancements and demonstrations made in TMFD technology. One final point to present before turning to the TMFD demonstrations is the context for discussing real-time monitoring of SNM. It is useful to review the spectrum of isotopes generated within nuclear fuel during reactor operations. Used nuclear fuel (UNF) from a light water reactor (LWR) contains fission products as well as TRU elements formed through neutron absorption/decay chains. The majority of the fission products are gamma and beta emitters and they represent the more significant hazards from a radiation protection standpoint. However, alpha and neutron emitting uranium and TRU elements represent the more significant safeguards and security concerns. Table 1.1 presents a representative PWR inventory of the uranium and actinide isotopes present in a used fuel assembly. The uranium and actinide isotopes (chiefly the Pu, Am and Cm elements) are all emitters of alpha particles and some of them release significant quantities of neutrons through spontaneous fissions« less

Towards a Consolidated Approach for the Assessment of Evaluation Models of Nuclear Power Reactors

DOE PAGES

Epiney, A.; Canepa, S.; Zerkak, O.; ...

2016-11-02

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

Advanced Neutronics Tools for BWR Design Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santamarina, A.; Hfaiedh, N.; Letellier, R.

2006-07-01

This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007more » BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)« less

Sensitivity enhancement of the high-resolution xMT multi-trigger resist for EUV lithography

NASA Astrophysics Data System (ADS)

Popescu, Carmen; Frommhold, Andreas; McClelland, Alexandra; Roth, John; Ekinci, Yasin; Robinson, Alex P. G.

2017-03-01

Irresistible Materials is developing a new molecular resist system that demonstrates high-resolution capability based on the multi-trigger concept. A series of studies such as resist purification, developer choice,and enhanced resist crosslinking were conducted in order to optimize the performance of this material. The optimized conditions allowed patterning 14 nm half-pitch (hp) lines with a line width roughness (LWR) of 2.7 nm at the XIL beamline of the Swiss Light source. Furthermore it was possible to pattern 14 nm hp features with dose of 14 mJ/cm2 with an LWR of 4.9 nm. We have also begun to investigate the addition of high-Z additives to EUV photoresist as a means to increase sensitivity and modify secondary electron blur.

Integral Inherently Safe Light Water Reactor (I 2S-LWR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petrovic, Bojan; Memmott, Matthew; Boy, Guy

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I 2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to addressmore » the preference of some utilities in the US power market for unit power level on the order of 1 GWe.« less

Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

NASA Astrophysics Data System (ADS)

Burdo, James S.

This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the difference was less than 1.00, allowing for the existence of the difference within the margin of error. The second was whether the difference between the two values was big enough to prevent their error bars from overlapping. Error analysis was performed both using a one second count and pseudo-Maxwell statistics for a projected 60 second count, giving four criteria for detection. The number of guide tubes meeting these criteria was compared and graphed for each case. Further analysis at extremes of high and low enrichment and long and short burnup times was done using data from assemblies at the Beaver Valley 1 and 2 PWR. In all neutron flux cases, at least two guide tube locations meet all the criteria for detection of pin diversion. At least one location in almost all of the gamma flux cases does. These results show that placing detectors in the empty guide tubes of spent fuel bundles to identify possible pin diversion is feasible.

Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

NASA Astrophysics Data System (ADS)

Carroll, Spencer

As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel performance code developed by PNNL and used by the Nuclear Regulatory Commission (NRC) as a licensing code for US reactors. FRAPCON will give insight into how these new fuel-cladding combinations will affect cladding hoop stress and help determine if the new materials are feasible for use in a reactor. To accurately simulate the interaction between the new materials, a soft pellet model that allows for stresses on the pellet to affect pellet deformation will have to be implemented. Currently, FRAPCON uses a rigid pellet model that does not allow for feedback of the cladding onto the pellet. Since SiC does not creep at the temperatures being considered and is not ductile, any PCMI create a much higher interfacial pressure than is possible with Zircaloy. Because of this, it is necessary to implement a model that allows for pellet creep to alleviate some of these cladding stresses. These results will then be compared to FEMAXI-6, a Japanese fuel performance code that already calculates pellet stress and allows for cladding feedback onto the pellet. This research is intended to be a continuation and verification of previous work done by USC on the analysis of accident tolerant fuels with alternative claddings and is intended to prove that a soft pellet model is necessary to accurately model any fuel with SiC cladding.

Optimization of Deep Borehole Systems for HLW Disposal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Driscoll, Michael; Baglietto, Emilio; Buongiorno, Jacopo

2015-09-09

This is the final report on a project to update and improve the conceptual design of deep boreholes for high level nuclear waste disposal. The effort was concentrated on application to intact US legacy LWR fuel assemblies, but conducted in a way in which straightforward extension to other waste forms, host rock types and countries was preserved. The reference fuel design version consists of a vertical borehole drilled into granitic bedrock, with the uppermost kilometer serving as a caprock zone containing a diverse and redundant series of plugs. There follows a one to two kilometer waste canister emplacement zone havingmore » a hole diameter of approximately 40-50 cm. Individual holes are spaced 200-300 m apart to form a repository field. The choice of verticality and the use of a graphite based mud as filler between the waste canisters and the borehole wall liner was strongly influenced by the expectation that retrievability would continue to be emphasized in US and worldwide repository regulatory criteria. An advanced version was scoped out using zinc alloy cast in place to fill void space inside a disposal canister and its encapsulated fuel assembly. This excludes water and greatly improves both crush resistance and thermal conductivity. However the simpler option of using a sand fill was found adequate and is recommended for near-term use. Thermal-hydraulic modeling of the low permeability and porosity host rock and its small (≤ 1%) saline water content showed that vertical convection induced by the waste’s decay heat should not transport nuclides from the emplacement zone up to the biosphere atop the caprock. First order economic analysis indicated that borehole repositories should be cost-competitive with shallower mined repositories. It is concluded that proceeding with plans to drill a demonstration borehole to confirm expectations, and to carry out priority experiments, such as retention and replenishment of in-hole water is in order.« less

Steam Oxidation Testing in the Severe Accident Test Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pint, Bruce A.; McMurray, Jake W.

2016-08-01

Since 2011, Oak Ridge National Laboratory (ORNL) has been conducting high temperature steam oxidation testing of candidate alloys for accident tolerant fuel (ATF) cladding. These concepts are designed to enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the US ATF community, the Severe Accident Test Station (SATS) has been evaluating candidate materials (including coatings) since 2012. Compared to the current UO 2/Zr-based alloy fuel system, alternative cladding materials need to offer slower oxidation kinetics and a smaller enthalpy of oxidation in order to significantly reduce the rate of heat and hydrogen generation in the coremore » during a coolant-limited severe accident. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models. However, prior modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. In some cases, the results have been difficult to interpret and more fundamental information is needed such as the stability of alumina in flowing steam at 1400°-1500°C. This report summarizes recent work to measure the steam oxidation kinetics of candidate alloys, the evaporation rate of alumina in steam and the development of integral data on FeCrAl compared to conventional Zr-based cladding.« less

Evaluation of steam corrosion and water quenching behavior of zirconium-silicide coated LWR fuel claddings

NASA Astrophysics Data System (ADS)

Yeom, Hwasung; Lockhart, Cody; Mariani, Robert; Xu, Peng; Corradini, Michael; Sridharan, Kumar

2018-02-01

This study investigates steam corrosion of bulk ZrSi2, pure Si, and zirconium-silicide coatings as well as water quenching behavior of ZrSi2 coatings to evaluate its feasibility as a potential accident-tolerant fuel cladding coating material in light water nuclear reactor. The ZrSi2 coating and Zr2Si-ZrSi2 coating were deposited on Zircaloy-4 flats, SiC flats, and cylindrical Zircaloy-4 rodlets using magnetron sputter deposition. Bulk ZrSi2 and pure Si samples showed weight loss after the corrosion test in pure steam at 400 °C and 10.3 MPa for 72 h. Silicon depletion on the ZrSi2 surface during the steam test was related to the surface recession observed in the silicon samples. ZrSi2 coating (∼3.9 μm) pre-oxidized in 700 °C air prevented substrate oxidation but thin porous ZrO2 formed on the coating. The only condition which achieved complete silicon immobilization in the oxide scale in aqueous environments was the formation of ZrSiO4 via ZrSi2 coating oxidation in 1400 °C air. In addition, ZrSi2 coatings were beneficial in enhancing quenching heat transfer - the minimum film boiling temperature increased by 6-8% in the three different environmental conditions tested. During repeated thermal cycles (water quenching from 700 °C to 85 °C for 20 s) performed as a part of quench tests, no spallation and cracking was observed and the coating prevented oxidation of the underlying Zircaloy-4 substrate.

Role of ADS in the back-end of the fuel cycle strategies and associated design activities: The case of Japan

NASA Astrophysics Data System (ADS)

Oigawa, Hiroyuki; Tsujimoto, Kazufumi; Nishihara, Kenji; Sugawara, Takanori; Kurata, Yuji; Takei, Hayanori; Saito, Shigeru; Sasa, Toshinobu; Obayashi, Hironari

2011-08-01

Reduction of burden caused by radioactive waste management is one of the most critical issues for the sustainable utilization of nuclear power. The Partitioning and Transmutation (P&T) technology provides the possibility to reduce the amount of the radiotoxic inventory of the high-level radioactive waste (HLW) dramatically and to extend the repository capacity. The accelerator-driven system (ADS) is regarded as a powerful tool to effectively transmute minor actinides (MAs) in the "double-strata" fuel cycle strategy. The ADS has a potential to flexibly manage MA in the transient phase from light water reactors (LWRs) to fast breeder reactors (FBRs), and can co-exist with FBR symbiotically and complementarily to enhance the reliability and the safety of the commercial FBR cycle. The concept of ADS in JAEA is a lead-bismuth eutectic (LBE) cooled, tank-type subcritical reactor with the power of 800 MWth driven by a 30 MW superconducting LINAC. By such an ADS, 250 kg of MA can be transmuted annually, which corresponds to the amount of MA produced in 10 units of LWR with 1 GWe. The design study was performed mainly for the subcritical reactor and the spallation target with a beam window. In Japan, Atomic Energy Commission (AEC) has implemented the check and review (C&R) on P&T technology from 2008 to 2009. In the C&R, the benefit of P&T technology, the current status of the R&D, and the way forward to promote it were discussed.

Thermal Aging Phenomena in Cast Duplex Stainless Steels

DOE PAGES

Byun, T. S.; Yang, Y.; Overman, N. R.; ...

2015-11-12

We used cast stainless steels (CASSs)for the large components of light water reactor (LWR) power plants such as primary coolant piping and pump casing. The thermal embrittlement of CASS components is one of the most serious concerns related to the extended-term operation of nuclear power plants. Many past researches have concluded that the formation of Cr-rich alpha-phase by Spinodal decomposition of delta-ferrite phase is the primary mechanism for the thermal embrittlement. Cracking mechanism in the thermally-embrittled duplex stainless steels consists of the formation of cleavage at ferrite and its propagation via separation of ferrite-austenite interphase. This article intends to providemore » an introductory overview on the thermal aging phenomena in LWR-relevant conditions. Firstly, the thermal aging effect on toughness is discussed in terms of the cause of embrittlement and influential parameters. Moreover, an approximate analysis of thermal reaction using Arrhenius equation was carried out to scope the aging temperatures for the accelerated aging experiments to simulate the 60 and 80 years of services. Further, an equilibrium precipitation calculation was performed for model CASS alloys using the CALPHAD program, and the results are used to describe the precipitation behaviors in duplex stainless steels. Our results are also to be used to guide an on-going research aiming to provide knowledge-based conclusive prediction for the integrity of the CASS components of LWR power plants during the service life extended up to and beyond 60 years.« less

Thermal Aging Phenomena in Cast Duplex Stainless Steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Byun, T. S.; Yang, Y.; Overman, N. R.

Cast stainless steels (CASSs) have been extensively used for the large components of light water reactor (LWR) power plants such as primary coolant piping and pump casing. The thermal embrittlement of CASS components is one of the most serious concerns related to the extended-term operation of nuclear power plants. Many past researches have concluded that the formation of Cr–rich α'-phase by Spinodal decomposition of δ-ferrite phase is the primary mechanism for the thermal embrittlement. Cracking mechanism in the thermally-embrittled duplex stainless steels consists of the formation of cleavage at ferrite and its propagation via separation of ferrite-austenite interphase. This articlemore » intends to provide an introductory overview on the thermal aging phenomena in LWR relevant conditions. Firstly, the thermal aging effect on toughness is discussed in terms of the cause of embrittlement and influential parameters. An approximate analysis of thermal reaction using Arrhenius equation was carried out to scope the aging temperatures for the accelerated aging experiments to simulate the 60 and 80 years of services. Further, equilibrium precipitation calculation was performed for model CASS alloys using the CALPHAD program and the results are used to describe the precipitation behaviors in duplex stainless steels. These results are also to be used to guide an on-going research aiming to provide knowledge-based conclusive prediction for the integrity of the CASS components of LWR power plants during the service life extended up to and beyond 60 years.« less

Line-edge roughness performance targets for EUV lithography

NASA Astrophysics Data System (ADS)

Brunner, Timothy A.; Chen, Xuemei; Gabor, Allen; Higgins, Craig; Sun, Lei; Mack, Chris A.

2017-03-01

Our paper will use stochastic simulations to explore how EUV pattern roughness can cause device failure through rare events, so-called "black swans". We examine the impact of stochastic noise on the yield of simple wiring patterns with 36nm pitch, corresponding to 7nm node logic, using a local Critical Dimension (CD)-based fail criteria Contact hole failures are examined in a similar way. For our nominal EUV process, local CD uniformity variation and local Pattern Placement Error variation was observed, but no pattern failures were seen in the modest (few thousand) number of features simulated. We degraded the image quality by incorporating Moving Standard Deviation (MSD) blurring to degrade the Image Log-Slope (ILS), and were able to find conditions where pattern failures were observed. We determined the Line Width Roughness (LWR) value as a function of the ILS. By use of an artificial "step function" image degraded by various MSD blur, we were able to extend the LWR vs ILS curve into regimes that might be available for future EUV imagery. As we decreased the image quality, we observed LWR grow and also began to see pattern failures. For high image quality, we saw CD distributions that were symmetrical and close to Gaussian in shape. Lower image quality caused CD distributions that were asymmetric, with "fat tails" on the low CD side (under-exposed) which were associated with pattern failures. Similar non-Gaussian CD distributions were associated with image conditions that caused missing contact holes, i.e. CD=0.

Study on bubbly flow behavior in natural circulation reactor by thermal-hydraulic simulation tests with SF6-Gas and ethanol liquid

NASA Astrophysics Data System (ADS)

Kondo, Yoshiyuki; Suga, Keishi; Hibi, Koki; Okazaki, Toshihiko; Komeno, Toshihiro; Kunugi, Tomoaki; Serizawa, Akimi; Yoneda, Kimitoshi; Arai, Takahiro

2009-02-01

An advanced experimental technique has been developed to simulate two-phase flow behavior in a light water reactor (LWR). The technique applies three kinds of methods; (1) use of sulfur-hexafluoride (SF6) gas and ethanol (C2H5OH) liquid at atmospheric temperature and a pressure less than 1.0MPa, where the fluid properties are similar to steam-water ones in the LWR, (2) generation of bubble with a sintering tube, which simulates bubble generation on heated surface in the LWR, (3) measurement of detailed bubble distribution data with a bi-optical probe (BOP), (4) and measurement of liquid velocities with the tracer liquid. This experimental technique provides easy visualization of flows by using a large scale experimental apparatus, which gives three-dimensional flows, and measurement of detailed spatial distributions of two-phase flow. With this technique, we have carried out experiments simulating two-phase flow behavior in a single-channel geometry, a multi-rod-bundle one, and a horizontal-tube-bundle one on a typical natural circulation reactor system. Those experiments have clarified a) a flow regime map in a rod bundle on the transient region between bubbly and churn flow, b) three-dimensional flow behaviour in rod-bundles where inter-subassembly cross-flow occurs, c) bubble-separation behavior with consideration of reactor internal structures. The data have given analysis models for the natural circulation reactor design with good extrapolation.

NRC ARDC Guidance Support Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holbrook, Mark R.

This report provides a summary that reflects the progress and status of proposed regulatory design criteria for advanced non-light water reactor (LWR) designs in accordance with the Level 3 milestone M3AT-17IN2001013 in work package AT-17IN200101. These criteria have been designated as advanced reactor design criteria (ARDC) and they provide guidance to future applicants for addressing the general design criteria (GDC) that are currently applied specifically to LWR designs. This report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of ARDC regulatory guidance for sodium fast reactor (SFR) andmore » modular high-temperature gas-cooled reactor (HTGR) designs. Status Report Organization: Section 2 discusses the origin of the GDC and their application to LWRs. Section 3 addresses the objective of this initiative and how it benefits the advanced non-LWR reactor vendors. Section 4 discusses the scope and structure of the initiative. Section 5 provides background on the U.S. Department of Energy (DOE) ARDC team’s original development of the proposed ARDC that were submitted to the NRC for consideration. Section 6 provides a summary of recent ARDC Phase 2 activities. Appendices A through E document the DOE ARDC team’s public comments on various sections of the NRC’s draft regulatory guide DG–1330, “Guidance for Developing Principal Design Criteria for Non-Light Water Reactors.”« less

Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding

DOE PAGES

Besmann, T. M.; Yamamoto, Y.; Unocic, K. A.

2016-06-21

We assessed the thermochemical compatibility of potential replacement cladding materials for zirconium alloys in light water reactors. Considered were FeCrAl steel (similar to Kanthal APMT), Nb-1%Zr (similar to PWC-11), and a hybrid SiC-composite with a metallic barrier layer. The niobium alloy was also seen as requiring an oxidation protective layer, and a diffusion silicide was investigated. Metallic barrier layers for the SiC-composite reviewed included a FeCrAl alloy, Nb-1%Zr, and chromium. Thermochemical calculations were performed to determine oxidation behavior of the materials in steam, and for hybrid SiC-composites possible interactions between the metallic layer and SiC. Additionally, experimental exposures of SiC-alloymore » reaction couples at 673K, 1073K, and 1273K for 168 h in an inert atmosphere were made and microanalysis performed. Whereas all materials were determined to oxidize under higher oxygen partial pressures in the steam environment, these varied by material with expected protective oxides forming. Finally, the computed and experimental results indicate the formation of liquid phase eutectic in the FeCrAl-SiC system at the higher temperatures.« less

Opportunities for the LWR ATF materials development program to contribute to the LBE-cooled ADS materials qualification program

NASA Astrophysics Data System (ADS)

Gong, Xing; Li, Rui; Sun, Maozhou; Ren, Qisen; Liu, Tong; Short, Michael P.

2016-12-01

Accelerator-driven systems (ADS) are a promising approach for nuclear waste disposal. Nevertheless, the principal candidate materials proposed for ADS construction, such as the ferritic/martensitic steel, T91, and austenitic stainless steels, 316L and 15-15Ti, are not fully compatible with the liquid lead-bismuth eutectic (LBE) coolant. Under some operating conditions, liquid metal embrittlement (LME) or liquid metal corrosion (LMC) may occur in these steels when exposed to LBE. These environmentally-induced material degradation effects pose a threat to ADS reactor safety, as failure of the materials could initiate a severe accident, in which fission products are released into the coolant. Meanwhile, parallel efforts to develop accident-tolerant fuels (ATF) in light water reactors (LWRs) could provide both general materials design philosophies and specific material solutions to the ADS program. In this paper, the potential contributions of the ATF materials development program to the ADS materials qualification program are evaluated and discussed in terms of service conditions and materials performance requirements. Several specific areas where coordinated development may benefit both programs, including composite materials and selected coatings, are discussed.

Inclusion property of Cs, Sr, and Ba impurities in LiCl crystal formed by layer-melt crystallization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, Jung-Hoon; Cho, Yung-Zun; Lee, Tae-Kyo

Pyroprocessing is one of the promising technologies enabling the recycling of spent nuclear fuels from a commercial light water reactor (LWR). In general, pyroprocessing uses dry molten salts as electrolytes. In particular, LiCl waste salt after pyroprocessing contains highly radioactive I/II group fission products mainly composed of Cs, Sr, and Ba impurities. Therefore, it is beneficial to reuse LiCl salt in the pyroprocessing as an electrolyte for economic and environmental issues. Herein, to understand the inclusion property of impurities within LiCl crystal, the physical properties such as lattice parameter change, bulk modulus, and substitution enthalpy of a LiCl crystal havingmore » 0-6 at% Cs{sup +} or Ba{sup 2+} impurities under existence of 1 at% Sr{sup 2+} impurity were calculated via the first-principles density functional theory. The substitution enthalpy of LiCl crystals having 1 at% Sr{sup 2+} showed slightly decreased value than those without Sr{sup 2+} impurity. Therefore, through the substitution enthalpy calculation, it is expected that impurities will be incorporated within LiCl crystal as co-existed form rather than as a single component form. (authors)« less

The UO2 ex-ADU powder preparation and pellet sintering for optimum efficiency: experimental and modeling studies

NASA Astrophysics Data System (ADS)

Hung, Nguyen Trong; Thuan, Le Ba; Van Tung, Nguyen; Thuy, Nguyen Thanh; Lee, Jin-Young; Jyothi, Rajesh Kumar

2017-12-01

The UO2 nuclear fuel pellet process for light water reactors (LWR) includes the conversion of uranium hexafluoride (UF6) into UO2 powder and the fabrication of UO2 pellets from such UO2 powder. In the paper, studies on UO2 pellet process from ammonium diuranate-derived uranium dioxide powder (UO2 ex-ADU powder) were reported. The UO2 ex-ADU powders were converted from ADU at various temperatures of 973 K, 1023 K and 1073 K and then UO2 pellets prepared from the powders were sintered at temperatures of 1923 K, 1973 K and 2023 K for times of 4 h, 6 h and 8 h. Response surface methodology (RSM) based on quadratic central composite design (CCD) type of face centered (CCF) improved by Box and Hunter was used to model the UO2 pellet process, using MODDE 5.0 software as an assessing tool. On the base of the proposed model, the relationship between the technological parameters and density of the UO2 pellet product was suggested to control the UO2 ex-ADU pellet process as desired levels.

Experimental methodology of contact edge roughness on sub-100-nm pattern

NASA Astrophysics Data System (ADS)

Lee, Tae Yong; Ihm, Dongchul; Kang, Hyo Chun; Lee, Jun Bum; Lee, Byoung-Ho; Chin, Soo-Bok; Cho, Do-Hyun; Kim, Yang Hyong; Yang, Ho Dong; Yang, Kyoung Mo

2004-05-01

The measurement of edge roughness has become a hot issue in the semiconductor industry. Major vendors offer a variety of features to measure the edge roughness in their CD-SEMs. However, most of the features are limited by the applicable pattern types. For the line and space patterns, features such as Line Edge Roughness (LER) and Line Width Roughness (LWR) are available in current CD-SEMs. The edge roughness is more critical in contact process. However the measurement of contact edge roughness (CER) or contact space roughness (CSR) is more complicated than that of LER or LWR. So far, no formal standard measurement algorithm or definition of contact roughness measurement exists. In this article, currently available features are investigated to assess their representability for CER or CSR. Some new ideas to quantify CER and CSR were also suggested with preliminary experimental results.

Research, Development and Demonstration (RD&D) Needs for Light Water Reactor (LWR) Technologies A Report to the Reactor Technology Subcommittee of the Nuclear Energy Advisory Committee (NEAC) Office of Nuclear Energy U.S. Department of Energy

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCarthy, Kathryn A.; Adams, Bradley J.

