Fully Ceramic Microencapsulated Fuel Development for LWR Applications

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

Snead, Lance Lewis; Besmann, Theodore M; Terrani, Kurt A

2012-01-01

The concept, fabrication, and key feasibility issues of a new fuel form based on the microencapsulated (TRISO-type) fuel which has been specifically engineered for LWR application and compacted within a SiC matrix will be presented. This fuel, the so-called fully ceramic microencapsulated fuel is currently undergoing development as an accident tolerant fuel for potential UO2 replacement in commercial LWRs. While the ability of this fuel to facilitate normal LWR cycle performance is an ongoing effort within the program, this will not be a focus of this paper. Rather, key feasibility and performance aspects of the fuel will be presented includingmore » the ability to fabricate a LWR-specific TRISO, the need for and route to a high thermal conductivity and fully dense matrix that contains neutron poisons, and the performance of that matrix under irradiation and the interaction of the fuel with commercial zircaloy clad.« less

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

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

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

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

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

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.

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

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

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.

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

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

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

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

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.

Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

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

Jiao, Zhujie; Was, Gary; Bartels, David

2015-04-02

This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that themore » effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.« less

Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

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

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.

1994-04-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ``flooded cavity``, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array canmore » deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications.« less

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

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.

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

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.

Spent fuel data base: commercial light water reactors. [PWR; BWR

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

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

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

Michael A. Pope; R. Sonat Sen; Brian Boer

2011-09-01

The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code tomore » assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.« 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.

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.

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

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.

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

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.

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.

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%.

Evaluation of Methods for Decladding LWR Fuel for a Pyroprocessing-Based Reprocessing Plant

DTIC Science & Technology

1992-10-01

oAD-A275 326 ORN.rFM-1121o04 OAK RIDGE NATIONAL LABORATORY Evaluation of Methods for Decladding _LWR Fuel for a Pyroprocessing -Based Reprocessing...Dist. Category UC-526 EVALUATION OF METHODS FOR DECLADDING LWR FUEL FOR A PYROPROCESSING -BASED REPROCESSING PLANT W. D. Bond J. C. Mailen G. E...decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyroprocesses

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

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

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.

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.

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

Calculation evaluation of multiplying properties of LWR with thorium fuel

NASA Astrophysics Data System (ADS)

Shamanin, I. V.; Grachev, V. M.; Knyshev, V. V.; Bedenko, S. V.; Novikova, N. G.

2017-01-01

The results of multiplying properties design research of the unit cell and LWR fuel assembly with the high temperature gas-cooled thorium reactor fuel pellet are presented in the work. The calculation evaluation showed the possibility of using thorium in LWR effectively. In this case the amount of fissile isotope is 2.45 times smaller in comparison with the standard loading of LWR. The research and numerical experiments were carried out using the verified accounting code of the program MCU5, modern libraries of evaluated nuclear data and multigroup approximations.

A Specific Long-Term Plan for Management of U.S. Nuclear Spent Fuel

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

Levy, Salomon

2006-07-01

A specific plan consisting of six different steps is proposed to accelerate and improve the long-term management of U.S. Light Water Reactor (LWR) spent nuclear fuel. The first step is to construct additional, centralized, engineered (dry cask) spent fuel facilities to have a backup solution to Yucca Mountain (YM) delays or lack of capacity. The second step is to restart the development of the Integral Fast Reactor (IFR), in a burner mode, because of its inherent safety characteristics and its extensive past development in contrast to Acceleration Driven Systems (ADS). The IFR and an improved non-proliferation version of its pyro-processingmore » technology can burn the plutonium (Pu) and minor actinides (MA) obtained by reprocessing LWR spent fuel. The remaining IFR and LWR fission products will be treated for storage at YM. The radiotoxicity of that high level waste (HLW) will fall below that of natural uranium in less than one thousand years. Due to anticipated increased capital, maintenance, and research costs for IFR, the third step is to reduce the required number of IFRs and their potential delays by implementing multiple recycles of Pu and Neptunium (Np) MA in LWR. That strategy is to use an advanced separation process, UREX+, and the MIX Pu option where the role and degradation of Pu is limited by uranium enrichment. UREX+ will decrease proliferation risks by avoiding Pu separation while the MIX fuel will lead to an equilibrium fuel recycle mode in LWR which will reduce U. S. Pu inventory and deliver much smaller volumes of less radioactive HLW to YM. In both steps two and three, Research and Development (R and D) is to emphasize the demonstration of multiple fuel reprocessing and fabrication, while improving HLW treatment, increasing proliferation resistance, and reducing losses of fissile material. The fourth step is to license and construct YM because it is needed for the disposal of defense wastes and the HLW to be generated under the proposed plan. The fifth step consists of developing a risk informed methodology to assess the various options available for disposition of LWR spent fuel and to select among them. The sixth step is to modify the current U. S. infrastructure and to create a climate to increase the utilization of uranium and the sustainability of nuclear generated electricity. (author)« 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.