The LWR RD&D Working Group developed a detailed list of RD&D suggestions and recommendations, which are provided in Appendix D. The Working Group then undertook a systematic ranking process, described in Appendix E. The results of the ranking process are not meant to be a strict set of priorities, but rather should provide insight into how the items generally ranked within the Working Group. Future discussions and investigation into these items could provide information that would support a change in these priorities or in their emphasis. The results of this prioritization are provided below. Note that in general, many RD&Dmore » ideas are applicable to both new Advanced Light Water Reactor (ALWR) plants and currently operating plants.« less

Risk Informed Margins Management as part of Risk Informed Safety Margin Characterization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curtis Smith

2014-06-01

The ability to better characterize and quantify safety margin is important to improved decision making about Light Water Reactor (LWR) design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plantmore » safety and performance will become known. To support decision making related to economics, readability, and safety, the Risk Informed Safety Margin Characterization (RISMC) Pathway provides methods and tools that enable mitigation options known as risk informed margins management (RIMM) strategies.« less

Theoretical study on sensitivity enhancement in energy-deficit region of chemically amplified resists used for extreme ultraviolet lithography

NASA Astrophysics Data System (ADS)

Kozawa, Takahiro; Santillan, Julius Joseph; Itani, Toshiro

2017-10-01

The role of photons in lithography is to transfer the energy and information required for resist pattern formation. In the information-deficit region, a trade-off relationship is observed between line edge roughness (LER) and sensitivity. However, the sensitivity can be increased without increasing LER in the energy-deficit region. In this study, the sensitivity enhancement limit was investigated, assuming line-and-space patterns with a half-pitch of 11 nm. LER was calculated by a Monte Carlo method. It was unrealistic to increase the sensitivity twofold while keeping the line width roughness (LWR) within 10% critical dimension (CD), whereas the twofold sensitivity enhancement with 20% CD LWR was feasible. The requirements are roughly that the sensitization distance should be less than 2 nm and that the total sensitizer concentration should be higher than 0.3 nm-3.

BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) - Generation Methodology and Preliminary Testing of two ENEA-Bologna Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

NASA Astrophysics Data System (ADS)

Pescarini, Massimo; Sinitsa, Valentin; Orsi, Roberto; Frisoni, Manuela

2016-02-01

Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20 γ) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42 γ) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 SN code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses.

Overcoming etch challenges related to EUV based patterning (Conference Presentation)

NASA Astrophysics Data System (ADS)

Metz, Andrew W.; Cottle, Hongyun; Honda, Masanobu; Morikita, Shinya; Kumar, Kaushik A.; Biolsi, Peter

2017-04-01

Research and development activities related to Extreme Ultra Violet [EUV] defined patterning continue to grow for < 40 nm pitch applications. The confluence of high cost and extreme process control challenges of Self-Aligned Quad Patterning [SAQP] with continued momentum for EUV ecosystem readiness could provide cost advantages in addition to improved intra-level overlay performance relative to multiple patterning approaches. However, Line Edge Roughness [LER] and Line Width Roughness [LWR] performance of EUV defined resist images are still far from meeting technology needs or ITRS spec performance. Furthermore, extreme resist height scaling to mitigate flop over exacerbates the plasma etch trade-offs related to traditional approaches of PR smoothing, descum implementation and maintaining 2D aspect ratios of short lines or elliptical contacts concurrent with ultra-high photo resist [PR] selectivity. In this paper we will discuss sources of LER/LWR, impact of material choice, integration, and innovative plasma process techniques and describe how TELTM VigusTM CCP Etchers can enhance PR selectivity, reduce LER/LWR, and maintain 2D aspect ratio of incoming patterns. Beyond traditional process approaches this paper will show the utility of: [1] DC Superposition in enhancing EUV resist hardening and selectivity, increasing resistance to stress induced PR line wiggle caused by CFx passivation, and mitigating organic planarizer wiggle; [2] Quasi Atomic Layer Etch [Q-ALE] for ARC open eliminating the tradeoffs between selectivity, CD, and shrink ratio control; and [3] ALD+Etch FUSION technology for feature independent CD shrink and LER reduction. Applicability of these concepts back transferred to 193i based lithography is also confirmed.

Advancing Traffic Flow Theory Using Empirical Microscopic Data

DOT National Transportation Integrated Search

2015-01-01

As reviewed in Section 1.1, much of traffic flow theory depends a fundamental relationship (FR) between flow, density, and space mean speed; either explicitly, e.g., hydrodynamic models such as LWR (Lighthill and Whitham, 1955, and Richards, 1956) or...

78 FR 59982 - Revisions to Radiation Protection

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-30

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0268] Revisions to Radiation Protection AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan section; issuance. SUMMARY: The U.S. Nuclear Regulatory... for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition'': Section 12.1...

Results of using frequency banded SAFT for examining three types of defects

NASA Astrophysics Data System (ADS)

Clayton, Dwight; Barker, Alan; Santos-Villalobos, Hector

2017-02-01

A multitude of concrete-based structures are typically part of a light water reactor (LWR) plant to provide the foundation, support, shielding, and containment functions. Concrete has been used in the construction of nuclear power plants (NPPs) because of three primary properties; its low cost, structural strength, and ability to shield radiation. Examples of concrete structures important to the safety of LWR plants include the containment building, spent fuel pool, and cooling towers. This use has made concrete's long-term performance crucial for the safe operation of commercial NPPs. Extending reactor life to 60 years and beyond will likely increase susceptibility and severity of known forms of degradation. Additionally, new mechanisms of materials degradation are also possible. Specially designed and fabricated test specimens can provide realistic flaws that are similar to actual flaws in terms of how they interact with a particular Nondestructive Evaluation (NDE) technique. Artificial test blocks allow the isolation of certain testing problems as well as the variation of certain parameters. Because conditions in the laboratory are controlled, the number of unknown variables can be decreased, making it possible to focus on specific aspects, investigate them in detail, and gain further information on the capabilities and limitations of each method. To minimize artifacts caused by boundary effects, the dimensions of the specimens should not be too compact. In this paper, we apply the frequency banded Synthetic Aperture Focusing Technique (SAFT) technique to a 2.134 m × 2.134 m × 1.016 m concrete test specimen with twenty deliberately embedded defects. These twenty embedded defects simulate voids (honeycombs), delamination, and embedded organic construction debris. Using the time-frequency technique of wavelet packet decomposition and reconstruction, the spectral content of the signal can be divided into two resulting child nodes. The resulting two nodes can then also be divided into two child nodes with each child node containing half of the bandwidth (spectral content) of its parent node. This process can be repeated until the bandwidth of the child nodes is sufficiently small. Once the desired bandwidth has been obtained, the band limited signal can be analyzed using SAFT, enabling the visualization of reflectivity of a frequency band and that band's interaction with the contents of the concrete structure. This paper examines the benefits of using frequency banded SAFT.

Production of Liquid Synthetic Fuels from Carbon, Water and Nuclear Power on Ships and at Shore Bases for Military and Potential Commercial Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Locke Bogart, S.; Schultz, Ken; Brown, Lloyd

2006-07-01

It is demonstrable that synthetic fuels (jet/diesel/gasoline {approx_equal} (CH{sub 2}){sub n}) can be produced from carbon, water, and nuclear energy. What remains to be shown is that all system processes are scalable, integrable, and economical. Sources of carbon include but are not limited to CO{sub 2} from the atmosphere or seawater, CO{sub 2} from fossil-fired power plants, and elemental carbon from coal or biomass. For mobile defense (Navy) applications, the ubiquitous atmosphere is our chosen carbon source. For larger-scale sites such as Naval Advance Bases, the atmosphere may still be the choice should other sources not be readily available. However, at many locations suitable for defense and, potentially, commercial syn-fuel production, far higher concentrations of carbon may be available. The rationale for this study was manifold: fuel system security from terrorism and possible oil embargoes; rising demand and, eventually, peaking supply of conventional petroleum; and escalating costs and prices of fuels. For these reasons, the initial parts of the study were directed at Syn-fuel production for mobile Naval platforms and shore sites such as Rokkasho, Japan (as an exemplar). Nuclear reactors would provide the energy for H{sub 2} from water-splitting, Membrane Gas Absorption (MGA) would extract CO{sub 2} from the atmosphere, the Reverse Water-Gas Reaction (RWGR) would convert the CO{sub 2} to CO, and the resultant H{sub 2} and CO feeds would be converted to (CH{sub 2})n by the Fischer-Tropsch reaction. Many of these processes exist at commercial scale. Some, particularly MGA and RWGR, have been demonstrated at the bench-scale, requiring up-scaling. Likewise, the demonstration of an integrated system at some scale is yet to be done. For ship-based production, it has been shown that the system should be viable and, under reasonable assumptions, both scalable and economical for defense fuels. For the assumptions in the study, fuel cost estimates range from {approx}more » $2.55 to $$4.75 per gallon with a nominal cost of {approx} $$3.65 per gallon. For large installations and advanced nuclear power and hydrogen production systems (high temperature reactors and thermo-chemical hydrogen production), then fuel production might be produced at near-commercial fuel prices. For the H2-MHR and plausible assumptions and estimates of CO{sub 2} extraction and fuel synthesis capital and operating costs, such fuels might have nominal and low production costs ranging from {approx} $2.40 to $$1.70 per gallon, respectively, for a Public Sector Fixed Charge Rate of 5%. Next, it was shown that for CO{sub 2} provided from a fossil-fired power plant, a CO{sub 2} 'disposal' fee of $$30/tonne and a Fixed Charge Rate of 10%, then syn-fuel might be produced at {approx} $3.00 and $2.45 (nominal cost values) and $1.90 and $$1.85 (low cost values) per gallon by LWRs and H2-MHRs, respectively. Last, it was shown that nuclear-produced H{sub 2} and O{sub 2} could convert coal to liquid fuels at very low cost. For a Fixed Charge Rate of 10% and nominal plant costs, fuel costs ranged from {approx} $$1.60 (LWR) per gallon to {approx} $$1.30 (H2-MHR) for an assumed CO{sub 2} avoidance credit of $$30/Tonne. Our studies have shown that the addition of nuclear-produced hydrogen and oxygen to the coal syn-fuel process can greatly reduce CO{sub 2} production and, for modest CO{sub 2} credit, can further reduce the cost of the syn-fuel. Capturing CO{sub 2} from stack gas or even the air will further reduce the amount of CO{sub 2} that must be dealt with. This last case is independent of the price of fossil fuels and liquid fuel production costs and prices will have been capped. Of possibly even greater importance, the carbon fuel cycle will have been closed, thus minimizing or eliminating concerns with Global Climate Change. (authors)« less

Unbiased roughness measurements: the key to better etch performance

NASA Astrophysics Data System (ADS)

Liang, Andrew; Mack, Chris; Sirard, Stephen; Liang, Chen-wei; Yang, Liu; Jiang, Justin; Shamma, Nader; Wise, Rich; Yu, Jengyi; Hymes, Diane

2018-03-01

Edge placement error (EPE) has become an increasingly critical metric to enable Moore's Law scaling. Stochastic variations, as characterized for lines by line width roughness (LWR) and line edge roughness (LER), are dominant factors in EPE and known to increase with the introduction of EUV lithography. However, despite recommendations from ITRS, NIST, and SEMI standards, the industry has not agreed upon a methodology to quantify these properties. Thus, differing methodologies applied to the same image often result in different roughness measurements and conclusions. To standardize LWR and LER measurements, Fractilia has developed an unbiased measurement that uses a raw unfiltered line scan to subtract out image noise and distortions. By using Fractilia's inverse linescan model (FILM) to guide development, we will highlight the key influences of roughness metrology on plasma-based resist smoothing processes. Test wafers were deposited to represent a 5 nm node EUV logic stack. The patterning stack consists of a core Si target layer with spin-on carbon (SOC) as the hardmask and spin-on glass (SOG) as the cap. Next, these wafers were exposed through an ASML NXE 3350B EUV scanner with an advanced chemically amplified resist (CAR). Afterwards, these wafers were etched through a variety of plasma-based resist smoothing techniques using a Lam Kiyo conductor etch system. Dense line and space patterns on the etched samples were imaged through advanced Hitachi CDSEMs and the LER and LWR were measured through both Fractilia and an industry standard roughness measurement software. By employing Fractilia to guide plasma-based etch development, we demonstrate that Fractilia produces accurate roughness measurements on resist in contrast to an industry standard measurement software. These results highlight the importance of subtracting out SEM image noise to obtain quicker developmental cycle times and lower target layer roughness.

EUV process improvement with novel litho track hardware

NASA Astrophysics Data System (ADS)

Stokes, Harold; Harumoto, Masahiko; Tanaka, Yuji; Kaneyama, Koji; Pieczulewski, Charles; Asai, Masaya

2017-03-01

Currently, there are many developments in the field of EUV lithography that are helping to move it towards increased HVM feasibility. Targeted improvements in hardware design for advanced lithography are of interest to our group specifically for metrics such as CD uniformity, LWR, and defect density. Of course, our work is focused on EUV process steps that are specifically affected by litho track performance, and consequently, can be improved by litho track design improvement and optimization. In this study we are building on our experience to provide continual improvement for LWR, CDU, and Defects as applied to a standard EUV process by employing novel hardware solutions on our SOKUDO DUO coat develop track system. Although it is preferable to achieve such improvements post-etch process we feel, as many do, that improvements after patterning are a precursor to improvements after etching. We hereby present our work utilizing the SOKUDO DUO coat develop track system with an ASML NXE:3300 in the IMEC (Leuven, Belgium) cleanroom environment to improve aggressive dense L/S patterns.

Oxidation of 304 stainless steel in high-temperature steam

NASA Astrophysics Data System (ADS)

Ishida, Toshihisa; Harayama, Yasuo; Yaguchi, Sinnosuke

1986-08-01

An experiment on oxidation of 304 stainless steel was performed in steam between 900°C and 1350°C, using the spare cladding of the reactor of the nuclear-powered ship Mutsu. The temperature range was appropriate for a postulated loss of coolant accident (LOCA) analysis of a LWR. The oxidation kinetics were found to obey the parabolic law during the first period of 8 min. After the first period, the parabolic reaction rate constant decreased in the case of heating temperatures between 1100°C and 1250°C. At 1250°C, especially, a marked decrease was observed in the oxide scale-forming kinetics when the surface treated initially by mechanical polishing and given a residual stress. This enhanced oxidation resistance was attributed to the presence of a chromium-enriched layer which was detected by use of an X-ray microanalyzer. The oxidation kinetics equation obtained for the first 8 min is applicable to the model calculation of a hypothetical LOCA in a LWR, employing 304 stainless steel cladding.

[Analysis and experimental verification of sensitivity and SNR of laser warning receiver].

PubMed

Zhang, Ji-Long; Wang, Ming; Tian, Er-Ming; Li, Xiao; Wang, Zhi-Bin; Zhang, Yue

2009-01-01

In order to countermeasure increasingly serious threat from hostile laser in modern war, it is urgent to do research on laser warning technology and system, and the sensitivity and signal to noise ratio (SNR) are two important performance parameters in laser warning system. In the present paper, based on the signal statistical detection theory, a method for calculation of the sensitivity and SNR in coherent detection laser warning receiver (LWR) has been proposed. Firstly, the probabilities of the laser signal and receiver noise were analyzed. Secondly, based on the threshold detection theory and Neyman-Pearson criteria, the signal current equation was established by introducing detection probability factor and false alarm rate factor, then, the mathematical expressions of sensitivity and SNR were deduced. Finally, by using method, the sensitivity and SNR of the sinusoidal grating laser warning receiver developed by our group were analyzed, and the theoretic calculation and experimental results indicate that the SNR analysis method is feasible, and can be used in performance analysis of LWR.

Beer fermentation: monitoring of process parameters by FT-NIR and multivariate data analysis.

PubMed

Grassi, Silvia; Amigo, José Manuel; Lyndgaard, Christian Bøge; Foschino, Roberto; Casiraghi, Ernestina

2014-07-15

This work investigates the capability of Fourier-Transform near infrared (FT-NIR) spectroscopy to monitor and assess process parameters in beer fermentation at different operative conditions. For this purpose, the fermentation of wort with two different yeast strains and at different temperatures was monitored for nine days by FT-NIR. To correlate the collected spectra with °Brix, pH and biomass, different multivariate data methodologies were applied. Principal component analysis (PCA), partial least squares (PLS) and locally weighted regression (LWR) were used to assess the relationship between FT-NIR spectra and the abovementioned process parameters that define the beer fermentation. The accuracy and robustness of the obtained results clearly show the suitability of FT-NIR spectroscopy, combined with multivariate data analysis, to be used as a quality control tool in the beer fermentation process. FT-NIR spectroscopy, when combined with LWR, demonstrates to be a perfectly suitable quantitative method to be implemented in the production of beer. Copyright © 2014 Elsevier Ltd. All rights reserved.

Effects on electron scattering and resist characteristics using assisting underlayers for e-beam direct write lithography

NASA Astrophysics Data System (ADS)

Thrun, Xaver; Choi, Kang-Hoon; Hanisch, Norbert; Hohle, Christoph; Steidel, Katja; Guerrero, Douglas; Figueiro, Thiago; Bartha, Johann W.

2013-03-01

Resist processing for future technology nodes becomes more and more complex. The resist film thickness is getting thinner and hardmask concepts (trilayer) are needed for reproducible etch transfer into the stack. Additional layers between resist and substrate are influencing the electron scattering in e-beam lithography and may also improve sensitivity and resolution. In this study, bare silicon wafers with different assisting underlayers were processed in a 300 mm CMOS manufacturing environment and were exposed on a 50 keV VISTEC SB3050DW variable-shaped electron beam direct writer at Fraunhofer CNT. The underlayers are organic-inorganic hybrid coatings with different metal additives. The negative-tone resist was evaluated in terms of contrast, sensitivity, resolution and LWR/LER as a function of the stack. The interactions between resist and different assisting underlayers on e-beam direct writing will be investigated. These layers could be used to optimize the trade-off among resolution, LWR and sensitivity in future applications.

GEM*STAR: Time for an Alternative Way Forward

NASA Astrophysics Data System (ADS)

Vogelaar, R. Bruce

2011-10-01

The presumption that nuclear reactors will retain their role in global energy production is constantly being challenged - even more so following recent events at Fukushima. Nuclear energy, despite being ``green,'' has inexorably been coupled in the public mind with three paramount concerns: safety, weapons proliferation, and waste (and then ultimately cost). Over the past four decades, the safety of deployed fleets has greatly improved, yet the capital and political costs of a ``nuclear energy option'' appear insurmountable in several countries. The US approach to civilian nuclear energy has become deeply entrenched, first through choices made by the military, and then by the deployed nuclear reactor fleet. This extends to the research agencies as well, to the point where basic sciences and nuclear energy operate in separate spheres. But technologies and priorities have changed, and the time has arrived where a transformative re-think of nuclear energy is not only possible, but urgent. And nuclear physicists are uniquely positioned to accomplish this. This talk will show that by asking, and answering,``what would an accelerator-driven civilian nuclear energy program look like,'' ADNA Corporation's GEM*STAR design directly addresses all three fundamental concerns: safety, proliferation, and waste - and also the final hurdle: cost. GEM*STAR is not an ``add-on'' (to either Project-X, or GEN III+), but rather a base-line energy production capacity, for either electricity or transport fuel production. It integrates and advances the molten-salt reactor technology developed at ORNL, the MW beam accelerator technologies developed by basic sciences, and a reactor/target design optimized for accelerator driven-systems. The results include: the ability to use LWR spent fuel without reprocessing or additional waste; the ability to use natural uranium; no critical mass ever present; orders-of-magnitude less volatile radioactivity in the core; more efficient use of, and deeper burn of actinides, without additional waste; proliferation resistance (no enrichment or reprocessing); high-tolerance to ``beam-trips'' and ultimately, and perhaps most importantly, lower cost electricity or diesel fuel than any currently envisioned new energy source.

An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn T; Tobin, Stephen J; Fensin, Mike L

2009-01-01

There are a variety of reasons for quantifying plutonium (Pu) in spent fuel. Below, five motivations are listed: (1) To verify the Pu content of spent fuel without depending on unverified information from the facility, as requested by the IAEA ('independent verification'). New spent fuel measurement techniques have the potential to allow the IAEA to recover continuity of knowledge and to better detect diversion. (2) To assure regulators that all of the nuclear material of interest leaving a nuclear facility actually arrives at another nuclear facility ('shipper/receiver'). Given the large stockpile of nuclear fuel at reactor sites around the world,more » it is clear that in the coming decades, spent fuel will need to be moved to either reprocessing facilities or storage sites. Safeguarding this transportation is of significant interest. (3) To quantify the Pu in spent fuel that is not considered 'self-protecting.' Fuel is considered self-protecting by some regulatory bodies when the dose that the fuel emits is above a given level. If the fuel is not self-protecting, then the Pu content of the fuel needs to be determined and the Pu mass recorded in the facility's accounting system. This subject area is of particular interest to facilities that have research-reactor spent fuel or old light-water reactor (LWR) fuel. It is also of interest to regulators considering changing the level at which fuel is considered self-protecting. (4) To determine the input accountability value at an electrochemical processing facility. It is not expected that an electrochemical reprocessing facility will have an input accountability tank, as is typical in an aqueous reprocessing facility. As such, one possible means of determining the input accountability value is to measure the Pu content in the spent fuel that arrives at the facility. (5) To fully understand the composition of the fuel in order to efficiently and safely pack spent fuel into a long-term repository. The NDA of spent fuel can be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC&A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument will be quantified for the full assembly library for measurements in three different media: air, water and borated water. The 12 NDA techniques being researched are the following: Delayed Gamma, Delayed Neutrons, Differential Die-Away, Lead Slowing Down Spectrometer, Neutron Multiplicity, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Passive Neutron Albedo Reactivity, Self-integration Neutron Resonance Densitometry, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection.« less

Evaluation of PLS, LS-SVM, and LWR for quantitative spectroscopic analysis of soils

USDA-ARS?s Scientific Manuscript database

Soil testing requires the analysis of large numbers of samples in laboratory that are often time consuming and expensive. Mid-infrared spectroscopy (mid-IR) and near-infrared spectroscopy (NIRS) are fast, non-destructive, and inexpensive analytical methods that have been used for soil analysis, in l...

78 FR 4477 - Review of Safety Analysis Reports for Nuclear Power Plants, Introduction

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-22

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0268] Review of Safety Analysis Reports for Nuclear Power... Analysis Reports for Nuclear Power Plants: LWR Edition.'' The new subsection is the Standard Review Plan... Nuclear Power Plants: Integral Pressurized Water Reactor (iPWR) Edition.'' DATES: Comments must be filed...

78 FR 48503 - Proposed Revision to Missiles Generated by Extreme Winds

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-08

...-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR..., ``Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants,'' and Interim Staff Guidance DC... and Hurricane Missiles for Nuclear Power Plants'' (ADAMS, Accession No. ML110940300), and Interim...

78 FR 48504 - Proposed Revisions to Maintenance Rule Standard Review Plan

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-08

... Review Plan AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan-draft section revision... Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition,'' Section 17... and Management System (ADAMS): You may access publicly available documents online in the NRC Library...

78 FR 41434 - Proposed Revisions to Design of Structures, Components, Equipment and Systems

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-10

..., Components, Equipment and Systems AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan-draft... Systems, Piping Components and their Associated Supports,'' of NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition.'' DATES: Submit comments by...

Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robinson, Sharon M.; Chattin, Marc Rhea; Giaquinto, Joseph

2015-09-01

It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T 2O. In a standard processing flowsheet, tritium management would bemore » accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. Samples of Surry-2 and H.B. Robinson pressurized water reactor cladding were heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. Cladding samples were also heated within the temperature range of 480–600ºC expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. The tritium content of the Surry-2 and H.B. Robinson cladding was measured to be ~234 and ~500 µCi/g, respectively. Heating the Surry-2 cladding at 500°C for 24 h removed ~0.2% of the tritium from the cladding, and heating at 700°C for 24 h removed ~9%. Heating the H.B. Robinson cladding at 700°C for 24 h removed ~11% of the tritium. When samples of the Surry-2 and H.B. Robinson claddings were heated at 700°C for 96 h, essentially all of the tritium in the cladding was removed. However, only ~3% of the tritium was removed when a sample of Surry-2 cladding was heated at 600°C for 96 h. These data indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in operations involving cladding recycle.« less

78 FR 54864 - Light-Walled Rectangular Pipe and Tube From Mexico: Preliminary Results and Partial Rescission of...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-06

... DEPARTMENT OF COMMERCE International Trade Administration [A-201-836] Light-Walled Rectangular... the antidumping duty order on light-walled rectangular pipe and tube (LWR pipe and tube) from Mexico... The merchandise subject to the order is certain welded carbon- quality light-walled steel pipe and...

Quantitative Residual Strain Analyses on Strain Hardened Nickel Based Alloy

NASA Astrophysics Data System (ADS)

Yonezawa, Toshio; Maeguchi, Takaharu; Goto, Toru; Juan, Hou

Many papers have reported about the effects of strain hardening by cold rolling, grinding, welding, etc. on stress corrosion cracking susceptibility of nickel based alloys and austenitic stainless steels for LWR pipings and components. But, the residual strain value due to cold rolling, grinding, welding, etc. is not so quantitatively evaluated.

78 FR 13911 - Proposed Revision to Design of Structures, Components, Equipment and Systems

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-01

... Analysis Reports for Nuclear Power Plants: LWR Edition,'' Section 3.7.1, ``Seismic Design Parameters,'' Section 3.7.2, ``Seismic System Analysis,'' Section 3.7.3, ``Seismic Subsystem Analysis,'' Section 3.8.1... and analysis issues, (2) updates to review interfaces to improve the efficiency and consistency of...

78 FR 41436 - Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-10

... Safety Analysis Reports for Nuclear Power Plants: LWR Edition,'' on a proposed new section to its... revised position on the treatment of the high winds external hazard for certain RTNSS structures, systems... winds external hazard for certain RTNSS structures, systems and components (SSCs). This position differs...

78 FR 31614 - Implementation of Regulatory Guide 1.221 on Design-Basis Hurricane and Hurricane Missiles

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-24

... for Nuclear Power Plants,'' in support of NRC reviews of early site permit (ESP), standard design... NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants... License Applications for Nuclear Power Plants, (LWR Edition)'' (ML070630003) In addition, this ISG...

78 FR 48727 - Proposed Revisions to Design of Structures, Components, Equipment and Systems

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-09

... Analysis Reports for Nuclear Power Plants: LWR Edition,'' Section 3.9.3 ``ASME Code Class 1, 2, and 3...'s Agencywide Documents Access and Management System (ADAMS): You may access publicly available... operational readiness of snubbers (ADAMS Accession No. ML070720041), and review interfaces have been updated...

Monte Carlo simulation of edge placement error

NASA Astrophysics Data System (ADS)

Kobayashi, Shinji; Okada, Soichiro; Shimura, Satoru; Nafus, Kathleen; Fonseca, Carlos; Estrella, Joel; Enomoto, Masashi

2018-03-01

In the discussion of edge placement error (EPE), we proposed interactive pattern fidelity error (IPFE) as an indicator to judge pass/fail of integrated patterns. IPFE consists of lower and upper layer EPEs (CD and center of gravity: COG) and overlay, which is decided from the combination of each maximum variation. We succeeded in obtaining the IPFE density function by Monte Carlo simulation. In the results, we also found that the standard deviation (σ) of each indicator should be controlled by 4.0σ, at the semiconductor grade, such as 100 billion patterns per die. Moreover, CD, COG and overlay were analyzed by analysis of variance (ANOVA); we can discuss all variations from wafer to wafer (WTW), pattern to pattern (PTP), line edge roughness (LWR) and stochastic pattern noise (SPN) on an equal footing. From the analysis results, we can determine that these variations belong to which process and tools. Furthermore, measurement length of LWR is also discussed in ANOVA. We propose that the measurement length for IPFE analysis should not be decided to the micro meter order, such as >2 μm length, but for which device is actually desired.

Setting up a proper power spectral density (PSD) and autocorrelation analysis for material and process characterization

NASA Astrophysics Data System (ADS)

Rutigliani, Vito; Lorusso, Gian Francesco; De Simone, Danilo; Lazzarino, Frederic; Rispens, Gijsbert; Papavieros, George; Gogolides, Evangelos; Constantoudis, Vassilios; Mack, Chris A.

2018-03-01

Power spectral density (PSD) analysis is playing more and more a critical role in the understanding of line-edge roughness (LER) and linewidth roughness (LWR) in a variety of applications across the industry. It is an essential step to get an unbiased LWR estimate, as well as an extremely useful tool for process and material characterization. However, PSD estimate can be affected by both random to systematic artifacts caused by image acquisition and measurement settings, which could irremediably alter its information content. In this paper, we report on the impact of various setting parameters (smoothing image processing filters, pixel size, and SEM noise levels) on the PSD estimate. We discuss also the use of PSD analysis tool in a variety of cases. Looking beyond the basic roughness estimate, we use PSD and autocorrelation analysis to characterize resist blur[1], as well as low and high frequency roughness contents and we apply this technique to guide the EUV material stack selection. Our results clearly indicate that, if properly used, PSD methodology is a very sensitive tool to investigate material and process variations

The LER/LWR metrology challenge for advance process control through 3D-AFM and CD-SEM

NASA Astrophysics Data System (ADS)

Faurie, P.; Foucher, J.; Foucher, A.-L.

2009-12-01

The continuous shrinkage in dimensions of microelectronic devices has reached such level, with typical gate length in advance R&D of less than 20nm combine with the introduction of new architecture (FinFET, Double gate...) and new materials (porous interconnect material, 193 immersion resist, metal gate material, high k materials...), that new process parameters have to be well understood and well monitored to guarantee sufficient production yield in a near future. Among these parameters, there are the critical dimensions (CD) associated to the sidewall angle (SWA) values, the line edge roughness (LER) and the line width roughness (LWR). Thus, a new metrology challenge has appeared recently and consists in measuring "accurately" the fabricated patterns on wafers in addition to measure the patterns on a repeatable way. Therefore, a great effort has to be done on existing techniques like CD-SEM, Scatterometry and 3D-AFM in order to develop them following the two previous criteria: Repeatability and Accuracy. In this paper, we will compare the 3D-AFM and CD-SEM techniques as a mean to measure LER and LWR on silicon and 193 resist and point out CD-SEM impact on the material during measurement. Indeed, depending on the material type, the interaction between the electron beam and the material or between the AFM tip and the material can vary a lot and subsequently can generate measurements bias. The first results tend to show that depending on CD-SEM conditions (magnification, number of acquisition frames) the final outputs can vary on a large range and therefore show that accuracy in such measurements are really not obvious to obtain. On the basis of results obtained on various materials that present standard sidewall roughness, we will show the limit of each technique and will propose different ways to improve them in order to fulfil advance roadmap requirements for the development of the next IC generation.

Status Report on the High-Temperature Steam Electrolysis Plant Model Developed in the Modelica Framework (FY17)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Jong Suk; Bragg-Sitton, Shannon M.; Boardman, Richard D.

This report has been prepared as part of an effort to design and build a modeling and simulation (M&S) framework to assess the economic viability of a nuclear-renewable hybrid energy system (N-R HES). In order to facilitate dynamic M&S of such an integrated system, research groups in multiple national laboratories have been developing various subsystems as dynamic physics-based components using the Modelica programming language. In fiscal year 2015 (FY15), Idaho National Laboratory (INL) performed a dynamic analysis of two region-specific N-R HES configurations, including the gas-to-liquid (natural gas to Fischer-Tropsch synthetic fuel) and brackish water reverse osmosis desalination plants asmore » industrial processes. In FY16, INL developed two additional subsystems in the Modelica framework: (1) a high-temperature steam electrolysis (HTSE) plant as a high priority industrial plant to be integrated with a light water reactor (LWR) within an N-R HES and (2) a gas turbine power plant as a secondary energy supply. In FY17, five new components (i.e., a feedwater pump, a multi-stage compression system, a sweep-gas turbine, flow control valves, and pressure control valves) have been incorporated into the HTSE system proposed in FY16, aiming to better realistically characterize all key components of concern. Special attention has been given to the controller settings based on process models (i.e., direct synthesis method), aiming to improve process dynamics and controllability. A dynamic performance analysis of the improved LWR/HTSE integration case was carried out to evaluate the technical feasibility (load-following capability) and safety of such a system operating under highly variable conditions requiring flexible output. The analysis (evaluated in terms of the step response) clearly shows that the FY17 model resulted in superior output responses with much smaller settling times and less oscillatory behavior in response to disturbances in the electric load than those observed with the FY16 model. Simulation results involving several case studies show that the suggested control scheme could maintain the controlled variables (including the steam utilization factor, cathode stream inlet composition, and temperatures and pressures of the process streams at various locations) within desired limits under various plant operating conditions. The results also indicate that the proposed HTSE plant could provide operational flexibility to participate in energy management at the utility scale by dynamically optimizing the use of excess plant capacity within an N-R HES.« less

Steam Oxidation Testing in the Severe Accident Test Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pint, Bruce A.

After the March 2011 accident at Fukushima Daiichi, Oak Ridge National Laboratory (ORNL) began conducting high temperature steam oxidation testing of candidate materials for accident tolerant fuel (ATF) cladding in August 2011 [1-11]. The ATF concept is to enhance safety margins in light water reactors (LWR) during severe accident scenarios by identifying materials with 100× slower steam oxidation rates compared to current Zr-based alloys. In 2012, the ORNL laboratory equipment was expanded and made available to the entire ATF community as the Severe Accident Test Station (SATS) [4,12]. Compared to the current UO2/Zr-based alloy fuel system, an ATF alternative wouldmore » significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [13-14]. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models [15-17]. However, initial modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. Also, because many accident scenarios include steadily increasing temperatures, the required data are not traditional isothermal exposures but exposures with varying “ramp” rates. In some cases, the steam oxidation behavior has been surprising and difficult to interpret. Thus, more fundamental information continues to be collected. In addition, more work continues to focus on commercially-manufactured tube material. This report summarizes recent work to characterize the behavior of candidate alloys exposed to high temperature steam, evaluate steam oxidation behavior in various ramp scenarios and continue to collect integral data on FeCrAl compared to conventional Zr-based cladding.« less

Thermo-mechanical assessment of full SiC/SiC composite cladding for LWR applications with sensitivity analysis

NASA Astrophysics Data System (ADS)

Singh, Gyanender; Terrani, Kurt; Katoh, Yutai

2018-02-01

SiC/SiC composites are considered among leading candidates for accident tolerant fuel cladding in light water reactors. However, when SiC-based materials are exposed to neutron irradiation, they experience significant changes in dimensions and physical properties. Under a large heat flux application (i.e. fuel cladding), the non-uniform changes in the dimensions and physical properties will lead to build-up of stresses in the structure over the course of time. To ensure reliable and safe operation of such a structure it is important to assess its thermo-mechanical performance under in-reactor conditions of irradiation and elevated temperature. In this work, the foundation for 3D thermo-mechanical analysis of SiC/SiC cladding is put in place and a set of analyses with simplified boundary conditions has been performed. The analyses were carried out with two different codes that were benchmarked against one another and prior results in the literature. A constitutive model is constructed and solved numerically to predict the stress distribution and variation in the cladding under normal operating conditions. The dependence of dimensions and physical properties variation with irradiation and temperature has been incorporated. These robust models may now be modified to take into account the axial and circumferential variation in neutron and heat flux to fully account for 3D effects. The results from the simple analyses show the development of high tensile stresses especially in the circumferential and axial directions at the inner region of the cladding. Based on the results obtained, design guidelines are recommended. For lack of certainty in or tailor-ability for the physical and mechanical properties of SiC/SiC composite material a sensitivity analysis is conducted. The analysis results establish a precedence order of the properties based on the extent to which these properties influence the temperature and the stresses.

76 FR 77025 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Section 1.0, Revision...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-09

... the Final Staff Guidance Section 1.0, Revision 2 on Introduction and Interfaces AGENCY: Nuclear... Plants: LWR Edition,'' Section 1.0, Revision 2 on ``Introduction and Interfaces'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML112730393). The NRC staff issues revisions to SRP...

78 FR 1199 - Light-Walled Rectangular Pipe and Tube From Mexico: Final Results of Antidumping Duty...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-08

... DEPARTMENT OF COMMERCE International Trade Administration [A-201-836] Light-Walled Rectangular... order on light-walled rectangular pipe and tube (LWR pipe and tube) from Mexico. This review covers two... but received no such comments. We also did not receive a request for a hearing. \\1\\ See Light-Walled...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2013 CFR

2013-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2014 CFR

2014-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

Evaluation on the Effect of Composition on Radiation Hardening and Embrittlement in Model FeCrAl Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Briggs, Samuel A.; Edmondson, Philip

2015-09-18

This report details the findings of post-radiation mechanical testing and microstructural characterization performed on a series of model and commercial FeCrAl alloys to assist with the development of a cladding technology with enhanced accident tolerance. The samples investigated include model alloys with simple ferritic grain structure and two commercial alloys with minor solute additions. These samples were irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) up to nominal doses of 7.0 dpa near or at Light Water Reactor (LWR) relevant temperatures (300-400 C). Characterization included a suite of techniques including small angle neutron scatteringmore » (SANS), atom probe tomography (APT), and transmission based electron microscopy techniques. Mechanical testing included tensile tests at room temperature on sub-sized tensile specimens. The goal of this work was to conduct detailed characterization and mechanical testing to begin establishing empirical and/or theoretical structure-property relationships for radiation-induced hardening and embrittlement in the FeCrAl alloy class. Development of such relationships will provide insight on the performance of FeCrAl alloys in an irradiation environment and will enable further development of the alloy class for applications within a LWR environment. A particular focus was made on establishing trends, including composition and radiation dose. The report highlights in detail the pertinent findings based on this work. This report shows that radiation hardening in the alloys is primarily composition dependent due to the phase separation in the high-Cr FeCrAl alloys. Other radiation induced/enhanced microstructural features were less dependent on composition and when observed at low number densities, were not a significant contributor to the observed mechanical responses. Pre-existing microstructure in the alloys was found to be important, with grain boundaries and pre-existing dislocation networks acting as defect sinks, resulting in variations in the observed microstructures after irradiation. Dose trends were also observed, with increasing radiation dose promoting changes in the size and number density of the Cr-rich α' precipitates. Based on the microstructural analysis, performed tensile testing, and prior knowledge from FeCr literature it was hypothesized that the formation of the Cr-rich α' precipitates could lead to significant radiation-induced embrittlement in the alloys, and this could be composition dependent, a result which would mirror the trends observed for radiation-induced hardening. Due to the limited database on embrittlement in the FeCrAl alloy class after irradiation, a series of radiation experiments have been implemented. The overarching point of view within this report is the radiation tolerance of FeCrAl is complex, with many mechanisms and factors to be considered at once. Further development of the FeCrAl alloy class for enhanced accident tolerant applications requires detailed, single (or at least limited) variable experiments to fully comprehend and predict the performance of this alloy in LWRs. This report has been submitted as fulfillment of milestone M2FT-15OR0202321 titled, Summary report on the effect of composition on the irradiation embrittlement of Gen 1 ATF FeCrAl for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.« less

Detection of bacteria using bacteriophage with hollow gold nanostructures immobilized fiber optic sensor

NASA Astrophysics Data System (ADS)

Halkare, Pallavi; Punjabi, Nirmal; Wangchuk, Jigme; Kondabagil, Kiran; Mukherji, Soumyo

2016-04-01

Hollow gold nanostructures (HGNS) have been used in variety of optical biosensors due to their inherent advantage of operating at near infra red (NIR) wavelength, large extinction coefficient and high dielectric sensitivity. The absorption wavelength of these nanostructures can be modulated by changing the ratio of hollow region to the core shell thickness. The aim of the present study is to incorporate the properties of HGNS, to develop LSPR based U-bent fiber optic sensor for detection of pathogens. The detection was carried out using an experimental set up consisting of a white light source, 200 μm diameter optical fiber having bend diameter of 1.6 mm +/- 0. 2 mm and a spectrometer. The HGNS were immobilized on the decladded portion of the fiber optic probe by chemisorptions. The effective plasmon penetration depth of the HGNS immobilized fiber optic sensor was approximated by using alternating layers of positively and negatively charged polyelectrolytes. The HGNS immobilized U-bent fiber optic sensor was used for detection of E.coli B40 strain using bacteriophage T4. The preliminary experiments were carried out with 104 cfu/ml of E.coli B40 and the change in absorbance obtained was approx. 0.042 +/- 0.0045 abs. units (n = 3). The response of this sensor was found to be better than spherical gold nanoparticle immobilized sensing platforms.

Accommodating Remedial Readers in the General Education Setting: Is Listening-while-Reading Sufficient to Improve Factual and Inferential Comprehension?

ERIC Educational Resources Information Center

Schmitt, Ara J.; Hale, Andrea D.; McCallum, Elizabeth; Mauck, Brittany

2011-01-01

Word reading accommodations are commonly applied in the general education setting in an attempt to improve student comprehension and learning of curriculum content. This study examined the effects of listening-while-reading (LWR) and silent reading (SR) using text-to-speech assistive technology on the comprehension of 25 middle-school remedial…

75 FR 5632 - Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-03

....8 and Regulatory Guide 1.206, ``Combined License Applications for Nuclear Power Plants (LWR Edition... Management System (ADAMS) Accession No. ML092640035). This ISG provides new guidance information for... (SRP), Section 8.3.1 and Sections 9.5.4 through 9.5.8. The NRC staff issues DC/COL-ISGs to facilitate...

Uniform corrosion of FeCrAl alloys in LWR coolant environments

NASA Astrophysics Data System (ADS)

Terrani, K. A.; Pint, B. A.; Kim, Y.-J.; Unocic, K. A.; Yang, Y.; Silva, C. M.; Meyer, H. M.; Rebak, R. B.

2016-10-01

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300-500 μm thick cladding.

Mechanical property degradation and microstructural evolution of cast austenitic stainless steels under short-term thermal aging

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lach, Timothy G.; Byun, Thak Sang; Leonard, Keith J.

Mechanical testing and microstructural characterization were performed on short-term thermally aged cast austenitic stainless steels (CASS) to understand the severity and mechanisms of thermal-aging degradation experienced during extended operation of light water reactor (LWR) coolant systems. Four CASS materials – CF3, CF3M, CF8, and CF8M – were thermally aged for 1500 hours at 290 °C, 330 °C, 360 °C, and 400 °C. All four alloys experienced insignificant change in strength and ductility properties but a significant reduction in absorbed impact energy. The primary microstructural and compositional changes during thermal aging were spinodal decomposition of the δ-ferrite into α/ α`, precipitationmore » of G-phase in the δ-ferrite, segregation of solute to the austenite/ ferrite interphase boundary, and growth of M23C6 carbides on the austenite/ferrite interphase boundary. These changes were shown to be highly dependent on chemical composition, particularly the concentration of C and Mo, and aging temperature. A comprehensive model is being developed to correlate the microstructural evolution with mechanical behavior and simulation for predictive evaluations of LWR coolant system components.« less

Radiation effects in concrete for nuclear power plants Part I: Quantification of radiation exposure and radiation effects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G; Pape, Yann Le; Remec, Igor

A large fraction of light water reactor (LWR) construction utilizes concrete, including safety-related structures such as the biological shielding and containment building. Concrete is an inherently complex material, with the properties of concrete structures changing over their lifetime due to the intrinsic nature of concrete and influences from local environment. As concrete structures within LWRs age, the total neutron fluence exposure of the components, in particular the biological shield, can increase to levels where deleterious effects are introduced as a result of neutron irradiation. This work summarizes the current state of the art on irradiated concrete, including a review ofmore » the current literature and estimates the total neutron fluence expected in biological shields in typical LWR configurations. It was found a first-order mechanism for loss of mechanical properties of irradiated concrete is due to radiation-induced swelling of aggregates, which leads to volumetric expansion of the concrete. This phenomena is estimated to occur near the end of life of biological shield components in LWRs based on calculations of estimated peak neutron fluence in the shield after 80 years of operation.« less

Introduction of pre-etch deposition techniques in EUV patterning

NASA Astrophysics Data System (ADS)

Xiang, Xun; Beique, Genevieve; Sun, Lei; Labonte, Andre; Labelle, Catherine; Nagabhirava, Bhaskar; Friddle, Phil; Schmitz, Stefan; Goss, Michael; Metzler, Dominik; Arnold, John

2018-04-01

The thin nature of EUV (Extreme Ultraviolet) resist has posed significant challenges for etch processes. In particular, EUV patterning combined with conventional etch approaches suffers from loss of pattern fidelity in the form of line breaks. A typical conventional etch approach prevents the etch process from having sufficient resist margin to control the trench CD (Critical Dimension), minimize the LWR (Line Width Roughness), LER (Line Edge Roughness) and reduce the T2T (Tip-to-Tip). Pre-etch deposition increases the resist budget by adding additional material to the resist layer, thus enabling the etch process to explore a wider set of process parameters to achieve better pattern fidelity. Preliminary tests with pre-etch deposition resulted in blocked isolated trenches. In order to mitigate these effects, a cyclic deposition and etch technique is proposed. With optimization of deposition and etch cycle time as well as total number of cycles, it is possible to open the underlying layers with a beneficial over etch and simultaneously keep the isolated trenches open. This study compares the impact of no pre-etch deposition, one time deposition and cyclic deposition/etch techniques on 4 aspects: resist budget, isolated trench open, LWR/LER and T2T.

Progress and process improvements for multiple electron-beam direct write

NASA Astrophysics Data System (ADS)

Servin, Isabelle; Pourteau, Marie-Line; Pradelles, Jonathan; Essomba, Philippe; Lattard, Ludovic; Brandt, Pieter; Wieland, Marco

2017-06-01

Massively parallel electron beam direct write (MP-EBDW) lithography is a cost-effective patterning solution, complementary to optical lithography, for a variety of applications ranging from 200 to 14 nm. This paper will present last process/integration results to achieve targets for both 28 and 45 nm nodes. For 28 nm node, we mainly focus on line-width roughness (LWR) mitigation by playing with stack, new resist platform and bias design strategy. The lines roughness was reduced by using thicker spin-on-carbon (SOC) hardmask (-14%) or non-chemically amplified (non-CAR) resist with bias writing strategy implementation (-20%). Etch transfer into trilayer has been demonstrated by preserving pattern fidelity and profiles for both CAR and non-CAR resists. For 45 nm node, we demonstrate the electron-beam process integration within optical CMOS flows. Resists based on KrF platform show a full compatibility with multiple stacks to fit with conventional optical flow used for critical layers. Electron-beam resist performances have been optimized to fit the specifications in terms of resolution, energy latitude, LWR and stack compatibility. The patterning process overview showing the latest achievements is mature enough to enable starting the multi-beam technology pre-production mode.

Uniform corrosion of FeCrAl alloys in LWR coolant environments

DOE PAGES

Terrani, K. A.; Pint, B. A.; Kim, Y. -J.; ...

2016-06-29

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation ofmore » very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. Finally, the maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ~2 μm, which is inconsequential for a ~300–500 μm thick cladding.« less

Proceedings of the IAEA specialists` meeting on cracking in LWR RPV head penetrations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pugh, C.E.; Raney, S.J.

1996-07-01

This report contains 17 papers that were presented in four sessions at the IAEA Specialists` meeting on Cracking in LWR RPV Head Penetrations held at ASTM Headquarters in Philadelphia on May 2-3, 1995. The papers are compiled here in the order that presentations were made in the sessions, and they relate to operational observations, inspection techniques, analytical modeling, and regulatory control. The goal of the meeting was to allow international experts to review experience in the field of ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. The emphasis was to allow a better understanding of RPV materialmore » behavior, to provide guidance supporting reliability and adequate performance, and to assist in defining directions for further investigations. The international nature of the meeting is illustrated by the fact that papers were presented by researchers from 10 countries. There were technical experts present form other countries who participated in discussions of the results presented. This present document incorporates the final version of the papers as received from the authors. The final chapter includes conclusions and recommendations. Individual papers have been cataloged separately.« less

SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

IJ van Rooyen; WR Lloyd; TL Trowbridge

2013-09-01

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designsmore » being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points and clad configurations. The 2-ply sleeve samples show a higher bend momentum compared to those of the 1-ply sleeve samples. This is applicable to both the hybrid mock-up and bare SiC-CMC sleeve samples. Comparatively both the 1- and 2-ply hybrid mock-up samples showed a higher bend stiffness and strength compared with the standard Zr-4 mock-up sample. The characterization of the hybrid mock-up samples showed signs of distress and preliminary signs of fraying at the protective Zr-4 sleeve areas for the 1-ply SiC-CMC sleeve. In addition, the microstructure of the SiC matrix near the cracks at the region of highest compressive bending strain shows significant cracking and flaking. The 2-ply SiC-CMC sleeve samples showed a more bonded, cohesive SiC matrix structure. This cracking and fraying causes concern for increased fretting during the actual use of the design. Tomography was proven as a successful tool to identify open porosity during pre-test characterization. Although there is currently insufficient data to make conclusive statements regarding the overall merit of the hybrid cladding design, preliminary characterization of this novel design has been demonstrated.« less

THAI Multi-Compartment Containment Test Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kanzleiter, T.; Poss, G.; Funke, F.

2006-07-01

The THAI experimental programme includes combined-effect investigations on thermal hydraulics, hydrogen, and fission product (iodine and aerosols) behaviour in LWR containments under severe accident conditions. An overview on the experiments performed up to now and on the future test program is presented, in combination with a selection of typical results to illustrate the versatility of the test facility and the broad variety of topics investigated. (authors)

Integrating Safety Assessment Methods using the Risk Informed Safety Margins Characterization (RISMC) Approach

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curtis Smith; Diego Mandelli

Safety is central to the design, licensing, operation, and economics of nuclear power plants (NPPs). As the current light water reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of systems, structures, and components (SSC) degradations or failures that initiate safety significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very highmore » degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated primarily based on engineering judgment backed by a set of conservative engineering calculations. The ability to better characterize and quantify safety margin is important to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development (R&D) in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plant safety and performance will become known. To support decision making related to economics, readability, and safety, the RISMC Pathway provides methods and tools that enable mitigation options known as margins management strategies. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. As the lead Department of Energy (DOE) Laboratory for this Pathway, the Idaho National Laboratory (INL) is tasked with developing and deploying methods and tools that support the quantification and management of safety margin and uncertainty.« less

Development of LWR Fuels with Enhanced Accident Tolerance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lahoda, Edward J.; Boylan, Frank A.

2015-10-30

Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U 15N and U 3Si 2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U 3Si 2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U 3Si 2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pelletsmore » for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U 3Si 2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO 2 pellets. Pellets and powders of UO 2, UN, and U 3Si 2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO 2. Cold spray application of either the amorphous steel or the Ti 2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which is the current baseline structure and a SiC to SiC tube closure approach. Permeability tests and mechanical tests were developed to verify the operation of the SiC cladding. Steam autoclave (420°C), high temperature (1200°C) flowing steam tests and quench tests were carried out with minimal corrosion, mechanical or hermeticity degradation effect on the SiC cladding or end plug closure. However, in-reactor loop tests carried out in the MIT reactor indicated an unacceptable degree of corrosion, likely due to the corrosive effect of radiolysis products which attacked the SiC.« less

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Honma, George

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will bemore » used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).« less

Treatment of industrial wastewater effluents using hydrodynamic cavitation and the advanced Fenton process.

PubMed

Chakinala, Anand G; Gogate, Parag R; Burgess, Arthur E; Bremner, David H

2008-01-01

For the first time, hydrodynamic cavitation induced by a liquid whistle reactor (LWR) has been used in conjunction with the advanced Fenton process (AFP) for the treatment of real industrial wastewater. Semi-batch experiments in the LWR were designed to investigate the performance of the process for two different industrial wastewater samples. The effect of various operating parameters such as pressure, H2O2 concentration and the initial concentration of industrial wastewater samples on the extent of mineralization as measured by total organic carbon (TOC) content have been studied with the aim of maximizing the extent of degradation. It has been observed that higher pressures, sequential addition of hydrogen peroxide at higher loadings and lower concentration of the effluent are more favourable for a rapid TOC mineralization. In general, the novel combination of hydrodynamic cavitation with AFP results in about 60-80% removal of TOC under optimized conditions depending on the type of industrial effluent samples. The combination described herein is most useful for treatment of bio-refractory materials where the diminution in toxicity can be achieved up to a certain level and then conventional biological oxidation can be employed for final treatment. The present work is the first to report the use of a hydrodynamic cavitation technique for real industrial wastewater treatment.

Status of the MeLoDIE experiment, an advanced device for the study of the irradiation creep of LWR cladding with full online capabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guimbal, P.; Huotilainen, S.; Taehtinen, S.

2015-07-01

As a prototype of future instrumented material experiments in the Jules Horowitz Reactor (JHR), the MELODIE project was launched in 2009 by the CEA in collaboration with VTT. Being designed as a biaxial creep experiment with online capability, MELODIE is able to apply an online-controlled biaxial loading on a LWR clad sample up to 120 MPa and to perform an online measurement of its biaxial deformation. An important experimental challenge was to perform reliably accurate measurements under the high nuclear heat load of in-core locations while keeping within their tight space. For that purpose, specific sensors were co-designed with andmore » built by IFE Halden. Manufacturing of the MELODIE components was completed one year ago. The complexity of its in-pile section and of the pressurization system requested a step-by-step tuning of the setup. The toughest part of this process dealt with the Diameter gauge which required a partial redesign to take into account unexpected and unwanted electromagnetic interactions with the hosting device. Final cold performance tests of the on-board instrumentation will be presented. The MELODIE device is now ready and irradiation should start in OSIRIS reactor this spring. (authors)« less

Exploration of suitable dry etch technologies for directed self-assembly

NASA Astrophysics Data System (ADS)

Yamashita, Fumiko; Nishimura, Eiichi; Yatsuda, Koichi; Mochiki, Hiromasa; Bannister, Julie

2012-03-01

Directed self-assembly (DSA) has shown the potential to replace traditional resist patterns and provide a lower cost alternative for sub-20-nm patterns. One of the possible roadblocks for DSA implementation is the ability to etch the polymers to produce quality masks for subsequent etch processes. We have studied the effects of RF frequency and etch chemistry for dry developing DSA patterns. The results of the study showed a capacitively-coupled plasma (CCP) reactor with very high frequency (VHF) had superior pattern development after the block co-polymer (BCP) etch. The VHF CCP demonstrated minimal BCP height loss and line edge roughness (LER)/line width roughness (LWR). The advantage of CCP over ICP is the low dissociation so the etch rate of BCP is maintained low enough for process control. Additionally, the advantage of VHF is the low electron energy with a tight ion energy distribution that enables removal of the polymethyl methacrylate (PMMA) with good selectivity to polystyrene (PS) and minimal LER/LWR. Etch chemistries were evaluated on the VHF CCP to determine ability to treat the BCPs to increase etch resistance and feature resolution. The right combination of RF source frequencies and etch chemistry can help overcome the challenges of using DSA patterns to create good etch results.

Optimum ArFi laser bandwidth for 10nm node logic imaging performance

NASA Astrophysics Data System (ADS)

Alagna, Paolo; Zurita, Omar; Timoshkov, Vadim; Wong, Patrick; Rechtsteiner, Gregory; Baselmans, Jan; Mailfert, Julien

2015-03-01

Lithography process window (PW) and CD uniformity (CDU) requirements are being challenged with scaling across all device types. Aggressive PW and yield specifications put tight requirements on scanner performance, especially on focus budgets resulting in complicated systems for focus control. In this study, an imec N10 Logic-type test vehicle was used to investigate the E95 bandwidth impact on six different Metal 1 Logic features. The imaging metrics that track the impact of light source E95 bandwidth on performance of hot spots are: process window (PW), line width roughness (LWR), and local critical dimension uniformity (LCDU). In the first section of this study, the impact of increasing E95 bandwidth was investigated to observe the lithographic process control response of the specified logic features. In the second section, a preliminary assessment of the impact of lower E95 bandwidth was performed. The impact of lower E95 bandwidth on local intensity variability was monitored through the CDU of line end features and the LWR power spectral density (PSD) of line/space patterns. The investigation found that the imec N10 test vehicle (with OPC optimized for standard E95 bandwidth of300fm) features exposed at 200fm showed pattern specific responses, suggesting areas of potential interest for further investigation.

LWR pressure vessel surveillance dosimetry improvement program: LWR power reactor surveillance physics-dosimetry data base compendium

DOE Office of Scientific and Technical Information (OSTI.GOV)

McElroy, W.N.

1985-08-01

This NRC physics-dosimetry compendium is a collation of information and data developed from available research and commercial light water reactor vessel surveillance program (RVSP) documents and related surveillance capsule reports. The data represents the results of the HEDL least-squares FERRET-SAND II Code re-evaluation of exposure units and values for 47 PWR and BWR surveillance capsules for W, B and W, CE, and GE power plants. Using a consistent set of auxiliary data and dosimetry-adjusted reactor physics results, the revised fluence values for E > 1 MeV averaged 25% higher than the originally reported values. The range of fluence values (new/old)more » was from a low of 0.80 to a high of 2.38. These HEDL-derived FERRET-SAND II exposure parameter values are being used for NRC-supported HEDL and other PWR and BWR trend curve data development and testing studies. These studies are providing results to support Revision 2 of Regulatory Guide 1.99. As stated by Randall (Ra84), the Guide is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence.« less

Ice Thermal Storage Systems for LWR Supplemental Cooling and Peak Power Shifting

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haihua Zhao; Hongbin Zhang; Phil Sharpe

2010-06-01

Availability of enough cooling water has been one of the major issues for the nuclear power plant site selection. Cooling water issues have frequently disrupted the normal operation at some nuclear power plants during heat waves and long draught. The issues become more severe due to the new round of nuclear power expansion and global warming. During hot summer days, cooling water leaving a power plant may become too hot to threaten aquatic life so that environmental regulations may force the plant to reduce power output or even temporarily to be shutdown. For new nuclear power plants to be builtmore » at areas without enough cooling water, dry cooling can be used to remove waste heat directly into the atmosphere. However, dry cooling will result in much lower thermal efficiency when the weather is hot. One potential solution for the above mentioned issues is to use ice thermal storage systems (ITS) that reduce cooling water requirements and boost the plant’s thermal efficiency in hot hours. ITS uses cheap off-peak electricity to make ice and uses those ice for supplemental cooling during peak demand time. ITS is suitable for supplemental cooling storage due to its very high energy storage density. ITS also provides a way to shift large amount of electricity from off peak time to peak time. Some gas turbine plants already use ITS to increase thermal efficiency during peak hours in summer. ITSs have also been widely used for building cooling to save energy cost. Among three cooling methods for LWR applications: once-through, wet cooling tower, and dry cooling tower, once-through cooling plants near a large water body like an ocean or a large lake and wet cooling plants can maintain the designed turbine backpressure (or condensation temperature) during 99% of the time; therefore, adding ITS to those plants will not generate large benefits. For once-through cooling plants near a limited water body like a river or a small lake, adding ITS can bring significant economic benefits and avoid forced derating and shutdown during extremely hot weather. For the new plants using dry cooling towers, adding the ice thermal storage systems can effectively reduce the efficiency loss and water consumption during hot weather so that new LWRs could be considered in regions without enough cooling water. \\ This paper presents the feasibility study of using ice thermal storage systems for LWR supplemental cooling and peak power shifting. LWR cooling issues and ITS application status will be reviewed. Two ITS application case studies will be presented and compared with alternative options: one for once-through cooling without enough cooling for short time, and the other with dry cooling. Because capital cost, especially the ice storage structure/building cost, is the major cost for ITS, two different cost estimation models are developed: one based on scaling method, and the other based on a preliminary design using Building Information Modeling (BIM), an emerging technology in Architecture/Engineering/Construction, which enables design options, performance analysis and cost estimating in the early design stage.« less

Fracture toughness evaluation of select advanced replacement alloys for LWR core internals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tan, Lizhen; Chen, Xiang

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications bymore » 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Fracture toughness is one of the key engineering properties required for core internal materials. Together with other properties, which are being examined such as high-temperature steam oxidation resistance, radiation hardening, and irradiation-assisted stress corrosion cracking resistance, the alloys will be down-selected for neutron irradiation study and comprehensive post-irradiation examinations. According to the candidate alloys selected under the ARRM program, ductile fracture toughness of eight alloys was evaluated at room temperature and the LWR-relevant temperatures. The tested alloys include two ferritic alloys (Grade 92 and an oxide-dispersion-strengthened alloy 14YWT), two austenitic stainless steels (316L and 310), four Ni-base superalloys (718A, 725, 690, and X750). Alloy 316L and X750 are included as reference alloys for low- and high-strength alloys, respectively. Compact tension specimens in 0.25T and 0.2T were machined from the alloys in the T-L and R-L orientations according to the product forms of the alloys. This report summarizes the final results of the specimens tested and analyzed per ASTM Standard E1820. Unlike the ferritic alloys showing slight decreases (Grade 92) or significant decreases (14YWT) in fracture toughness at elevated temperatures, the fracture toughness of the austenitic stainless steels and Ni-base superalloys were not strongly dependent upon the test temperatures. The fracture toughness of the alloys at the LWR-relevant temperatures was estimated by averaging the toughness values within 250– 350°C, which suggested the fracture toughness of the alloys in a descending order as 316L (752±98 MPa√m), 310 (513±66 MPa√m), 718A (313±43 MPa√m), 690 (267±48 MPa√m), 725 (218±55 MPa√m), X750 (145±16 MPa√m), Grade 92 (112±12 MPa√m), and 14YWT (63±3 MPa√m). Tearing modulus of the alloys was analyzed in the meantime, which were not strongly dependent upon the test temperatures. The high-strength alloys 718A, 725, X750, and 14YWT had the lowest tearing modulus, ranging from ~45 to ~7. Alloy 690 exhibited the highest tearing modulus on the order of 450, followed by 316L and 310 on the order of 260. Grade 92 had a noticeably lower tearing modulus on the order of 70.« less

The status of ABWR-II development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hiroyuki, Okada; Hideya Kitamura; Kumiaki, Moriya

This paper reports on the current development status of the ABWR-II project, a next generation reactor design based on the ABWR. In the early 90's, a program to develop the next generation reactor for the 21. century was launched, at a time when the first ABWR was still under construction. At the initial stage of this project, development of a 'user friendly' plant design was the primary objective. Thus, the main focus was placed on selecting a design with features promoting ease of operation and maintenance. Meanwhile, the circumstances surrounding the Japanese nuclear power industry changed. The delay of FBRmore » development and the deregulation of the power generation market have significantly boosted the role of light water reactors, and accelerated the need to improve LWR economics. For these reasons, economic competitiveness became an overriding objective in the development of the ABWR-II, with no less importance placed on achieving the highest standards of safety. Several new features were adopted to enhance economic performance: 1700 MW electric output, large fuel bundles, simplified MSIV, large capacity SRV. An output of 1700 MWe was selected for compatibility with the Japanese power grid, and with consideration of current reactor pressure vessel manufacturing capability. Large fuel bundles will contribute to a shortened refueling outage period and a reduction of CRDs. For enhanced safety, the reference design implements a modified ECCS with four subdivision RHR, a diversified power source incorporating gas turbine generators (GTG), an advanced RCIC (ARCIC) and passive heat removal systems consisting of a passive containment cooling system (PCCS) and a passive reactor cooling system (PRCS). The modified ECCS configuration also enables on-line maintenance. While current reactors rely on complex accident management (AM) procedures, implemented by operators in the event of a serious accident, the ABWR-II incorporated severe accident countermeasures at the design stage, to eliminate the need of operator induced AM procedures. The ABWR-II represents one of the most promising and reliable options for the future replacement of older units, without incurring excessive R and D costs. (authors)« less

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koyanagi, Takaaki; Petrie, Christian M.

Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Officemore » of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high heat flux configurations have been assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. These rabbits contain a wide variety of specimens including monolith tubes, SiC fiber SiC matrix (SiC/SiC) composites, duplex specimens (inner composite, outer monolith), and specimens with a variety of metallic or ceramic coatings on the outer surface. The rabbits are targeted for insertion during HFIR cycle 475, which is scheduled for September 2017.« less

Piping Inelastic Fracture Mechanics Analysis.

DTIC Science & Technology

1980-06-30

LOCATIONd THERM4AL SLEEVE REPAIR WELD TYPE 310 STAINLESS TEL C FVICt AREA SPO PCE Fig. 3.1-Duane Arnold recirculation-inlet-nozzle safe end configuration...Environment The most commonly used materials in the LWR piping system are Types 304 and 316 austenitic stainless steel ( cast /wrought). However, for various...seismic and water hammering), the contribu- tion of the residual stress due to the welding plays a very important role in initiation and propagation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

The SPS Concept Development and Evaluation Program includes a comparative assessment. An early first step in the assessment process is the selection and characterization of alternative technologies. This document describes the cost and performance (i.e., technical and environmental) characteristics of six central station energy alternatives: (1) conventional coal-fired powerplant; (2) conventional light water reactor (LWR); (3) combined cycle powerplant with low-Btu gasifiers; (4) liquid metal fast breeder reactor (LMFBR); (5) photovoltaic system without storage; and (6) fusion reactor.

Bulk Shielding Facility quarterly report, April, May and June 1984

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corbett, B.L.; Lance, E.D.

1984-12-01

The BSR operated at an average power level of 1310 kW for 3.8% of the time during April, May, and June. Water-quality control in both the reactor primary and secondary cooling systems was satisfactory. The PCA was used in training startups and was operated on five occasions for the NBS and HEDL recheck of a previous experiment run on the LWR pressure vessel surveillance dosimetry improvement program.

Improvement of INVS Measurement Uncertainty for Pu and U-Pu Nitrate Solution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn Thomas; Menlove, Howard Olsen; Marlow, Johnna Boulds

2017-04-27

In the Tokai Reprocessing Plant (TRP) and the Plutonium Conversion Development Facility (PCDF), a large amount of plutonium nitrate solution which is recovered from light water reactor (LWR) and advanced thermal reactor (ATR), FUGEN are being stored. Since the solution is designated as a direct use material, the periodical inventory verification and flow verification are being conducted by Japan Safeguard Government Office (JSGO) and International Atomic Agency (IAEA).

BNL program in support of LWR degraded-core accident analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ginsberg, T.; Greene, G.A.

1982-01-01

Two major sources of loading on dry watr reactor containments are steam generatin from core debris water thermal interactions and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described in this paper. 8 figures.

3D-profile measurement of advanced semiconductor features by using FIB as reference metrology

NASA Astrophysics Data System (ADS)

Takamasu, Kiyoshi; Iwaki, Yuuki; Takahashi, Satoru; Kawada, Hiroki; Ikota, Masami

2017-03-01

A novel method of sub-nanometer uncertainty for the 3D-profile measurement and LWR (Line Width Roughness) measurement by using FIB (Focused Ion Beam) processing, and TEM (Transmission Electron Microscope) and CD-SEM (Critical Dimension Scanning Electron Microscope) images measurement is proposed to standardize 3D-profile measurement through reference metrology. In this article, we apply the methodology to line profile measurements and roughness measurement of advanced FinFET (Fin-shaped Field-Effect Transistor) features. The FinFET features are horizontally sliced as a thin specimen by FIB micro sampling system. Horizontally images of the specimens are obtained then by a planar TEM. LWR is calculated from the edges positions on TEM images. Moreover, we already have demonstrated the novel on-wafer 3D-profile metrology as "FIB-to-CDSEM method" with FIB slope cut and CD-SEM measuring. Using the method, a few micrometers wide on a wafer is coated and cut by 45-degree slope using FIB tool. Then, the wafer is transferred to CD-SEM to measure the cross section image by top down CD-SEM measurement. We applied FIB-to-CDSEM method to a CMOS image sensor feature. The 45-degree slope cut surface is observed using AFM. The surface profile of slope cut surface and line profiles are analyzed for improving the accuracy of FIB-to-CDSEM method.

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

DOE Office of Scientific and Technical Information (OSTI.GOV)

Phillips, Chris; Willis, William; Carter, Robert

2013-07-01

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for usemore » as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)« less

Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for FY 2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hallbert, Bruce Perry; Thomas, Kenneth David

2015-10-01

Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Camous, F.; Jacq, F.; Chatelard, P.

1997-07-01

In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

Self-aligned blocking integration demonstration for critical sub-30nm pitch Mx level patterning with EUV self-aligned double patterning

NASA Astrophysics Data System (ADS)

Raley, Angélique; Lee, Joe; Smith, Jeffrey T.; Sun, Xinghua; Farrell, Richard A.; Shearer, Jeffrey; Xu, Yongan; Ko, Akiteru; Metz, Andrew W.; Biolsi, Peter; Devilliers, Anton; Arnold, John; Felix, Nelson

2018-04-01

We report a sub-30nm pitch self-aligned double patterning (SADP) integration scheme with EUV lithography coupled with self-aligned block technology (SAB) targeting the back end of line (BEOL) metal line patterning applications for logic nodes beyond 5nm. The integration demonstration is a validation of the scalability of a previously reported flow, which used 193nm immersion SADP targeting a 40nm pitch with the same material sets (Si3N4 mandrel, SiO2 spacer, Spin on carbon, spin on glass). The multi-color integration approach is successfully demonstrated and provides a valuable method to address overlay concerns and more generally edge placement error (EPE) as a whole for advanced process nodes. Unbiased LER/LWR analysis comparison between EUV SADP and 193nm immersion SADP shows that both integrations follow the same trend throughout the process steps. While EUV SADP shows increased LER after mandrel pull, metal hardmask open and dielectric etch compared to 193nm immersion SADP, the final process performance is matched in terms of LWR (1.08nm 3 sigma unbiased) and is only 6% higher than 193nm immersion SADP for average unbiased LER. Using EUV SADP enables almost doubling the line density while keeping most of the remaining processes and films unchanged, and provides a compelling alternative to other multipatterning integrations, which present their own sets of challenges.

High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Odette, G. Robert

Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences thanmore » have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.« less

Convergence studies of deterministic methods for LWR explicit reflector methodology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Canepa, S.; Hursin, M.; Ferroukhi, H.

2013-07-01

The standard approach in modem 3-D core simulators, employed either for steady-state or transient simulations, is to use Albedo coefficients or explicit reflectors at the core axial and radial boundaries. In the latter approach, few-group homogenized nuclear data are a priori produced with lattice transport codes using 2-D reflector models. Recently, the explicit reflector methodology of the deterministic CASMO-4/SIMULATE-3 code system was identified to potentially constitute one of the main sources of errors for core analyses of the Swiss operating LWRs, which are all belonging to GII design. Considering that some of the new GIII designs will rely on verymore » different reflector concepts, a review and assessment of the reflector methodology for various LWR designs appeared as relevant. Therefore, the purpose of this paper is to first recall the concepts of the explicit reflector modelling approach as employed by CASMO/SIMULATE. Then, for selected reflector configurations representative of both GII and GUI designs, a benchmarking of the few-group nuclear data produced with the deterministic lattice code CASMO-4 and its successor CASMO-5, is conducted. On this basis, a convergence study with regards to geometrical requirements when using deterministic methods with 2-D homogenous models is conducted and the effect on the downstream 3-D core analysis accuracy is evaluated for a typical GII deflector design in order to assess the results against available plant measurements. (authors)« less

Low leaching and low LWR photoresist development for 193 nm immersion lithography

NASA Astrophysics Data System (ADS)

Ando, Nobuo; Lee, Youngjoon; Miyagawa, Takayuki; Edamatsu, Kunishige; Takemoto, Ichiki; Yamamoto, Satoshi; Tsuchida, Yoshinobu; Yamamoto, Keiko; Konishi, Shinji; Nakano, Katsushi; Tomoharu, Fujiwara

2006-03-01

With no apparent showstopper in sight, the adoption of ArF immersion technology into device mass production is not a matter of 'if' but a matter of 'when'. As the technology matures at an unprecedented speed, many of initial technical difficulties have been cleared away and the use of a protective layer known as top coat, initially regarded as a must, now becomes optional, for example. Our focus of interest has also sifted to more practical and production related issues such as defect reducing and performance enhancement. Two major types of immersion specific defects, bubbles and a large number of microbridges, were observed and reported elsewhere. The bubble defects seem to decrease by improvement of exposure tool. But the other type defect - probably from residual water spots - is still a problem. We suspect that the acid leaching from resist film causes microbridges. When small water spots were remained on resist surface after exposure, acid catalyst in resist film is leaching into the water spots even though at room temperature. After water from the spot is dried up, acid molecules are condensed at resist film surface. As a result, in the bulk of resist film, acid depletion region is generated underneath the water spot. Acid catalyzed deprotection reaction is not completed at this acid shortage region later in the PEB process resulting in microbridge type defect formation. Similar mechanism was suggested by Kanna et al, they suggested the water evaporation on PEB plate. This hypothesis led us to focus on reducing acid leaching to decrease residual water spot-related defect. This paper reports our leaching measurement results and low leaching photoresist materials satisfying the current leaching requirements outlined by tool makers without topcoat layer. On the other hand, Nakano et al reported that the higher receding contact angle reduced defectivity. The higher receding contact angle is also a key item to increase scan speed. The effort to increase the receding contact angle become very important issue for not only defectivity but also scanner throughput. Some of our experimental results along this line of study are also included in the report. The last topic covered is LWR (Line Width Roughness) as an essential leverage for performance improvement, especially for the smaller CD that immersion lithography is aiming to define. Our recent effort to find effect and working concept to reduce LWR with low leaching materials is also described.

Thermal conductivity of heterogeneous LWR MOX fuels

NASA Astrophysics Data System (ADS)

Staicu, D.; Barker, M.

2013-11-01

It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference was observed for hypostoichiometric fuels, that correspond to the condition used for irradiation. However, if these two formulas are evaluated for O/M = 2.000, the difference between the predictions is negligible (Fig. 1). The difference becomes significant for non-stoichiometric fuels, as shown for O/M = 1.975 in Fig. 1. The microstructure of the FBR fuel with 21.4 wt.% Pu was not described in the paper of Duriez. Taking into account the rigorous experimental methodology used by Duriez (characterisation of the stoichiometry), a possible explanation is an interaction between the plutonium distribution and the stoichiometry. Another parameter having a strong impact on the conductivity is the porosity correction used to obtain the values for 95% TD. This correction is small in the work of Duriez as the samples density is very close to 95% TD. This was also the case for the samples selected by Philipponneau in order to obtain his recommendation. An effect due to differences in the pores shape can also be excluded, as the results are identical for stoichiometric fuels (Fig. 1). Usually the apparent stoichiometry is obtained by heat treatments and checked before and after the measurements, either by XRD or thermogravimetry. However, for non-perfectly homogeneous samples, the gradients in the plutonium distribution induce a non-uniform oxygen distribution, which is difficult to characterise experimentally. It has been proposed by Baron that the deviation from stoichiometry is the main cause for the differences observed between fresh UO2 and MOX [14,15], this effect is quantified in the next section. In the first model ("Model 1"), the effect of Pu is neglected over the entire relevant Pu compositions range (up to 24 wt.% PuO2), and a correlation obtained for non-stoichiometric homogeneous (U,Pu)O2 is used. In the second model ("Model 2", the effect of Pu is supposed to be present at all compositions, with the stoichiometry effect. The thermal conductivity is described by the correlations of Fink [16] for the UO2 matrix, Duriez at low PuO2 contents (coating phase) and of Philipponneau at high PuO2 contents (agglomerates). For the first model, applying a correlation for non-stoichiometric UO2 would be relevant, but such a correlation does not exist for physical reasons in the hypostoichiometric domain. A correlation for homogeneous (U,Pu)O2+x has to be obtained in order to predict the thermal conductivity of heterogeneous MOX fuel, supposing that the effect of Pu can be neglected, i.e. supposing that the thermal conductivities of homogeneous (U,Pu)O2 and UO2 are equal both for stoichiometric and non-stoichiometric fuels. Such a correlation has to be obtained considering reliable data for stoichiometric UO2 and stoichiometry dependence. Different correlations for non-stoichiometric fuels were reviewed [2,8,12,13,15,35,36]. The correlation of Martin [36], available for hyperstoichiometric UO2, was evaluated in the hypostoichiometric domain and the predictions were found to give a stoichiometry dependence very similar to a correlation already proposed [15]. Investigations by Molecular Dynamics [37] have confirmed the almost symmetric effect of the hypo- and hyper-stoichiometry in UO2. We therefore use the correlation of Martin, with however a correction, as for stoichiometric fuels it over predicts the conductivity of stoichiometric UO2 at high temperatures, when compared to the recommendation of Fink [16] (Fig. 4). Analysis has shown that this over-prediction was due to the high temperature term in the correlation of Martin, and that, if this term is removed, the predictions of Martin and Fink were identical for stoichiometric fuels in the temperature range 500-1500 K. The correlation proposed for homogeneous MOX is therefore given by the following equation. k=(0.035 The series and parallel bounds (Eq. (2)) were calculated using the thermal conductivity values given by Eq. (5) for the heterogeneous MOX constituents and the maximum difference between these two bounds is 2% over the considered temperature range. The predictions obtained with the equations of Maxwell-Eucken (Eq. (3)) and Bergman (Eq. (4)) are equal and are in the interval between the series and parallel bounds. This result shows that the use of a sophisticated analytical or numerical model to predict the thermal conductivity is not justified [38]. The model of Maxwell-Eucken [31] was therefore chosen to predict the equivalent thermal conductivity of the heterogeneous MOX.The equivalent thermal conductivity of the stoichiometric heterogeneous MOX with an average PuO2 content of 7.2 wt.% (constituted by a stoichiometric UO2 matrix containing 15 vol.% of (U0.76Pu0.24)O1.975 agglomerates and 55 vol.% of a coating phase of (U0.94Pu0.06)O1.995) was calculated. The results show that the apparent thermal conductivity of the heterogeneous MOX, calculated using homogeneous MOX data (Eq. (5)) with O/M = 2.000, 1.995 and 1.975 (labeled Model 1 in Fig. 4) is not significantly different from the values measured by Duriez. The latter values are also very similar to the thermal conductivity of homogeneous MOX with O/M = 1.995. This simple model shows that the stoichiometry effect is sufficient to explain the lower thermal conductivity of LWR MOX fuel as compared to UO2. The advantage of this simple model is its consistency, as the calculations for the heterogeneous MOX are based on a unique formula for non-stoichiometric homogeneous (U,Pu)O2.In the second model, the effect of the plutonium is taken into account for the coating phase and for the Pu-rich agglomerates. The thermal conductivity is described by the correlations of Fink [16] for UO2.000, of Duriez et al. [2] for (U0.94Pu0.06)O1.995 (coating phase with low PuO2 content) and of Philipponneau [8] for (U0.76Pu0.24)O1.975 (Pu-rich agglomerates with high PuO2 content). The results (labeled Model 2 in Fig. 5) show that the calculated thermal conductivity of the heterogeneous 'stoichiometric' MOX is lower than UO2 and also lower than for stoichiometric MOX as given by Duriez. Therefore taking into account both the heterogeneity in the oxygen distribution and the Pu content leads to an underprediction of the thermal conductivity of heterogeneous MOX. A possible cause for the lower thermal conductivity of unirradiated heterogeneous MOX is therefore the intrinsic fluctuations of the local stoichiometry and only to a lesser extent the perturbation of the heat transfer due to the substitution of the U by Pu atoms in the crystal lattice. This interpretation was already proposed by Baron [14,15]. This assumption is acceptable if the size of the heterogeneities is much smaller that the thickness of the sample. A theoretical criterion for the impact of these parameters, initially proposed by Kerrish [40], was checked experimentally by Lee and Taylor [41] and was found to be too restrictive. The conclusions resulting from the investigations of Lee are that for diffusivity ratios between 1 and 3.5 and volume fractions up to 30%, a ratio of 5 between the sample thickness and inclusions diameter is sufficient. Our heterogeneous samples fulfill this criterion, taking into account that the thermal diffusivity ratio is close to 1 in MOX, that the volume fraction of Pu rich agglomerates is under 30%, and that the agglomerates have a diameter of less than 200 μm compared to the sample (disc) thickness of 1 mm. The most severe requirement that one could use to define a medium behaving like a homogeneous material is that the heat transfer is not affected by the heterogeneities. This is the case for instance if we have a heterogeneous material where the two constituents have equal thermal diffusivity and no thermal resistance is present at the interfaces. This requirement is very close to be perfectly verified for the heterogeneous MOX, as UO2 and (U,Pu)O2 have very close values of the thermal diffusivity. An effect of the sample heterogeneity can also be excluded from the point of view of the location where the thermograms are recorded: the temperature transients on the rear face of the samples are measured with a pyrometer and the system is provided with a lens assembly which enables a 1 mm diameter spot of the sample surface to be focused onto the signal collecting fibre. The thermograms are therefore averaged over a 1 mm diameter surface, which is much larger than the size of the heterogeneities (Pu rich agglomerates with a size of less than 200 μm).The impact of sample thickness on the measured thermal diffusivity was experimentally investigated for the MIMAS MOX with 7.0 wt.% Pu. For this purpose, discs of 0.5, 1, 2, and 3 mm thickness were cut and the thermal diffusivity was measured. The same investigation was done for standard UO2, in order to verify the accuracy of the inverse technique used for the identification of the thermal diffusivity from the thermograms. The inverse technique [39] explicitly takes into account the sample thickness in the calculation of the heat losses. The results for UO2 (Fig. 6) show that the measured thermal diffusivity does not depend on sample thickness, and is in good agreement with the recommendation of Fink [16]. The results for the heterogeneous MOX (Fig. 7) also show no dependence on sample thickness.

A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grabaskas, Dave; Brunett, Acacia J.; Bucknor, Matthew

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release ofmore » radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.« less

MPACT Standard Input User s Manual, Version 2.2.0

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, Benjamin S.; Downar, Thomas; Fitzgerald, Andrew

The MPACT (Michigan PArallel Charactistics based Transport) code is designed to perform high-fidelity light water reactor (LWR) analysis using whole-core pin-resolved neutron transport calculations on modern parallel-computing hardware. The code consists of several libraries which provide the functionality necessary to solve steady-state eigenvalue problems. Several transport capabilities are available within MPACT including both 2-D and 3-D Method of Characteristics (MOC). A three-dimensional whole core solution based on the 2D-1D solution method provides the capability for full core depletion calculations.

FMDP reactor alternative summary report. Volume 1 - existing LWR alternative

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greene, S.R.; Bevard, B.B.

1996-10-07

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] are becoming surplus to national defense needs in both the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. This document summarizes the results of analysis concerned with existing light water reactor plutonium disposition alternatives.

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Velkov, K.; Lizorkin, M.

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chopra, O. K.; Chung, H. M.; Gruber, E. E.

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vesselmore » and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments.« less

Efficacy of whitening oral rinses and dentifrices on color stability of bleached teeth

PubMed Central

Karadas, Muhammet

2015-01-01

Abstract Objective: This study aimed to evaluate the effect of whitening toothpastes and mouthrinses on the color stability of teeth bleached with 16% carbamide peroxide (CP) after immersion in coffee solution. Materials and methods: Specimens obtained from bovine incisors were bleached with 16% CP for 14 days. After bleaching, the specimens were stained in coffee solution for 24 h and randomly divided into eight groups according to the following products (n = 10): distilled water (control group, DW), Scope White mouthrinse (SW), Crest 3D White mouthrinse (CWR), Crest 3D White toothpaste (CWT), Crest 3D White toothpaste and Crest 3D White mouthrinse (CWT + CWR), Listerine Whitening toothpaste (LWT), Listerine Whitening mouthrinse (LWR), and Listerine Whitening mouthrinse and Listerine Whitening toothpaste (LWR + LWT). Color measurements were conducted using a spectrophotometer. The data were assessed by analysis of variance for repeated measures and Tukey’s multiple comparison test (p < 0.05). Results: Immersion in coffee solution after bleaching caused perceptible staining on tooth specimens (ΔE > 3.46). The whitening effect of CWR on teeth stained after bleaching was significantly greater than that in the other groups (p < 0.001). Tooth whitening (ΔE) in each group showed no significant difference from 6 to 12 weeks (p > 0.05). The combination of mouthrinse and toothpaste did not increase the degree of tooth whitening. Conclusion: Whitening mouthrinse and toothpaste had similar effects on the control group in terms of whitening of teeth stained after bleaching. Nevertheless, Crest 3D White mouthrinse produced the greatest recovery whitening effect among all the products tested. PMID:28642898

NRC Licensing Status Summary Report for NGNP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moe, Wayne Leland; Kinsey, James Carl

2014-11-01

The Next Generation Nuclear Plant (NGNP) Project, initiated at Idaho National Laboratory (INL) by the U.S. Department of Energy (DOE) pursuant to provisions of the Energy Policy Act of 2005, is based on research and development activities supported by the Department of Energy Generation IV Nuclear Energy Systems Initiative. The principal objective of the NGNP Project is to support commercialization of high temperature gas-cooled reactor (HTGR) technology. The HTGR is a helium-cooled and graphite moderated reactor that can operate at temperatures much higher than those of conventional light water reactor (LWR) technologies. The NGNP will be licensed for construction andmore » operation by the Nuclear Regulatory Commission (NRC). However, not all elements of current regulations (and their related implementation guidance) can be applied to HTGR technology at this time. Certain policies established during past LWR licensing actions must be realigned to properly accommodate advanced HTGR technology. A strategy for licensing HTGR technology was developed and executed through the cooperative effort of DOE and the NRC through the NGNP Project. The purpose of this report is to provide a snapshot of the current status of the still evolving pre-license application regulatory framework relative to commercial HTGR technology deployment in the U.S. The following discussion focuses on (1) describing what has been accomplished by the NGNP Project up to the time of this report, and (2) providing observations and recommendations concerning actions that remain to be accomplished to enable the safe and timely licensing of a commercial HTGR facility in the U.S.« less

Fingerprinting the type of line edge roughness

NASA Astrophysics Data System (ADS)

Fernández Herrero, A.; Pflüger, M.; Scholze, F.; Soltwisch, V.

2017-06-01

Lamellar gratings are widely used diffractive optical elements and are prototypes of structural elements in integrated electronic circuits. EUV scatterometry is very sensitive to structure details and imperfections, which makes it suitable for the characterization of nanostructured surfaces. As compared to X-ray methods, EUV scattering allows for steeper angles of incidence, which is highly preferable for the investigation of small measurement fields on semiconductor wafers. For the control of the lithographic manufacturing process, a rapid in-line characterization of nanostructures is indispensable. Numerous studies on the determination of regular geometry parameters of lamellar gratings from optical and Extreme Ultraviolet (EUV) scattering also investigated the impact of roughness on the respective results. The challenge is to appropriately model the influence of structure roughness on the diffraction intensities used for the reconstruction of the surface profile. The impact of roughness was already studied analytically but for gratings with a periodic pseudoroughness, because of practical restrictions of the computational domain. Our investigation aims at a better understanding of the scattering caused by line roughness. We designed a set of nine lamellar Si-gratings to be studied by EUV scatterometry. It includes one reference grating with no artificial roughness added, four gratings with a periodic roughness distribution, two with a prevailing line edge roughness (LER) and another two with line width roughness (LWR), and four gratings with a stochastic roughness distribution (two with LER and two with LWR). We show that the type of line roughness has a strong impact on the diffuse scatter angular distribution. Our experimental results are not described well by the present modelling approach based on small, periodically repeated domains.

The Effects of Liquid Propellant Motion on the Attitude Stability of Spin Stabilized Spacecraft

DTIC Science & Technology

1990-03-01

3.733539 3.455776 3.455703 0.314183 STABLE STABLE 80 1 UN 𔃻. Yp Ar 11 12 p IT ho Sigmal Siqma2 SiimaO SignaP SigmaR PTedict Result El’.ATI N MJM9ER (15...8217~ . I lWr 11 12 Ip U 5ig"jl SigW2 5ig,130 5i aP SigmaR Dredict Pesult -iu 79!r, I-lt = 1. 5 1 l, rE’L ) .0.2273 0. nC072 3.141519 39 06 3139.6 5660.7

Nanoparticle photoresist studies for EUV lithography

NASA Astrophysics Data System (ADS)

Kasahara, Kazuki; Xu, Hong; Kosma, Vasiliki; Odent, Jeremy; Giannelis, Emmanuel P.; Ober, Christopher K.

2017-03-01

EUV (extreme ultraviolet) lithography is one of the most promising candidates for next generation lithography. The main challenge for EUV resists is to simultaneously satisfy resolution, LWR (line-width roughness) and sensitivity requirements according to the ITRS roadmap. Though polymer type CAR (chemically amplified resist) is the currently standard photoresist, entirely new resist platforms are required due to the performance targets of smaller process nodes. In this paper, recent progress in nanoparticle photoresists which Cornell University has intensely studied is discussed. Lithography performance, especially scum elimination, improvement studies with the dissolution rate acceleration concept and new metal core applications are described.

Programmed LWR metrology by multi-techniques approach

NASA Astrophysics Data System (ADS)

Reche, Jérôme; Besacier, Maxime; Gergaud, Patrice; Blancquaert, Yoann; Freychet, Guillaume; Labbaye, Thibault

2018-03-01

Nowadays, roughness control presents a huge challenge for the lithography step. For advanced nodes, this morphological aspect reaches the same order of magnitude than the Critical Dimension. Hence, the control of roughness needs an adapted metrology. In this study, specific samples with designed roughness have been manufactured using e-beam lithography. These samples have been characterized with three different methodologies: CD-SEM, OCD and SAXS. The main goal of the project is to compare the capability of each of these techniques in terms of reliability, type of information obtained, time to obtain the measurements and level of maturity for the industry.

SCALE Code System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE depletion with TRITON (T5-DEPL/T6-DEPL),• CE sensitivity/uncertainty analysis with TSUNAMI-3D,• Simplified and efficient LWR lattice physics with Polaris,• Large scale detailed spent fuel characterization with ORIGAMI and ORIGAMI Automator,• Advanced fission source convergence acceleration capabilities with Sourcerer,• Nuclear data library generation with AMPX, and• Integrated user interface with Fulcrum.Enhanced capabilities include:• Accurate and efficient CE Monte Carlo methods for eigenvalue and fixed source calculations,• Improved MG resonance self-shielding methodologies and data,• Resonance self-shielding with modernized and efficient XSProc integrated into most sequences,• Accelerated calculations with TRITON/NEWT (generally 4x faster than SCALE 6.1),• Spent fuel characterization with 1470 new reactor-specific libraries for ORIGEN,• Modernization of ORIGEN (Chebyshev Rational Approximation Method [CRAM] solver, API for high-performance depletion, new keyword input format)• Extension of the maximum mixture number to values well beyond the previous limit of 2147 to ~2 billion,• Nuclear data formats enabling the use of more than 999 energy groups,• Updated standard composition library to provide more accurate use of natural abundances, andvi• Numerous other enhancements for improved usability and stability.« less

Delayed access of low body weight-selected chicks to food at hatch is associated with up-regulated pancreatic glucagon and glucose transporter gene expression.

PubMed

Parker, Grace A; Sumners, Lindsay H; Zhao, Xiaoling; Honaker, Christa F; Siegel, Paul B; Cline, Mark A; Gilbert, Elizabeth R

2015-11-01

Chickens selected for low (LWS) and high (HWS) juvenile body weight (BW) for 55 generations differ in BW by 10-fold at selection age. High (HWR) and low (LWR) body weight-relaxed lines have been random-bred since the 46th generation. Our objective was to evaluate the developmental and nutritional regulation of pancreatic mRNA abundance of pancreatic and duodenal homeobox 1 (PDX1), preproinsulin (PPI), preproglucagon (PPG), and glucose transporter 2 (GLUT2). At day of hatch (DOH) and days 1, 3, 7, and 15 (D1, 3, 7 and 15, respectively), pancreas was collected and real time PCR was performed in Experiment 1. In Experiment 2, HWS and LWS were fed or delayed access to food for 72 h post-hatch, and pancreas collected at D15. There was an interaction of line and age for GLUT2 (P=0.001), PPI (P<0.0001), PPG (P=0.034), and PDX1 (P<0.0001). Expression was greater in chicks from LWR and LWS than HWR and HWS. There was an interaction of line and nutrition on PPG (P<0.0001) and GLUT2 (P=0.001) mRNA, where expression was similar among chicks that were fed but greater in LWS than HWS when chicks were delayed access to food. Thus, the first two weeks is important for maturation of pancreatic endocrine function. Long-term selection for BW is associated with differences in pancreas development, and delaying access to food at hatch may have persisting effects on glucose regulatory function. Copyright © 2015 Elsevier Inc. All rights reserved.

Production and Testing of the VITAMIN-B7 Fine-Group and BUGLE-B7 Broad-Group Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Risner, J. M.; Wiarda, D.; Dunn, M. E.

2011-09-30

New coupled neutron-gamma cross-section libraries have been developed for use in light water reactor (LWR) shielding applications, including pressure vessel dosimetry calculations. The libraries, which were generated using Evaluated Nuclear Data File/B Version VII Release 0 (ENDF/B-VII.0), use the same fine-group and broad-group energy structures as the VITAMIN-B6 and BUGLE-96 libraries. The processing methodology used to generate both libraries is based on the methods used to develop VITAMIN-B6 and BUGLE-96 and is consistent with ANSI/ANS 6.1.2. The ENDF data were first processed into the fine-group pseudo-problem-independent VITAMIN-B7 library and then collapsed into the broad-group BUGLE-B7 library. The VITAMIN-B7 library containsmore » data for 391 nuclides. This represents a significant increase compared to the VITAMIN-B6 library, which contained data for 120 nuclides. The BUGLE-B7 library contains data for the same nuclides as BUGLE-96, and maintains the same numeric IDs for those nuclides. The broad-group data includes nuclides which are infinitely dilute and group collapsed using a concrete weighting spectrum, as well as nuclides which are self-shielded and group collapsed using weighting spectra representative of important regions of LWRs. The verification and validation of the new libraries includes a set of critical benchmark experiments, a set of regression tests that are used to evaluate multigroup crosssection libraries in the SCALE code system, and three pressure vessel dosimetry benchmarks. Results of these tests confirm that the new libraries are appropriate for use in LWR shielding analyses and meet the requirements of Regulatory Guide 1.190.« less

Evaluation of flaws in carbon steel piping. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zahoor, A.; Gamble, R.M.; Mehta, H.S.

1986-10-01

The objective of this program was to develop flaw evaluation procedures and allowable flaw sizes for ferritic piping used in light water reactor (LWR) power generation facilities. The program results provide relevant ASME Code groups with the information necessary to define flaw evaluation procedures, allowable flaw sizes, and their associated bases for Section XI of the code. Because there are several possible flaw-related failure modes for ferritic piping over the LWR operating temperature range, three analysis methods were employed to develop the evaluation procedures. These include limit load analysis for plastic collapse, elastic plastic fracture mechanics (EPFM) analysis for ductilemore » tearing, and linear elastic fracture mechanics (LEFM) analysis for non ductile crack extension. To ensure the appropriate analysis method is used in an evaluation, a step by step procedure also is provided to identify the relevant acceptance standard or procedure on a case by case basis. The tensile strength and toughness properties required to complete the flaw evaluation for any of the three analysis methods are included in the evaluation procedure. The flaw evaluation standards are provided in tabular form for the plastic collapse and ductile tearing modes, where the allowable part through flaw depth is defined as a function of load and flaw length. For non ductile crack extension, linear elastic fracture mechanics analysis methods, similar to those in Appendix A of Section XI, are defined. Evaluation flaw sizes and procedures are developed for both longitudinal and circumferential flaw orientations and normal/upset and emergency/faulted operating conditions. The tables are based on margins on load of 2.77 and 1.39 for circumferential flaws and 3.0 and 1.5 for longitudinal flaws for normal/upset and emergency/faulted conditions, respectively.« less

Intensification of oxidation capacity using chloroalkanes as additives in hydrodynamic and acoustic cavitation reactors.

PubMed

Chakinala, Anand G; Gogate, Parag R; Chand, Rashmi; Bremner, David H; Molina, Raúl; Burgess, Arthur E

2008-03-01

The effect of the presence and absence of the chloroalkanes, dichloromethane (CH(2)Cl(2)), chloroform (CHCl(3)) and carbon tetrachloride (CCl(4)) on the extent of oxidation of aqueous I(-) to I(3)(-) has been investigated in (a) a liquid whistle reactor (LWR) generating hydrodynamic cavitation and (b) an ultrasonic probe, which produces acoustic cavitation. The aim has been to examine the intensification achieved in the extent of oxidation due to the generation of additional free radicals/oxidants in the reactor as a result of the presence of chloroalkanes. It has been observed that the extent of increase in the oxidation reaction is strongly dependent on the applied pressure in the case of the LWR. Also, higher volumes of the chloroalkanes favour the intensification and the order of effectiveness is CCl(4)>CHCl(3)>CH(2)Cl(2). However, the results with the ultrasonic probe suggest that an optimum concentration of CH(2)Cl(2) or CHCl(3) exists beyond which there is little increase in the extent of observed intensification. For CCl(4), however, no such optimum concentration was observed and the extent of increase in the rates of oxidation reaction rose with the amount of CCl(4) added. Stage wise addition of the chloroalkanes was found to give marginally better results in the case of the ultrasonic probe as compared to bulk addition at the start of the run. Although CCl(4) is the most effective, its toxicity and carcinogenicity may mean that CH(2)Cl(2) and CHCl(3) offer a safer viable alternative and the present work should be useful in establishing the amount of chloroalkanes required for obtaining a suitable degree of intensification.

Investment in different sized SMRs: Economic evaluation of stochastic scenarios by INCAS code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barenghi, S.; Boarin, S.; Ricotti, M. E.

2012-07-01

Small Modular LWR concepts are being developed and proposed to investors worldwide. They capitalize on operating track record of GEN II LWR, while introducing innovative design enhancements allowed by smaller size and additional benefits from the higher degree of modularization and from deployment of multiple units on the same site. (i.e. 'Economy of Multiple' paradigm) Nevertheless Small Modular Reactors pay for a dis-economy of scale that represents a relevant penalty on a capital intensive investment. Investors in the nuclear power generation industry face a very high financial risk, due to high capital commitment and exceptionally long pay-back time. Investment riskmore » arise from uncertainty that affects scenario conditions over such a long time horizon. Risk aversion is increased by current adverse conditions of financial markets and general economic downturn, as is the case nowadays. This work investigates both the investment profitability and risk of alternative investments in a single Large Reactor or in multiple SMR of different sizes drawing information from project's Internal Rate of Return stochastic distribution. multiple SMR deployment on a single site with total power installed, equivalent to a single LR. Uncertain scenario conditions and stochastic input assumptions are included in the analysis, representing investment uncertainty and risk. Results show that, despite the combination of much larger number of stochastic variables in SMR fleets, uncertainty of project profitability is not increased, as compared to LR: SMR have features able to smooth IRR variance and control investment risk. Despite dis-economy of scale, SMR represent a limited capital commitment and a scalable investment option that meet investors' interest, even in developed and mature markets, that are traditional marketplace for LR. (authors)« less

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.

The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 –7 dpa/s and 390 °C) and thenmore » use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 –8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni 3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

DOE PAGES

Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.; ...

2017-03-11

The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 –7 dpa/s and 390 °C) and thenmore » use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 –8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni 3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

Material Issues of Blanket Systems for Fusion Reactors - Compatibility with Cooling Water -

NASA Astrophysics Data System (ADS)

Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

Heating Parameter Estimation Using Coaxial Thermocouple Gages in Wind Tunnel Test Articles.

DTIC Science & Technology

1984-12-01

Attack a Emissivity G Parameter Vector Pn Measurement Vector at nth Time Point p Density 0 Stefan-Boltzmann Constant 6 Transition Matrix APc Scaling...for. The radiation is modeled using the Stefan-Boltzmann Law, q = 60(U 4 - U, 4 ) (A-9) where 8 radiative emissivity a Stefan-Bol tzmann constant U...w00 I- 000 0 0111c :0 i zZ Z-4lwr I- E . - t J K - IL HHO "W 6i 0WZWZWO&000OW *0 . 0 - .- - -4 4 1"- 1 Lii w LiiU Li LI Li Lij Liw w ~ o 0 0wm ~wW6~w d

Thermal energy storage and transport

NASA Technical Reports Server (NTRS)

Hausz, W.

1980-01-01

The extraction of thermal energy from large LWR and coal fired plants for long distance transport to industrial and residential/commercial users is analyzed. Transport of thermal energy as high temperature water is shown to be considerably cheaper than transport as steam, hot oil, or molten salt over a wide temperature range. The delivered heat is competitive with user-generated heat from oil, coal, or electrode boilers at distances well over 50 km when the pipeline operates at high capacity factor. Results indicate that thermal energy storage makes meeting of even very low capacity factor heat demands economic and feasible and gives the utility flexibility to meet coincident electricity and heat demands effectively.

Status on the Component Models Developed in the Modelica Framework: High-Temperature Steam Electrolysis Plant & Gas Turbine Power Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Suk Kim, Jong; McKellar, Michael; Bragg-Sitton, Shannon M.

This report has been prepared as part of an effort to design and build a modeling and simulation (M&S) framework to assess the economic viability of a nuclear-renewable hybrid energy system (N-R HES). In order to facilitate dynamic M&S of such an integrated system, research groups in multiple national laboratories have been developing various subsystems as dynamic physics-based components using the Modelica programming language. In fiscal year (FY) 2015, Idaho National Laboratory (INL) performed a dynamic analysis of two region-specific N-R HES configurations, including the gas-to-liquid (natural gas to Fischer-Tropsch synthetic fuel) and brackish water reverse osmosis desalination plants asmore » industrial processes. In FY 2016, INL has developed two additional subsystems in the Modelica framework: a high-temperature steam electrolysis (HTSE) plant and a gas turbine power plant (GTPP). HTSE has been proposed as a high priority industrial process to be integrated with a light water reactor (LWR) in an N-R HES. This integrated energy system would be capable of dynamically apportioning thermal and electrical energy (1) to provide responsive generation to the power grid and (2) to produce alternative industrial products (i.e., hydrogen and oxygen) without generating any greenhouse gases. A dynamic performance analysis of the LWR/HTSE integration case was carried out to evaluate the technical feasibility (load-following capability) and safety of such a system operating under highly variable conditions requiring flexible output. To support the dynamic analysis, the detailed dynamic model and control design of the HTSE process, which employs solid oxide electrolysis cells, have been developed to predict the process behavior over a large range of operating conditions. As first-generation N-R HES technology will be based on LWRs, which provide thermal energy at a relatively low temperature, complementary temperature-boosting technology was suggested for integration with the HTSE process that requires higher temperature input. Simulation results involving several case studies show that the suggested control scheme could maintain the controlled variables (including the steam utilization factor, cathode stream inlet composition, and temperatures of the process streams at various locations) within desired limits under various plant operating conditions. The results also indicate that the proposed HTSE plant could provide operational flexibility to participate in energy management at the utility scale by dynamically optimizing the use of excess plant capacity within an N-R HES. A natural-gas fired GTPP has been proposed as a secondary energy supply to be included in an N-R HES. This auxiliary generator could be used to cover rapid dynamics in grid demand that cannot be met by the remainder of the N-R HES. To evaluate the operability and controllability of the proposed process during transients between load (demand) levels, the dynamic model and control design were developed. Special attention was given to the design of feedback controllers to regulate the power frequency, and exhaust gas and turbine firing temperatures. Several case studies were performed to investigate the system responses to the major disturbance (power load demand) in such a control system. The simulation results show that the performance of the proposed control strategies was satisfactory under each test when the GTPP experienced high rapid variations in the load.« less

Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomas, Ken; Lawrie, Sean; Hart, Adam

The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systemsmore » of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on demonstrating actual cost reductions that can be credited to budgets and thereby truly reduce O&M or capital costs. Technology enhancements, while enhancing work methods and making work more efficient, often fail to eliminate workload such that it changes overall staffing and material cost requirements. It is critical to demonstrate cost reductions or impacts on non-cost performance objectives in order for the business case to justify investment by nuclear operators. The Business Case Methodology (BCM) addresses the “benefit” side of the analysis—as opposed to the cost side—and how the organization evaluates discretionary projects (net present value (NPV), accounting effects of taxes, discount rates, etc.). The cost and analysis side is not particularly difficult for the organization and can usually be determined with a fair amount of precision (not withstanding implementation project cost overruns). It is in determining the "benefits" side of the analysis that utilities have more difficulty in technology projects and that is the focus of this methodology.« less

VERAIn

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simunovic, Srdjan

2015-02-16

CASL's modeling and simulation technology, the Virtual Environment for Reactor Applications (VERA), incorporates coupled physics and science-based models, state-of-the-art numerical methods, modern computational science, integrated uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs), single-effect experiments, and integral tests. The computational simulation component of VERA is the VERA Core Simulator (VERA-CS). The core simulator is the specific collection of multi-physics computer codes used to model and deplete a LWR core over multiple cycles. The core simulator has a single common input file that drives all of the different physics codes. The parser code, VERAIn, converts VERAmore » Input into an XML file that is used as input to different VERA codes.« less

Analysis of social optimum for staggered shifts in a single-entry traffic corridor with no late arrivals

NASA Astrophysics Data System (ADS)

Li, Chuan-Yao; Huang, Hai-Jun; Tang, Tie-Qiao

2017-03-01

This paper investigates the traffic flow dynamics under the social optimum (SO) principle in a single-entry traffic corridor with staggered shifts from the analytical and numerical perspectives. The LWR (Lighthill-Whitham and Richards) model and the Greenshield's velocity-density function are utilized to describe the dynamic properties of traffic flow. The closed-form SO solution is analytically derived and some numerical examples are used to further testify the analytical solution. The optimum proportion of the numbers of commuters with different desired arrival times is further discussed, where the analytical and numerical results both indicate that the cumulative outflow curve under the SO principle is piecewise smooth.

Hybrid method for numerical modelling of LWR coolant chemistry

NASA Astrophysics Data System (ADS)

Swiatla-Wojcik, Dorota

2016-10-01

A comprehensive approach is proposed to model radiation chemistry of the cooling water under exposure to neutron and gamma radiation at 300 °C. It covers diffusion-kinetic processes in radiation tracks and secondary reactions in the bulk coolant. Steady-state concentrations of the radiolytic products have been assessed based on the simulated time dependent concentration profiles. The principal reactions contributing to the formation of H2, O2 and H2O2 were indicated. Simulation was carried out depending on the amount of extra hydrogen dissolved in the coolant to reduce concentration of corrosive agents. High sensitivity to the rate of reaction H+H2O=OH+H2 is shown and discussed.

Some less conventional options for plutonium disposal

NASA Astrophysics Data System (ADS)

Stoll, Dr. Wolfgang, Prof.

2000-07-01

Disposition of weapons Pu (W-Pu) aims at the replacement of military access restrictions by inherent longlasting technical barriers to make the return into the weapons state difficult and not rewarding anymore. At the time of the NAS-study in 1994, two ways were perceived to be mature and selected: Fissioning of W-Pu as LWR-MOX and the disposal in a vitrified radionuclide-spiked form.1 Both options since have been questioned for equality, met different acceptance at both superpowers and showed slow progress. A criterion to measure disarmament would be the amount of W-Pu in the different proliferation resistant forms, multiplied by the effort needed for each form to return to weapons quality.

United States Department of Energy severe accident research following the Fukushima Daiichi accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.; Corradini, M.; Rempe, J.

The U.S. Department of Energy (DOE) has played a major role in the U.S. response to the events at Fukushima Daiichi. During the first several weeks following the accident, U.S. assistance efforts were guided by results from a significant and diverse set of analyses. In the months that followed, a coordinated analysis activity aimed at gaining a more thorough understanding of the accident sequence was completed using laboratory-developed, system-level best-estimate accident analysis codes, while a parallel analysis was conducted by U.S. industry. A comparison of predictions for Unit 1 from these two studies indicated significant differences between MAAP and MELCORmore » results for key plant parameters, such as in-core hydrogen production. On that basis, a crosswalk was completed to determine the key modeling variations that led to these differences. In parallel with these activities, it became clear that there was a need to perform a technology gap evaluation on accident-tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist given the current state of light water reactor (LWR) severe accident research and augmented by insights from Fukushima. In addition, there is growing international recognition that data from Fukushima could significantly reduce uncertainties related to severe accident progression, particularly for boiling water reactors. On these bases, a group of U. S. experts in LWR safety and plant operations was convened by the DOE Office of Nuclear Energy (DOE-NE) to complete technology gap analysis and Fukushima forensics data needs identification activities. The results from these activities were used as the basis for refining DOE-NE's severe accident research and development (R&D) plan. Finally, this paper provides a high-level review of DOE-sponsored R&D efforts in these areas, including planned activities on accident-tolerant components and accident analysis methods.« less

United States Department of Energy severe accident research following the Fukushima Daiichi accidents

DOE PAGES

Farmer, M. T.; Corradini, M.; Rempe, J.; ...

2016-11-02

The U.S. Department of Energy (DOE) has played a major role in the U.S. response to the events at Fukushima Daiichi. During the first several weeks following the accident, U.S. assistance efforts were guided by results from a significant and diverse set of analyses. In the months that followed, a coordinated analysis activity aimed at gaining a more thorough understanding of the accident sequence was completed using laboratory-developed, system-level best-estimate accident analysis codes, while a parallel analysis was conducted by U.S. industry. A comparison of predictions for Unit 1 from these two studies indicated significant differences between MAAP and MELCORmore » results for key plant parameters, such as in-core hydrogen production. On that basis, a crosswalk was completed to determine the key modeling variations that led to these differences. In parallel with these activities, it became clear that there was a need to perform a technology gap evaluation on accident-tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist given the current state of light water reactor (LWR) severe accident research and augmented by insights from Fukushima. In addition, there is growing international recognition that data from Fukushima could significantly reduce uncertainties related to severe accident progression, particularly for boiling water reactors. On these bases, a group of U. S. experts in LWR safety and plant operations was convened by the DOE Office of Nuclear Energy (DOE-NE) to complete technology gap analysis and Fukushima forensics data needs identification activities. The results from these activities were used as the basis for refining DOE-NE's severe accident research and development (R&D) plan. Finally, this paper provides a high-level review of DOE-sponsored R&D efforts in these areas, including planned activities on accident-tolerant components and accident analysis methods.« less

Roughness and uniformity improvements on self-aligned quadruple patterning technique for 10nm node and beyond by wafer stress engineering

NASA Astrophysics Data System (ADS)

Liu, Eric; Ko, Akiteru; O'Meara, David; Mohanty, Nihar; Franke, Elliott; Pillai, Karthik; Biolsi, Peter

2017-05-01

Dimension shrinkage has been a major driving force in the development of integrated circuit processing over a number of decades. The Self-Aligned Quadruple Patterning (SAQP) technique is widely adapted for sub-10nm node in order to achieve the desired feature dimensions. This technique provides theoretical feasibility of multiple pitch-halving from 193nm immersion lithography by using various pattern transferring steps. The major concept of this approach is to a create spacer defined self-aligned pattern by using single lithography print. By repeating the process steps, double, quadruple, or octuple are possible to be achieved theoretically. In these small architectures, line roughness control becomes extremely important since it may contribute to a significant portion of process and device performance variations. In addition, the complexity of SAQP in terms of processing flow makes the roughness improvement indirective and ineffective. It is necessary to discover a new approach in order to improve the roughness in the current SAQP technique. In this presentation, we demonstrate a novel method to improve line roughness performances on 30nm pitch SAQP flow. We discover that the line roughness performance is strongly related to stress management. By selecting different stress level of film to be deposited onto the substrate, we can manipulate the roughness performance in line and space patterns. In addition, the impact of curvature change by applied film stress to SAQP line roughness performance is also studied. No significant correlation is found between wafer curvature and line roughness performance. We will discuss in details the step-by-step physical performances for each processing step in terms of critical dimension (CD)/ critical dimension uniformity (CDU)/line width roughness (LWR)/line edge roughness (LER). Finally, we summarize the process needed to reach the full wafer performance targets of LWR/LER in 1.07nm/1.13nm on 30nm pitch line and space pattern.

Body composition and power performance improved after weight reduction in male athletes without hampering hormonal balance.

PubMed

Huovinen, Heikki T; Hulmi, Juha J; Isolehto, Juha; Kyröläinen, Heikki; Puurtinen, Risto; Karila, Tuomo; Mackala, Krzysztof; Mero, Antti A

2015-01-01

The aim of this study was to investigate the effects of a 4-week weight reduction period with high protein and reduced carbohydrate intake on body composition, explosive power, speed, serum hormones, and acid-base balance in male track and field jumpers and sprinters. Eight participants were assigned to a high weight reduction group (HWR; energy restriction 750 kcal·d) and 7 to a low weight reduction group (LWR; energy restriction 300 kcal·d). Energy and carbohydrate intake decreased significantly (p ≤ 0.05) only in HWR by 740 ± 330 kcal·d and 130 ± 29 g·d, respectively. Furthermore, total body mass and fat mass decreased (p ≤ 0.05) only in HWR by 2.2 ± 1.0 kg and 1.7 ± 1.6 kg, respectively. Fat-free mass (FFM), serum testosterone, cortisol, and sex hormone-binding globulin did not change significantly. Ca ion and pH decreased (p ≤ 0.05) only in HWR (3.1 ± 2.8% and 0.8 ± 0.8%, respectively), whereas (Equation is included in full-text article.)declined (p ≤ 0.05) in both groups by 19.3 ± 6.2% in HWR and by 13.1 ± 8.5% in LWR. The countermovement jump and 20-m sprint time improved consistently (p ≤ 0.05) only in HWR, by 2.6 ± 2.5 cm and 0.04 ± 0.04 seconds, respectively. Finally, athletes with a fat percentage of 10% or more at the baseline were able to preserve FFM. In conclusion, altered acid-base balance but improved weight-bearing power performance was observed without negative consequences on serum hormones and FFM after a 4-week weight reduction of 0.5 kg·wk achieved by reduced carbohydrate but maintained high protein intake.

First Conclusions of the WPEC/Subgroup-22 Nuclear Data for Improved LEU-LWR Reactivity Predictions

NASA Astrophysics Data System (ADS)

Courcelle, Arnaud

2005-05-01

This paper is a summary of a collective work in the framework of the Working Party in International Nuclear Data Evaluation and Co-operation (WPEC) to investigate the reasons for systematic reactivity underprediction of thermal LEU-LWR (Low-Enriched Uranium, Light-Water Reactor). This keff underprediction (≈ -500 pcm) is observed with the most recent nuclear data libraries (ENDF/B-VI.8, JENDL3.3 and JEFF3.0) This report reviews the evaluation work performed at several laboratories [Oak Ridge National Laboratory (ORNL), Los Alamos National Laboratory (LANL), Commissariat a l'énergie atomique de Bruyeres-Le-Chatel (CEA-BRC), International Atomic Energy Agency (IAEA)] as well as the integral tests (mainly at LANL, Knoll Atomic Power Laboratory (KAPL), Bettis Atomic Power Laboratory (BAPL), Nuclear Research and Consultancy Group NRG-Petten, CEA and IAEA) of the successive versions of the new evaluated files. The present status of the work can be summarized as follows: • Improved evaluations of 238U inelastic data proposed by LANL and CEA-BRC were tested against integral benchmarks and partially improve the reactivity prediction. • The thermal capture cross-section of 238U has been revised, and a new evaluation of 238U resonance parameters, up to 20 keV, is in progress at ORNL. Integral tests have ensured that the modifications of 238U capture cross-section in the thermal and resolved range were still compatible with 238U integral measurements (238U capture rate ratios measured in critical facilities and 239Pu build-up prediction in a depleted pressurized water reactor (PWR) assembly). It is demonstrated that the combination of the new inelastic data (LANL or BRC) with the preliminary ORNL resonance parameter set gives a good correction of the reactivity under-estimation. The provisional conclusions of this collective work are expected to contribute toward the improvement of the future versions of nuclear data libraries.

High-velocity interstellar gas in the line of sight to the Wolf-Rayet star HD 50896

NASA Technical Reports Server (NTRS)

Nichols-Bohlin, J.; Fesen, R. A.

1986-01-01

The large shell of interstellar gas (IG) discovered toward HD 50896 by Heckathorn and Fesen (1984) is characterized on the basis of high-dispersion IUE SWP and LWR spectra of 19 objects located within 4 deg of HD 50896 (but outside the optical ring nebula S308) at distances 0.6-2.9 kpc (compared to 1.5 kpc for HD 50896). The IG is found to have two components (at velocities -80 and -125 km/s), diameter 90 pc or greater, and distance 1.0 + or - 0.2 kpc, demonstrating that it is not related to HD 50896 and suggesting that it is a highly evolved supernova remnant associated with cluster Cr 121.

Coater/developer based techniques to improve high-resolution EUV patterning defectivity

NASA Astrophysics Data System (ADS)

Hontake, Koichi; Huli, Lior; Lemley, Corey; Hetzer, Dave; Liu, Eric; Ko, Akiteru; Kawakami, Shinichiro; Shimoaoki, Takeshi; Hashimoto, Yusaku; Tanaka, Koichiro; Petrillo, Karen; Meli, Luciana; De Silva, Anuja; Xu, Yongan; Felix, Nelson; Johnson, Richard; Murray, Cody; Hubbard, Alex

2017-10-01

Extreme ultraviolet lithography (EUVL) technology is one of the leading candidates under consideration for enabling the next generation of devices, for 7nm node and beyond. As the focus shifts to driving down the 'effective' k1 factor and enabling the full scaling entitlement of EUV patterning, new techniques and methods must be developed to reduce the overall defectivity, mitigate pattern collapse, and eliminate film-related defects. In addition, CD uniformity and LWR/LER must be improved in terms of patterning performance. Tokyo Electron Limited (TEL™) and IBM Corporation are continuously developing manufacturing quality processes for EUV. In this paper, we review the ongoing progress in coater/developer based processes (coating, developing, baking) that are required to enable EUV patterning.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adams, S.R.

A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different frommore » the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.« less

Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, Mitchell T.; Bunt, R.; Corradini, M.

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affectmore » reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).« less

Groundwars Analysis of the Laser Warning Receiver (LWR). Phase 2

DTIC Science & Technology

1992-06-02

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A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR safety analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

An advanced system analysis tool is being developed for advanced reactor safety analysis. This paper describes the underlying physics and numerical models used in the code, including the governing equations, the stabilization schemes, the high-order spatial and temporal discretization schemes, and the Jacobian Free Newton Krylov solution method. The effects of the spatial and temporal discretization schemes are investigated. Additionally, a series of verification test problems are presented to confirm the high-order schemes. Furthermore, it is demonstrated that the developed system thermal-hydraulics model can be strictly verified with the theoretical convergence rates, and that it performs very well for amore » wide range of flow problems with high accuracy, efficiency, and minimal numerical diffusions.« less

A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR safety analyses

DOE PAGES

Hu, Rui

2016-11-19

An advanced system analysis tool is being developed for advanced reactor safety analysis. This paper describes the underlying physics and numerical models used in the code, including the governing equations, the stabilization schemes, the high-order spatial and temporal discretization schemes, and the Jacobian Free Newton Krylov solution method. The effects of the spatial and temporal discretization schemes are investigated. Additionally, a series of verification test problems are presented to confirm the high-order schemes. Furthermore, it is demonstrated that the developed system thermal-hydraulics model can be strictly verified with the theoretical convergence rates, and that it performs very well for amore » wide range of flow problems with high accuracy, efficiency, and minimal numerical diffusions.« less

Determination of riverbank erosion probability using Locally Weighted Logistic Regression

NASA Astrophysics Data System (ADS)

Ioannidou, Elena; Flori, Aikaterini; Varouchakis, Emmanouil A.; Giannakis, Georgios; Vozinaki, Anthi Eirini K.; Karatzas, George P.; Nikolaidis, Nikolaos

2015-04-01

Riverbank erosion is a natural geomorphologic process that affects the fluvial environment. The most important issue concerning riverbank erosion is the identification of the vulnerable locations. An alternative to the usual hydrodynamic models to predict vulnerable locations is to quantify the probability of erosion occurrence. This can be achieved by identifying the underlying relations between riverbank erosion and the geomorphological or hydrological variables that prevent or stimulate erosion. Thus, riverbank erosion can be determined by a regression model using independent variables that are considered to affect the erosion process. The impact of such variables may vary spatially, therefore, a non-stationary regression model is preferred instead of a stationary equivalent. Locally Weighted Regression (LWR) is proposed as a suitable choice. This method can be extended to predict the binary presence or absence of erosion based on a series of independent local variables by using the logistic regression model. It is referred to as Locally Weighted Logistic Regression (LWLR). Logistic regression is a type of regression analysis used for predicting the outcome of a categorical dependent variable (e.g. binary response) based on one or more predictor variables. The method can be combined with LWR to assign weights to local independent variables of the dependent one. LWR allows model parameters to vary over space in order to reflect spatial heterogeneity. The probabilities of the possible outcomes are modelled as a function of the independent variables using a logistic function. Logistic regression measures the relationship between a categorical dependent variable and, usually, one or several continuous independent variables by converting the dependent variable to probability scores. Then, a logistic regression is formed, which predicts success or failure of a given binary variable (e.g. erosion presence or absence) for any value of the independent variables. The erosion occurrence probability can be calculated in conjunction with the model deviance regarding the independent variables tested. The most straightforward measure for goodness of fit is the G statistic. It is a simple and effective way to study and evaluate the Logistic Regression model efficiency and the reliability of each independent variable. The developed statistical model is applied to the Koiliaris River Basin on the island of Crete, Greece. Two datasets of river bank slope, river cross-section width and indications of erosion were available for the analysis (12 and 8 locations). Two different types of spatial dependence functions, exponential and tricubic, were examined to determine the local spatial dependence of the independent variables at the measurement locations. The results show a significant improvement when the tricubic function is applied as the erosion probability is accurately predicted at all eight validation locations. Results for the model deviance show that cross-section width is more important than bank slope in the estimation of erosion probability along the Koiliaris riverbanks. The proposed statistical model is a useful tool that quantifies the erosion probability along the riverbanks and can be used to assist managing erosion and flooding events. Acknowledgements This work is part of an on-going THALES project (CYBERSENSORS - High Frequency Monitoring System for Integrated Water Resources Management of Rivers). The project has been co-financed by the European Union (European Social Fund - ESF) and Greek national funds through the Operational Program "Education and Lifelong Learning" of the National Strategic Reference Framework (NSRF) - Research Funding Program: THALES. Investing in knowledge society through the European Social Fund.

Establishment and assessment of code scaling capability

NASA Astrophysics Data System (ADS)

Lim, Jaehyok

In this thesis, a method for using RELAP5/MOD3.3 (Patch03) code models is described to establish and assess the code scaling capability and to corroborate the scaling methodology that has been used in the design of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. It was sponsored by the United States Nuclear Regulatory Commission (USNRC) under the program "PUMA ESBWR Tests". PUMA-E facility was built for the USNRC to obtain data on the performance of the passive safety systems of the General Electric (GE) Nuclear Energy Economic Simplified Boiling Water Reactor (ESBWR). Similarities between the prototype plant and the scaled-down test facility were investigated for a Gravity-Driven Cooling System (GDCS) Drain Line Break (GDLB). This thesis presents the results of the GDLB test, i.e., the GDLB test with one Isolation Condenser System (ICS) unit disabled. The test is a hypothetical multi-failure small break loss of coolant (SB LOCA) accident scenario in the ESBWR. The test results indicated that the blow-down phase, Automatic Depressurization System (ADS) actuation, and GDCS injection processes occurred as expected. The GDCS as an emergency core cooling system provided adequate supply of water to keep the Reactor Pressure Vessel (RPV) coolant level well above the Top of Active Fuel (TAF) during the entire GDLB transient. The long-term cooling phase, which is governed by the Passive Containment Cooling System (PCCS) condensation, kept the reactor containment system that is composed of Drywell (DW) and Wetwell (WW) below the design pressure of 414 kPa (60 psia). In addition, the ICS continued participating in heat removal during the long-term cooling phase. A general Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach was discussed in detail relative to safety analyses of Light Water Reactor (LWR). The major components of the CSAU methodology that were highlighted particularly focused on the scaling issues of experiments and models and their applicability to the nuclear power plant transient and accidents. The major thermal-hydraulic phenomena to be analyzed were identified and the predictive models adopted in RELAP5/MOD3.3 (Patch03) code were briefly reviewed.

Environmentally assisted cracking in light water reactors : semiannual report, July 2000 - December 2000.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chopra, O. K.; Chung, H. M.; Gruber, E. E.

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from July 2000 to December 2000. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. The fatigue strain-vs.-life data are summarized for the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Effects of the reactor coolant environment on themore » mechanism of fatigue crack initiation are discussed. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. A fracture toughness J-R curve test was conducted on a commercial heat of Type 304 SS that was irradiated to {approx}2.0 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. The results were compared with the data obtained earlier on steels irradiated to 0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) (0.45 and 1.35 dpa). Neutron irradiation at 288 C was found to decrease the fracture toughness of austenitic SSs. Tests were conducted on compact-tension specimens of Alloy 600 under cyclic loading to evaluate the enhancement of crack growth rates in LWR environments. Then, the existing fatigue crack growth data on Alloys 600 and 690 were analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range on crack growth rates in air.« less

Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anh Bui; Nam Dinh; Brian Williams

In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Suchmore » sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable quantitative assessment of the CASL modeling of Crud-Induced Power Shift (CIPS) phenomenon, in particular, and the CASL advanced predictive capabilities, in general. This report is prepared for the Department of Energy’s Consortium for Advanced Simulation of LWRs program’s VUQ Focus Area.« less

The MaNGA integral field unit fiber feed system for the Sloan 2.5 m telescope

DOE Office of Scientific and Technical Information (OSTI.GOV)

Drory, N.; MacDonald, N.; Byler, N.

2015-02-01

We describe the design, manufacture, and performance of bare-fiber integral field units (IFUs) for the SDSS-IV survey Mapping Nearby Galaxies at Apache Point Observatory (MaNGA) on the the Sloan 2.5 m telescope at Apache Point Observatory. MaNGA is a luminosity-selected integral-field spectroscopic survey of 10{sup 4} local galaxies covering 360–1030 nm at R∼2200. The IFUs have hexagonal dense packing of fibers with packing regularity of 3 μm (rms), and throughput of 96 ± 0.5% from 350 nm to 1 μm in the lab. Their sizes range from 19 to 127 fibers (3–7 hexagonal layers) using Polymicro FBP 120:132:150 μm core:clad:buffermore » fibers to reach a fill fraction of 56%. High throughput (and low focal-ratio degradation (FRD)) is achieved by maintaining the fiber cladding and buffer intact, ensuring excellent surface polish, and applying a multi-layer anti-reflection (AR) coating of the input and output surfaces. In operations on-sky, the IFUs show only an additional 2.3% FRD-related variability in throughput despite repeated mechanical stressing during plate plugging (however other losses are present). The IFUs achieve on-sky throughput 5% above the single-fiber feeds used in SDSS-III/BOSS, attributable to equivalent performance compared to single fibers and additional gains from the AR coating. The manufacturing process is geared toward mass-production of high-multiplex systems. The low-stress process involves a precision ferrule with a hexagonal inner shape designed to lead inserted fibers to settle in a dense hexagonal pattern. The ferrule ID is tapered at progressively shallower angles toward its tip and the final 2 mm are straight and only a few microns larger than necessary to hold the desired number of fibers. Our IFU manufacturing process scales easily to accommodate other fiber sizes and can produce IFUs with substantially larger fiber counts. To assure quality, automated testing in a simple and inexpensive system enables complete characterization of throughput and fiber metrology. Future applications include larger IFUs, higher fill factors with stripped buffer, de-cladding, and lenslet coupling.« less

The MaNGA Integral Field Unit Fiber Feed System for the Sloan 2.5 m Telescope

NASA Astrophysics Data System (ADS)

Drory, N.; MacDonald, N.; Bershady, M. A.; Bundy, K.; Gunn, J.; Law, D. R.; Smith, M.; Stoll, R.; Tremonti, C. A.; Wake, D. A.; Yan, R.; Weijmans, A. M.; Byler, N.; Cherinka, B.; Cope, F.; Eigenbrot, A.; Harding, P.; Holder, D.; Huehnerhoff, J.; Jaehnig, K.; Jansen, T. C.; Klaene, M.; Paat, A. M.; Percival, J.; Sayres, C.

2015-02-01

We describe the design, manufacture, and performance of bare-fiber integral field units (IFUs) for the SDSS-IV survey Mapping Nearby Galaxies at Apache Point Observatory (MaNGA) on the the Sloan 2.5 m telescope at Apache Point Observatory. MaNGA is a luminosity-selected integral-field spectroscopic survey of 104 local galaxies covering 360-1030 nm at R˜ 2200. The IFUs have hexagonal dense packing of fibers with packing regularity of 3 μm (rms), and throughput of 96 ± 0.5% from 350 nm to 1 μm in the lab. Their sizes range from 19 to 127 fibers (3-7 hexagonal layers) using Polymicro FBP 120:132:150 μm core:clad:buffer fibers to reach a fill fraction of 56%. High throughput (and low focal-ratio degradation (FRD)) is achieved by maintaining the fiber cladding and buffer intact, ensuring excellent surface polish, and applying a multi-layer anti-reflection (AR) coating of the input and output surfaces. In operations on-sky, the IFUs show only an additional 2.3% FRD-related variability in throughput despite repeated mechanical stressing during plate plugging (however other losses are present). The IFUs achieve on-sky throughput 5% above the single-fiber feeds used in SDSS-III/BOSS, attributable to equivalent performance compared to single fibers and additional gains from the AR coating. The manufacturing process is geared toward mass-production of high-multiplex systems. The low-stress process involves a precision ferrule with a hexagonal inner shape designed to lead inserted fibers to settle in a dense hexagonal pattern. The ferrule ID is tapered at progressively shallower angles toward its tip and the final 2 mm are straight and only a few microns larger than necessary to hold the desired number of fibers. Our IFU manufacturing process scales easily to accommodate other fiber sizes and can produce IFUs with substantially larger fiber counts. To assure quality, automated testing in a simple and inexpensive system enables complete characterization of throughput and fiber metrology. Future applications include larger IFUs, higher fill factors with stripped buffer, de-cladding, and lenslet coupling.

VizieR Online Data Catalog: Radial velocities of the Be star HR 2142 (Peters+, 2016)

NASA Astrophysics Data System (ADS)

Peters, G. J.; Wang, L.; Gies, D. R.; Grundstrom, E. D.

2016-11-01

Radial velocity measurements were made using the set of spectra summarized in Table 1. The main focus of this work is a set of 88 high resolution, SWP HIRES FUV spectra acquired over the lifetime of the International Ultraviolet Explorer (IUE) observatory. These were downloaded from MAST and resampled. We also collected a set of 49 LWR and LWP near-UV spectra that were used to inspect the orbital variations in the MgII2796,2803 feature. The UV spectra were supplemented with a large collection of Hα spectra that we secured with the KPNO Coude Feed telescope and that were obtained by amateur astronomers participating in the Be Star Spectra database project (Pollmann 2007IBVS.5778....1P; Neiner et al. 2011AJ....142..149N). (2 data files).

Correlative Microscopy of Neutron-Irradiated Materials

DOE PAGES

Briggs, Samuel A.; Sridharan, Kumar; Field, Kevin G.

2016-12-31

A nuclear reactor core is a highly demanding environment that presents several unique challenges for materials performance. Materials in modern light water reactor (LWR) cores must survive several decades in high-temperature (300-350°C) aqueous corrosion conditions while being subject to large amounts of high-energy neutron irradiation. Next-generation reactor designs seek to use more corrosive coolants (e.g., molten salts) and even greater temperatures and neutron doses. The high amounts of disorder and unique crystallographic defects and microchemical segregation effects induced by radiation inevitably lead to property degradation of materials. Thus, maintaining structural integrity and safety margins over the course of the reactor'smore » service life thus necessitates the ability to understand and predict these degradation phenomena in order to develop new, radiation-tolerant materials that can maintain the required performance in these extreme conditions.« less

The novel top-coat material for RLS trade-off reduction in EUVL

NASA Astrophysics Data System (ADS)

Onishi, Ryuji; Sakamoto, Rikimaru; Fujitani, Noriaki; Endo, Takafumi; Ho, Bang-ching

2012-03-01

For the next generation lithography (NGL), several technologies have been proposed to achieve the 22nm-node devices and beyond. Extreme ultraviolet (EUV) lithography is one of the candidates for the next generation lithography. In EUV light source development, low power is one of the critical issue because of the low throughput, and another issue is Out of Band (OoB) light existing in EUV light. OoB is concerned to be the cause of deterioration for the lithography performance. In order to avoid this critical issue, we focused on development of the resist top coat material with OoB absorption property as Out of Band Protection Layer (OBPL). We designed this material having high absorbance around 240nm wavelength and high transmittance for EUV light. And this material aimed to improve sensitivity, resolution and LWR performance.

Reducing Line Edge Roughness in Si and SiN through plasma etch chemistry optimization for photonic waveguide applications

NASA Astrophysics Data System (ADS)

Marchack, Nathan; Khater, Marwan; Orcutt, Jason; Chang, Josephine; Holmes, Steven; Barwicz, Tymon; Kamlapurkar, Swetha; Green, William; Engelmann, Sebastian

2017-03-01

The LER and LWR of subtractively patterned Si and SiN waveguides was calculated after each step in the process. It was found for Si waveguides that adjusting the ratio of CF4:CHF3 during the hard mask open step produced reductions in LER of 26 and 43% from the initial lithography for isolated waveguides patterned with partial and full etches, respectively. However for final LER values of 3.0 and 2.5 nm on fully etched Si waveguides, the corresponding optical loss measurements were indistinguishable. For SiN waveguides, introduction of C4H9F to the conventional CF4/CHF3 measurement was able to reduce the mask height budget by a factor of 5, while reducing LER from the initial lithography by 26%.

Analysis of the Browns Ferry Unit 3 irradiation experiments. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simmons, G.L.

1984-11-01

The results of the analysis of two experiments performed at the Browns Ferry-3 reactor are presented. These calculations utilize state-of-the-art neutron transport techniques and a new neutron cross-section library that has been developed for LWR applications. The calculations agree well with the experimental data obtained in irradiations inside the reactor vessel. For the measurements performed in the reactor cavity, the calculations agree well at the reactor midplane. Accurate determination of the axial distribution of the neutron fluence in the reactor cavity depends on having a concise representation of the axial-void distribution in the core. Detailed data are presented describing themore » procedures used in the generation of the new cross-section library that has been named SAILOR. This library is available from the Radiation-Shielding Information Center.« less

Current and anticipated uses of thermalhydraulic and neutronic codes at PSI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aksan, S.N.; Zimmermann, M.A.; Yadigaroglu, G.

1997-07-01

The thermalhydraulic and/or neutronic codes in use at PSI mainly provide the capability to perform deterministic safety analysis for Swiss NPPs and also serve as analysis tools for experimental facilities for LWR and ALWR simulations. In relation to these applications, physical model development and improvements, and assessment of the codes are also essential components of the activities. In this paper, a brief overview is provided on the thermalhydraulic and/or neutronic codes used for safety analysis of LWRs, at PSI, and also of some experiences and applications with these codes. Based on these experiences, additional assessment needs are indicated, together withmore » some model improvement needs. The future needs that could be used to specify both the development of a new code and also improvement of available codes are summarized.« less

Photomask linewidth comparison by PTB and NIST

NASA Astrophysics Data System (ADS)

Bergmann, D.; Bodermann, B.; Bosse, H.; Buhr, E.; Dai, G.; Dixson, R.; Häßler-Grohne, W.; Hahm, K.; Wurm, M.

2015-10-01

We report the initial results of a recent bilateral comparison of linewidth or critical dimension (CD) calibrations on photomask line features between two national metrology institutes (NMIs): the National Institute of Standards and Technology (NIST) in the United States and the Physikalisch-Technische Bundesanstalt (PTB) in Germany. For the comparison, a chrome on glass (CoG) photomask was used which has a layout of line features down to 100 nm nominal size. Different measurement methods were used at both institutes. These included: critical dimension atomic force microscopy (CD-AFM), CD scanning electron microscopy (CD-SEM) and ultraviolet (UV) transmission optical microscopy. The measurands are CD at 50 % height of the features as well as sidewall angle and line width roughness (LWR) of the features. On the isolated opaque features, we found agreement of the CD measurements at the 3 nm to 5 nm level on most features - usually within the combined expanded uncertainties of the measurements.

In-reactor performance of LWR-type tritium target rods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

International Ultraviolet Explorer observations of the peculiar variable spectrum of the eclipsing binary R Arae

NASA Technical Reports Server (NTRS)

Mccluskey, G. E.; Kondo, Y.

1983-01-01

The eclipsing binary system R Arae = HD 149730 is a relatively bright southern system with an orbital period of about 4.4 days. It is a single-lined spectroscopic binary. The spectral class of the primary component is B9 Vp. The system was included in a study of mass flow and evolution in close binary systems using the International Ultraviolet Explorer satellite (IUE). Four spectra in the wavelength range from 1150 to 1900 A were obtained with the far-ultraviolet SWP camera, and six spectra in the range from 1900 to 3200 range were obtained with the mid-ultraviolet LWR camera. The close binary R Arae exhibits very unusual ultraviolet spectra. It appears that no other close binary system, observed with any of the orbiting satellites, shows outside-eclipse ultraviolet continuum flux variations of this nature.

Light Water Reactor Sustainability Program: Risk-Informed Safety Margins Characterization (RISMC) Pathway Technical Program Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Curtis; Rabiti, Cristian; Martineau, Richard

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degreemore » of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”« less

Pattern fidelity improvement of chemo-epitaxy DSA process for high-volume manufacturing

NASA Astrophysics Data System (ADS)

Muramatsu, Makoto; Nishi, Takanori; You, Gen; Saito, Yusuke; Ido, Yasuyuki; Ito, Kiyohito; Tobana, Toshikatsu; Hosoya, Masanori; Chen, Weichien; Nakamura, Satoru; Somervell, Mark; Kitano, Takahiro

2016-03-01

Directed self-assembly (DSA) is one of the candidates for next generation lithography. Over the past few years, cylindrical and lamellar structures dictated by the block co-polymer (BCP) composition have been investigated for use in patterning contact holes or lines, and, Tokyo Electron Limited (TEL is a registered trademark or a trademark of Tokyo Electron Limited in Japan and /or other countries.) has presented the evaluation results and the advantages of each-1-5. In this report, we will present the latest results regarding the defect reduction work on a model line/space system. Especially it is suggested that the defectivity of the neutral layer has a large impact on the defectivity of the DSA patterns. Also, LER/LWR reduction results will be presented with a focus on the improvements made during the etch transferring the DSA patterns into the underlayer.

Application of subsize specimens in nuclear plant life extension

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosinski, S.T.; Kumar, A.S.; Cannon, S.C.

1991-01-01

The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missour-Rolla to test a correlation for the upper shelf energy (USE) values obtained from the impact testing of subsize Charpy V-notch specimens to those obtained from the testing of full size samples. The program involves the impact testing of unirradiated and irradiated full, half, and third size Charpy V-notch specimens. To verify the applicability of the correlation on LWR materials unirradiated and irradiated full, half, and third size Charpy V-notch specimens of a commercial pressure vessel steel (ASTM A533 Grade B) willmore » be tested. This paper will provide details of the program and present results obtained from the application of the developed correlation methodology to the impact testing of the unirradiated full, half, and third size A533 Grade B Charpy V-notch specimens.« less

Application of subsize specimens in nuclear plant life extension

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosinski, S.T.; Kumar, A.S.; Cannon, S.C.

1991-12-31

The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missour-Rolla to test a correlation for the upper shelf energy (USE) values obtained from the impact testing of subsize Charpy V-notch specimens to those obtained from the testing of full size samples. The program involves the impact testing of unirradiated and irradiated full, half, and third size Charpy V-notch specimens. To verify the applicability of the correlation on LWR materials unirradiated and irradiated full, half, and third size Charpy V-notch specimens of a commercial pressure vessel steel (ASTM A533 Grade B) willmore » be tested. This paper will provide details of the program and present results obtained from the application of the developed correlation methodology to the impact testing of the unirradiated full, half, and third size A533 Grade B Charpy V-notch specimens.« less

Overview of NRC Proactive Management of Materials Degradation (PMMD) Program

NASA Astrophysics Data System (ADS)

Carpenter, C. E. Gene; Hull, Amy; Oberson, Greg

Materials degradation phenomena, if not appropriately managed, have the potential to adversely impact the design functionality and safety margins of nuclear power plant (NPP) systems, structures and components (SSCs). Therefore, the U.S. Nuclear Regulatory Commission (NRC) has initiated an over-the-horizon multi-year research Proactive Management of Materials Degradation (PMMD) Research Program, which is presently evaluating longer time frames (i.e., 80 or more years) and including passive long-lived SSCs beyond the primary piping and core internals, such as concrete containment and cable insulation. This will allow the NRC to (1) identify significant knowledge gaps and new forms of degradation; (2) capture current knowledge base; and, (3) prioritize materials degradation research needs and directions for future efforts. This effort is being accomplished in collaboration with the U.S. Department of Energy's (DOE) LWR Sustainability (LWRS) program. This presentation will discuss the activities to date, including results, and the path forward.

Analysis of user equilibrium for staggered shifts in a single-entry traffic corridor with no late arrivals

NASA Astrophysics Data System (ADS)

Li, Chuan-Yao; Huang, Hai-Jun; Tang, Tie-Qiao

2017-05-01

In this paper, we investigate the effects of staggered shifts on the user equilibrium (UE) state in a single-entry traffic corridor with no late arrivals from the analytical and numerical perspective. The LWR (Lighthill-Whitham-Richards) model and the Greenshields' velocity-density function are used to describe the dynamic properties of traffic flow. Propositions for the properties of flow patterns in UE, and the quasi-analytic solutions for three possible situations in UE are deduced. Numerical tests are carried out to testify the analytical results, where the three-dimensional evolution diagram of traffic flow illustrates that shock and rarefaction wave exist in UE and the space-time diagram indicates that UE solutions satisfy the propagation properties of traffic flow. In addition, the cost curves show that the UE solutions satisfy the UE trip-timing condition.

Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nanstad, Randy K.; Odette, George R.; Almirall, Nathan

2017-05-01

The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughnessmore » loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.« less

Preliminary comparative assessment of land use for the Satellite Power System (SPS) and alternative electric energy technologies

NASA Technical Reports Server (NTRS)

Newsom, D. E.; Wolsko, T.

1980-01-01

A preliminary comparative assessment of land use for the satellite power system (SPS), other solar technologies, and alternative electric energy technologies was conducted. The alternative technologies are coal gasification/combined-cycle, coal fluidized-bed combustion (FBC), light water reactor (LWR), liquid metal fast breeder reactor (LMFBR), terrestrial photovoltaics (TPV), solar thermal electric (STE), and ocean thermal energy conversion (OTEC). The major issues of a land use assessment are the quantity, purpose, duration, location, and costs of the required land use. The phased methodology described treats the first four issues, but not the costs. Several past efforts are comparative or single technology assessment are reviewed briefly. The current state of knowledge about land use is described for each technology. Conclusions are drawn regarding deficiencies in the data on comparative land use and needs for further research.

Graphoepitaxy integration and pattern transfer of lamellar silicon-containing high-chi block copolymers

NASA Astrophysics Data System (ADS)

Bézard, P.; Chevalier, X.; Legrain, A.; Navarro, C.; Nicolet, C.; Fleury, G.; Cayrefourcq, I.; Tiron, R.; Zelsmann, M.

2018-03-01

In this work, we present our recent achievements on the integration and transfer etching of a novel silicon-containing high-χ block copolymer for lines/spaces applications. Developed carbo-silane BCPs are synthesized under industrial conditions and present periodicities as low as 14 nm. A full directed self-assembly by graphoepitaxy process is shown using standard photolithography stacks and all processes are performed on 300 mm wafer compatible tools. Specific plasma processes are developed to isolate perpendicular lamellae and sub-12 nm features are finally transferred into silicon substrates. The quality of the final BCP hard mask (CDU, LWR, LER) are also investigated. Finally, thanks to the development of dedicated neutral layers and top-coats allowing perpendicular orientations, it was possible to investigate plasma etching experiments on full-sheets at 7 nm resolution, opening the way to the integration of these polymers in chemoepitaxy stacks.

US Efforts in Support of Examinations at Fukushima Daiichi – 2016 Evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amway, P.; Andrews, N.; Bixby, Willis

Although it is clear that the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limitedmore » full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings (TEPCO) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document reports recent results from the US Forensics Effort to use information obtained by TEPCO to enhance the safety of existing and future nuclear power plant designs. This Forensics Effort, which is sponsored by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of US experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO information from Daiichi that address these needs. Examples presented in this report demonstrate that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities.« less

Sensitivity enhancement of chemically amplified resists and performance study using extreme ultraviolet interference lithography

NASA Astrophysics Data System (ADS)

Buitrago, Elizabeth; Nagahara, Seiji; Yildirim, Oktay; Nakagawa, Hisashi; Tagawa, Seiichi; Meeuwissen, Marieke; Nagai, Tomoki; Naruoka, Takehiko; Verspaget, Coen; Hoefnagels, Rik; Rispens, Gijsbert; Shiraishi, Gosuke; Terashita, Yuichi; Minekawa, Yukie; Yoshihara, Kosuke; Oshima, Akihiro; Vockenhuber, Michaela; Ekinci, Yasin

2016-07-01

Extreme ultraviolet lithography (EUVL, λ=13.5 nm) is the most promising candidate to manufacture electronic devices for future technology nodes in the semiconductor industry. Nonetheless, EUVL still faces many technological challenges as it moves toward high-volume manufacturing (HVM). A key bottleneck from the tool design and performance point of view has been the development of an efficient, high-power EUV light source for high throughput production. Consequently, there has been extensive research on different methodologies to enhance EUV resist sensitivity. Resist performance is measured in terms of its ultimate printing resolution, line width roughness (LWR), sensitivity [S or best energy (BE)], and exposure latitude (EL). However, there are well-known fundamental trade-off relationships (line width roughness, resolution and sensitivity trade-off) among these parameters for chemically amplified resists (CARs). We present early proof-of-principle results for a multiexposure lithography process that has the potential for high sensitivity enhancement without compromising other important performance characteristics by the use of a "Photosensitized Chemically Amplified Resist™" (PSCAR™). With this method, we seek to increase the sensitivity by combining a first EUV pattern exposure with a second UV-flood exposure (λ=365 nm) and the use of a PSCAR. In addition, we have evaluated over 50 different state-of-the-art EUV CARs. Among these, we have identified several promising candidates that simultaneously meet sensitivity, LWR, and EL high-performance requirements with the aim of resolving line space (L/S) features for the 7- and 5-nm logic node [16- and 13-nm half-pitch (HP), respectively] for HVM. Several CARs were additionally found to be well resolved down to 12- and 11-nm HP with minimal pattern collapse and bridging, a remarkable feat for CARs. Finally, the performance of two negative tone state-of-the-art alternative resist platforms previously investigated was compared to the CAR performance at and below 16-nm HP resolution, demonstrating the need for alternative resist solutions at 13-nm resolution and below. EUV interference lithography (IL) has provided and continues to provide a simple yet powerful platform for academic and industrial research, enabling the characterization and development of resist materials before commercial EUV exposure tools become available. Our experiments have been performed at the EUV-IL set-up in the Swiss Light Source (SLS) synchrotron facility located at the Paul Scherrer Institute (PSI).

Sensitivity enhancement of chemically amplified resists and performance study using EUV interference lithography

NASA Astrophysics Data System (ADS)

Buitrago, Elizabeth; Nagahara, Seiji; Yildirim, Oktay; Nakagawa, Hisashi; Tagawa, Seiichi; Meeuwissen, Marieke; Nagai, Tomoki; Naruoka, Takehiko; Verspaget, Coen; Hoefnagels, Rik; Rispens, Gijsbert; Shiraishi, Gosuke; Terashita, Yuichi; Minekawa, Yukie; Yoshihara, Kosuke; Oshima, Akihiro; Vockenhuber, Michaela; Ekinci, Yasin

2016-03-01

Extreme ultraviolet lithography (EUVL, λ = 13.5 nm) is the most promising candidate to manufacture electronic devices for future technology nodes in the semiconductor industry. Nonetheless, EUVL still faces many technological challenges as it moves toward high-volume manufacturing (HVM). A key bottleneck from the tool design and performance point of view has been the development of an efficient, high power EUV light source for high throughput production. Consequently, there has been extensive research on different methodologies to enhance EUV resist sensitivity. Resist performance is measured in terms of its ultimate printing resolution, line width roughness (LWR), sensitivity (S or best energy BE) and exposure latitude (EL). However, there are well-known fundamental trade-off relationships (LRS trade-off) among these parameters for chemically amplified resists (CARs). Here we present early proof-of-principle results for a multi-exposure lithography process that has the potential for high sensitivity enhancement without compromising other important performance characteristics by the use of a Photosensitized Chemically Amplified Resist (PSCAR). With this method, we seek to increase the sensitivity by combining a first EUV pattern exposure with a second UV flood exposure (λ = 365 nm) and the use of a PSCAR. In addition, we have evaluated over 50 different state-of-the-art EUV CARs. Among these, we have identified several promising candidates that simultaneously meet sensitivity, LWR and EL high performance requirements with the aim of resolving line space (L/S) features for the 7 and 5 nm logic node (16 nm and 13 nm half-pitch HP, respectively) for HVM. Several CARs were additionally found to be well resolved down to 12 nm and 11 nm HP with minimal pattern collapse and bridging, a remarkable feat for CARs. Finally, the performance of two negative tone state-of-the-art alternative resist platforms previously investigated was compared to the CAR performance at and below 16 nm HP resolution, demonstrating the need for alternative resist solutions at 13 nm resolution and below. EUV interference lithography (IL) has provided and continues to provide a simple yet powerful platform for academic and industrial research enabling the characterization and development of new resist materials before commercial EUV exposure tools become available. Our experiments have been performed at the EUV-IL set-up in the Swiss Light Source (SLS) synchrotron facility located at the Paul Scherrer Institute (PSI).

a Study of the Interferences with the On-Line Radioiodine Measurement Under Nuclear Accident Conditions

NASA Astrophysics Data System (ADS)

Tseng, Tung-Tse

In this research the interferences with the on -line detection of radioiodines, under nuclear accident conditions, were studied. The special tool employed for this research is the developed on-line radioiodine monitor (the Penn State Radioiodine Monitor), which is capable of detecting low levels of radioiodine on-line in air containing orders of magnitude higher levels of radioactive noble gases. Most of the data reported in this thesis were collected during a series of experiments called "Source -Term Experiment Program (STEP)." The experiments were conducted at the Argonne National Laboratory's TREAT reactor located at the Idaho National Engineering Laboratory (INEL). In these tests, fission products were released from the Light Water Reactor (LWR) test fuels as a result of simulating a reactor accident. The Penn State Monitor was then used to sample the fission products accumulated in a large container which simulated the reactor containment building. The test results proved that the Penn State Monitor was not affected significantly by the passage of large amounts of noble gases through the system. Also, it confirmed the predicted results that the operation of conventional on-line radioiodine detectors would, under nuclear accident conditions, be seriously impaired by the passage of high concentrations of radioactive noble gases through such systems. This work also demonstrated that under conditions of high noble gas concentrations and low radioiodine concentrations, the formation of noble-gas-decayed alkali metals can seriously interfere with the on-line detection of radioiodine, especially during the 24 hours immediately after the accident. The decayed alkali metal particulates were also found to be much more penetrating than the ordinary type of particulates, since a large fraction (15%) of the particulates were found to penetrate through the commonly used High Efficiency Particulate Air (HEPA) filter (rated >99.97% for 0.3 (mu)m particulate). Also, a significant fraction ((TURN)40%) of these particles became deposited on silver zeolite iodine filters inside the counting chamber. Finally, the Penn State Monitor proved itself to be a powerful research tool for the on-line source term studies since it can easily produce near noble-gas-free spectra during the real time studies occurring under simulated nuclear accident conditions.

L3.PHI.CTF.P10.02-rev2 Coupling of Subchannel T/H (CTF) and CRUD Chemistry (MAMBA1D)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K.; Palmtag, Scott; Collins, Benjamin S.

2015-05-15

The purpose of this milestone is to create a preliminary capability for modeling light water reactor (LWR) thermal-hydraulic (T/H) and CRUD growth using the CTF subchannel code and the subgrid version of the MAMBA CRUD chemistry code, MAMBA1D. In part, this is a follow-on to Milestone L3.PHI.VCS.P9.01, which is documented in Report CASL-U-2014-0188-000, titled "Development of CTF Capability for Modeling Reactor Operating Cycles with Crud Growth". As the title suggests, the previous milestone set up a framework for modeling reactor operation cycles with CTF. The framework also facilitated coupling to a CRUD chemistry capability for modeling CRUD growth throughout themore » reactor operating cycle. To demonstrate the capability, a simple CRUD \\surrogate" tool was developed and coupled to CTF; however, it was noted that CRUD growth predictions by the surrogate were not considered realistic. This milestone builds on L3.PHI.VCS.P9.01 by replacing this simple surrogate tool with the more advanced MAMBA1D CRUD chemistry code. Completing this task involves addressing unresolved tasks from Milestone L3.PHI.VCS.P9.01, setting up an interface to MAMBA1D, and extracting new T/H information from CTF that was not previously required in the simple surrogate tool. Speci c challenges encountered during this milestone include (1) treatment of the CRUD erosion model, which requires local turbulent kinetic energy (TKE) (a value that CTF does not calculate) and (2) treatment of the MAMBA1D CRUD chimney boiling model in the CTF rod heat transfer solution. To demonstrate this new T/H, CRUD modeling capability, two sets of simulations were performed: (1) an 18 month cycle simulation of a quarter symmetry model of Watts Bar and (2) a simulation of Assemblies G69 and G70 from Seabrook Cycle 5. The Watts Bar simulation is merely a demonstration of the capability. The simulation of the Seabrook cycle, which had experienced CRUD-related fuel rod failures, had actual CRUD-scrape data to compare with results. As results show, the initial CTF/MAMBA1D-predicted CRUD thicknesses were about half of their expected values, so further investigation will be required for this simulation.« less

Corrosion fatigue of alloys 600 and 690 in simulated LWR environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruther, W.E.; Soppett, W.K.; Kassner, T.F.

1996-04-01

Crack growth data were obtained on fracture-mechanics specimens of Alloys 600 and 690 to investigate environmentally assisted cracking (EAC) in simulated boiling water reactor and pressurized water reactor environments at 289 and 320 C. Preliminary information was obtained on the effect of temperature, load ratio, stress intensity (K), and the dissolved-oxygen and -hydrogen concentrations of the water on EAC. Specimens of Type 316NG and sensitized Type 304 stainless steel (SS) were included in several of the experiments to assess the behavior of these materials and Alloy 600 under the same water chemistry and loading conditions. The experimental data are comparedmore » with predictions from an Argonne National Laboratory (ANL) model for crack growth rates (CGRs) of SSs in water and the ASME Code Section 11 correlation for CGRs in air at the K{sub max} and load-ratio values in the various tests. The data for all of the materials were bounded by ANL model predictions and the ASME Section 11 ``air line.``« less

Microstructural evolution of type 304 and 316 stainless steels under neutron irradiation at LWR relevant conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tan, Lizhen; Stoller, Roger E.; Field, Kevin G.

Extension of light water reactors' useful life will expose austenitic internal core components to irradiation damage levels beyond 100 displacements per atom (dpa), which will lead to profound microstructural evolution and consequent degradation of macroscopic properties. Microstructural evolution, including Frank loops, cavities, precipitates, and segregation at boundaries and the resultant radiation hardening in type 304 and 316 stainless steel (SS) variants, were studied in this work via experimental characterization and multiple simulation methods. Experimental data for up to 40 heats of type 304SS and 316SS variants irradiated in different reactors to 0.6–120 dpa at 275–375°C were either generated from thismore » work or collected from literature reports. These experimental data were then combined with models of Frank loop and cavity evolution, computational thermodynamics and precipitation, and ab initio and rate theory integrated radiation-induced segregation models to provide insights into microstructural evolution and degradation at higher radiation doses.« less

Feasibility of using near infrared spectroscopy to detect and quantify an adulterant in high quality sandalwood oil.

PubMed

Kuriakose, Saji; Joe, I Hubert

2013-11-01

Determination of the authenticity of essential oils has become more significant, in recent years, following some illegal adulteration and contamination scandals. The present investigative study focuses on the application of near infrared spectroscopy to detect sample authenticity and quantify economic adulteration of sandalwood oils. Several data pre-treatments are investigated for calibration and prediction using partial least square regression (PLSR). The quantitative data analysis is done using a new spectral approach - full spectrum or sequential spectrum. The optimum number of PLS components is obtained according to the lowest root mean square error of calibration (RMSEC=0.00009% v/v). The lowest root mean square error of prediction (RMSEP=0.00016% v/v) in the test set and the highest coefficient of determination (R(2)=0.99989) are used as the evaluation tools for the best model. A nonlinear method, locally weighted regression (LWR), is added to extract nonlinear information and to compare with the linear PLSR model. Copyright © 2013 Elsevier B.V. All rights reserved.

Online Learning Flight Control for Intelligent Flight Control Systems (IFCS)

NASA Technical Reports Server (NTRS)

Niewoehner, Kevin R.; Carter, John (Technical Monitor)

2001-01-01

The research accomplishments for the cooperative agreement 'Online Learning Flight Control for Intelligent Flight Control Systems (IFCS)' include the following: (1) previous IFC program data collection and analysis; (2) IFC program support site (configured IFC systems support network, configured Tornado/VxWorks OS development system, made Configuration and Documentation Management Systems Internet accessible); (3) Airborne Research Test Systems (ARTS) II Hardware (developed hardware requirements specification, developing environmental testing requirements, hardware design, and hardware design development); (4) ARTS II software development laboratory unit (procurement of lab style hardware, configured lab style hardware, and designed interface module equivalent to ARTS II faceplate); (5) program support documentation (developed software development plan, configuration management plan, and software verification and validation plan); (6) LWR algorithm analysis (performed timing and profiling on algorithm); (7) pre-trained neural network analysis; (8) Dynamic Cell Structures (DCS) Neural Network Analysis (performing timing and profiling on algorithm); and (9) conducted technical interchange and quarterly meetings to define IFC research goals.

Driving down defect density in composite EUV patterning film stacks

NASA Astrophysics Data System (ADS)

Meli, Luciana; Petrillo, Karen; De Silva, Anuja; Arnold, John; Felix, Nelson; Johnson, Richard; Murray, Cody; Hubbard, Alex; Durrant, Danielle; Hontake, Koichi; Huli, Lior; Lemley, Corey; Hetzer, Dave; Kawakami, Shinichiro; Matsunaga, Koichi

2017-03-01

Extreme ultraviolet lithography (EUVL) technology is one of the leading candidates for enabling the next generation devices, for 7nm node and beyond. As the technology matures, further improvement is required in the area of blanket film defectivity, pattern defectivity, CD uniformity, and LWR/LER. As EUV pitch scaling approaches sub 20 nm, new techniques and methods must be developed to reduce the overall defectivity, mitigate pattern collapse and eliminate film related defect. IBM Corporation and Tokyo Electron Limited (TELTM) are continuously collaborating to develop manufacturing quality processes for EUVL. In this paper, we review key defectivity learning required to enable 7nm node and beyond technology. We will describe ongoing progress in addressing these challenges through track-based processes (coating, developer, baking), highlighting the limitations of common defect detection strategies and outlining methodologies necessary for accurate characterization and mitigation of blanket defectivity in EUV patterning stacks. We will further discuss defects related to pattern collapse and thinning of underlayer films.

Microstructural evolution of type 304 and 316 stainless steels under neutron irradiation at LWR relevant conditions

DOE PAGES

Tan, Lizhen; Stoller, Roger E.; Field, Kevin G.; ...

2015-12-11

Extension of light water reactors' useful life will expose austenitic internal core components to irradiation damage levels beyond 100 displacements per atom (dpa), which will lead to profound microstructural evolution and consequent degradation of macroscopic properties. Microstructural evolution, including Frank loops, cavities, precipitates, and segregation at boundaries and the resultant radiation hardening in type 304 and 316 stainless steel (SS) variants, were studied in this work via experimental characterization and multiple simulation methods. Experimental data for up to 40 heats of type 304SS and 316SS variants irradiated in different reactors to 0.6–120 dpa at 275–375°C were either generated from thismore » work or collected from literature reports. These experimental data were then combined with models of Frank loop and cavity evolution, computational thermodynamics and precipitation, and ab initio and rate theory integrated radiation-induced segregation models to provide insights into microstructural evolution and degradation at higher radiation doses.« less

Social optimum for evening commute in a single-entry traffic corridor with no early departures

NASA Astrophysics Data System (ADS)

Li, Chuan-Yao; Xu, Guang-Ming; Tang, Tie-Qiao

2018-07-01

In this paper, we investigate the evening commute behaviors on the social optimum (SO) state in a single-entry traffic corridor with no early departures. Differing from the previous studies on evening commute, the dynamic properties of traffic flow are analyzed with the LWR (Lighthill-Whitham-Richards) model. The properties of optimum cumulative inflow curve with general desired departure time distribution curve are deduced, and then the analytic solutions for common desired departure time in SO are obtained. Three numerical examples are carried out to capture the characteristics of evening commuting behaviors under different values of time. The analytic and numerical results both indicate that the rarefaction wave originating from the first entry point influences the whole or part of the outflow curve. No shock wave exists through the commuting process. In addition, the cost curves show that the trip cost increases and the departure delay cost decreases with departure time, whereas the travel time cost first increases then decreases with departure time under the SO principle.

Data decomposition of Monte Carlo particle transport simulations via tally servers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Romano, Paul K.; Siegel, Andrew R.; Forget, Benoit

An algorithm for decomposing large tally data in Monte Carlo particle transport simulations is developed, analyzed, and implemented in a continuous-energy Monte Carlo code, OpenMC. The algorithm is based on a non-overlapping decomposition of compute nodes into tracking processors and tally servers. The former are used to simulate the movement of particles through the domain while the latter continuously receive and update tally data. A performance model for this approach is developed, suggesting that, for a range of parameters relevant to LWR analysis, the tally server algorithm should perform with minimal overhead on contemporary supercomputers. An implementation of the algorithmmore » in OpenMC is then tested on the Intrepid and Titan supercomputers, supporting the key predictions of the model over a wide range of parameters. We thus conclude that the tally server algorithm is a successful approach to circumventing classical on-node memory constraints en route to unprecedentedly detailed Monte Carlo reactor simulations.« less

Corrosion fatigue characterization of reactor pressure vessel steels. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Der Sluys, W.A.

1982-12-01

During routine operation, light water reactor (LWR) pressure vessels are subjected to a variety of transients that result in time-varying stresses. Consequently, fatigue and environmentally-assisted fatigue are mechanisms of growth relevant to flaws in these pressure vessels. To provide a better understanding of the resistance of nuclear pressure vessel steels to these flaw growth processes, fracture mechanics data were generated on the rates of fatigue crack growth for SA508-2 and SA533B-1 steels in both room temperature air and 288/sup 0/C water. Areas investigated were: the relationship of crack growth rate to prior loading history; the effects of loading frequency andmore » R ratio (K/sub min//K/sub max/) on crack growth rate as a function of the stress intensity factor range (..delta..K); transient aspects of the fatigue crack growth behavior; the effect of material chemistry (sulphur content) on fatigue crack; and growth rate; water chemistry effects (high-purity water versus simulated pressurized water reactotr (PWR) primary coolant).« less

DEVELOPMENT OF HFE SECTIONS OF DG-1145.

DOE Office of Scientific and Technical Information (OSTI.GOV)

HIGGINS,J.C.; OHARA, J.M.; BONGARRA, J.

2007-03-26

For the licensing of the current fleet of commercial nuclear power plants (NPPs), the Nuclear Regulatory Commission (NRC) used two key documents, NUREG-0800 and Regulatory Guide (RG) 1.70. RG 1.70 provided guidance to applicants on the contents needed in their Safety Analysis Reports (SARs) submitted as part of their application to construct or operate an NPP. NUREG-0800, the NRC Standard Review Plan (SRP), provides guidance to the NRR staff reviewers on performing their safety reviews of these applications. As part of the preparation for a new wave of improved NPP designs the NRC is in the process of updating themore » SRP and is also developing a new RG designated as draft RG or DG-1145, ''Combined License Applications for Nuclear Power Plants (LWR Edition).'' This will eventually become RG 1.206 and will take the place of RG 1.70. This will provide guidance for combined license (COL) applicants, as well as for other 10CFR Part 52 variations that are permitted.« less

Curvature controlled wetting in two dimensions

NASA Astrophysics Data System (ADS)

Gil, Tamir; Mikheev, Lev V.

1995-07-01

A complete wetting transition at vanishing curvature of the substrate in two-dimensional circular geometry is studied by the transfer matrix method. We find an exact formal mapping of the partition function of the problem onto that of a (1+1)-dimensional wetting problem in planar geometry. As the radius of the substrate r0-->∞, the leading effect of the curvature is adding the Laplace pressure ΠL~r-10 to the pressure balance in the film. At temperatures and pressures under which the wetting is complete in planar geometry, Laplace pressure suppresses divergence of the mean thickness of the wetting layer lW, leading to a power law lW~r1/30. At a critical wetting transition of a planar substrate, curvature adds a relevant field; the corresponding multiscaling forms are readily available. The method allows for the systematic evaluation of corrections to the leading behavior; the next to the leading term reduces the thickness by the amount proportional to r-1/30

The peculiar, luminous early-type emission line stars of the Magellanic clouds: A preliminary taxonomy

NASA Technical Reports Server (NTRS)

Shore, S. N.; Sanduleak, N.

1982-01-01

A sample of some 20 early type emission supergiants in the Magellanic clouds was observed with both the SWP and LWR low resolution mode of IUE. All stars have strong H-emission, some showing P-Cygni structure as well with HeI, HeII, FeII and other ions also showing strong emission. It is found that the stars fall into three distinct groups on the basis of the HeII/HeI and HeI/HI strengths: (1) HeII strong, HeI, HI; (2) HeII absent, HeI, HI strong; (3) HeI absent, HI, FeII, FeII, strong in addition to low excitation ions. The two most extreme emission line stars found in the Clouds S 134/LMC and S 18/SMC are discussed. Results for the 2200A feature in these supergiants, and evidence for shells around the most luminous stars in the clouds are also described.

Shot noise limit of chemically amplified resists with photodecomposable quenchers used for extreme ultraviolet lithography

NASA Astrophysics Data System (ADS)

Kozawa, Takahiro; Santillan, Julius Joseph; Itani, Toshiro

2017-06-01

In lithography using high-energy photons such as an extreme ultraviolet (EUV) radiation, the shot noise of photons is a critical issue. The shot noise is a cause of line edge/width roughness (LER/LWR) and stochastic defect generation and limits the resist performance. In this study, the effects of photodecomposable quenchers were investigated from the viewpoint of the shot noise limit. The latent images of line-and-space patterns with 11 nm half-pitch were calculated using a Monte Carlo method. In the simulation, the effect of secondary electron blur was eliminated to clarify the shot noise limits regarding stochastic phenomena such as LER. The shot noise limit for chemically amplified resists with acid generators and photodecomposable quenchers was approximately the same as that for chemically amplified resists with acid generators and conventional quenchers when the total sensitizer concentration was the same. The effect of photodecomposable quenchers on the shot noise limit was essentially the same as that of acid generators.

High-volume manufacturing compatible dry development rinse process (DDRP): patterning and defectivity performance for EUVL

NASA Astrophysics Data System (ADS)

Sayan, Safak; Vanelderen, Pieter; Hetel, Iulian; Chan, BT; Raghavan, Praveen; Blanco, Victor; Foubert, Philippe; D'urzo, Lucia; De Simone, Danilo; Vandenberghe, Geert

2017-04-01

There are many knobs available that change the chemical and physical properties of the photoresists to "break" the RLS (Resolution, Sensitivity, Line edge/width roughness) trade-off, however those are not enough today to realize a material to satisfy all requirements at once for 7nm technology and beyond. DDRP improves the ultimate achievable resolution via pattern collapse mitigation, hence the priority of requirements for the EUV photoresist development may be changed with more focus on Sensitivity and LWR. This may potentially provide a new conceptual approach towards EUV PR development for DDRP applications. We have previously demonstrated pattern collapse (PC) mitigation via DDRP on different EUVL photoresists (including different resist platforms), achieving ultimate resolution and exposure latitude improvements [1,2]. In this contribution, we report patterning and material defect performance of HVM compatible (all aqueous) dry development rinse material. We will also report on process window improvement on 2-dimensional metal structures towards standard cell size reduction with elimination of mask layer(s) using single EUV exposure.

DSA process window expansion with novel DSA track hardware

NASA Astrophysics Data System (ADS)

Harumoto, Masahiko; Stokes, Harold; Tanaka, Yuji; Kaneyama, Koji; Pieczulewski, Chalres; Asai, Masaya; Argoud, Maxime; Servin, Isabelle; Chamiot-Maitral, Gaëlle; Claveau, Guillaume; Tiron, Raluca; Cayrefourcq, Ian

2017-03-01

PS-b-PMMA block copolymer is a well-known DSA material, and there are many DSA patterning methods that make effective the use of such 1st generation materials. Consequently, this variety of patterning methods opens a wide array of possibilities for DSA application[1-4]. Last year, during the inaugural International DSA Symposium, researchers and lithographers concurred on common key issues for DSA patterning methods such as: defect density, LWR, placement error, etc. Defect density was specifically expressed as the biggest obstacle for new processes. Coat-Develop track systems contribute to the DSA pattern fabrication and also influence the DSA pattern performances[4]. In this study, defectivity was investigated using an atmosphere-controlled chamber on the SOKUDO DUO track. As an initial step for expanding the DSA process window, fingerprint patterns were used for various atmospheric conditions during DSA self-assembly annealing. In this study, we will demonstrate an improved DSA process window, and then we will discuss the mechanism for this atmospheric effect.

Experimental study of contact edge roughness on sub-100 nm various circular shapes

NASA Astrophysics Data System (ADS)

Lee, Tae Y.; Ihm, Dongchul; Kang, Hyo C.; Lee, Jum B.; Lee, Byoung H.; Chin, Soo B.; Cho, Do H.; Song, Chang L.

2005-05-01

The measurement of edge roughness has become a hot issue in the semiconductor industry. Especially the contact roughness is being more critical as design rule shrinks. Major vendors offer a variety of features to measure the edge roughness in their CD-SEMs. For the line and space patterns, features such as Line Edge Roughness (LER) and Line Width Roughness (LWR) are available in current CD-SEMs. However the features currently available in commercial CD-SEM cannot provide a proper solution in monitoring the contact roughness. We had introduced a new parameter R, measurement algorithm and definition of contact edge roughness to quantify CER and CSR in previous paper. The parameter, R could provide an alternative solution to monitor contact or island pattern roughness. In this paper, we investigated to assess optimum number of CD measurement (1-D) and fitting method for CER or CSR. The study was based on a circular contact shape. Some new ideas to quantify CER or CSR were also suggested with preliminary experimental results.