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.

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.

Gas core reactors for actinide transmutation. [uranium hexafluoride

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.; Wan, P. T.; Chow, S.

1979-01-01

The preliminary design of a uranium hexafluoride actinide transmutation reactor to convert long-lived actinide wastes to shorter-lived fission product wastes was analyzed. It is shown that externally moderated gas core reactors are ideal radiators. They provide an abundant supply of thermal neutrons and are insensitive to composition changes in the blanket. For the present reactor, an initial load of 6 metric tons of actinides is loaded. This is equivalent to the quantity produced by 300 LWR-years of operation. At the beginning, the core produces 2000 MWt while the blanket generates only 239 MWt. After four years of irradiation, the actinide mass is reduced to 3.9 metric tonnes. During this time, the blanket is becoming more fissile and its power rapidly approaches 1600 MWt. At the end of four years, continuous refueling of actinides is carried out and the actinide mass is held constant. Equilibrium is essentially achieved at the end of eight years. At equilibrium, the core is producing 1400 MWt and the blanket 1600 MWt. At this power level, the actinide destruction rate is equal to the production rate from 32 LWRs.

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

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

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.

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

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

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.

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

Evaluation of nuclear fuel reprocessing strategies. 2. LWR fuel storage, recycle economics and plutonium logistics

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

Prince, B.E.; Hadley, S.W.

1983-10-27

This is the second of a two-part report intended as a critical review of certain issues involved with closing the Light Water Reactor (LWR) fuel cycle and establishing the basis for future transition to commercial breeder applications. The report is divided into four main sections consisting of (1) a review of the status of the LWR spent fuel management and storage problem; (2) an analysis of the economic incentives for instituting reprocessing and recycle in LWRs; (3) an analysis of the time-dependent aspects of plutonium economic value particularly as related to the LWR-breeder transition; and (4) an analysis of themore » time-dependent aspects of plutonium requirements and supply relative to this transition.« 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.

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

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

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

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

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

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

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.

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.

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

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

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

Blyth, Taylor S.; Avramova, Maria

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics- based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR)more » cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal- hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.« less

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

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

NASA Astrophysics Data System (ADS)

Blyth, Taylor S.

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics-based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal-hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.

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.

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

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.

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.

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

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

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

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

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.

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.

[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.

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

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

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.

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

Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

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

Miao, Yinbin; Ye, Bei; Hofman, Gerard

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U 3Si 2) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U 3Si 2 at LWR conditions. The fission gas behavior of U 3Si 2 can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranularmore » bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U 3Si 2 for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U 3Si 2 at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U 3Si 2 as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.« 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

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.

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

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.

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

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

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

2012-06-06

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spentmore » fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.« 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.

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.

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.

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

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

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

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

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 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.

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

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

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.

Gaseous swelling of U 3 Si 2 during steady-state LWR operation: A rate theory investigation

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

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

Rate theory simulations of fission gas behavior in U 3Si 2 are reported for light water reactor (LWR) steady-state operation scenarios. We developed a model of U 3Si 2 and implemented into the GRASS-SST code based on available research reactor post-irradiation examination (PIE) data, and density functional theory (DFT) calculations of key material properties. Simplified peripheral models were also introduced to capture the fuel-cladding interaction. The simulations identified three regimes of U 3Si 2 swelling behavior between 390 K and 1190 K. Under typical steady-state LWR operating conditions where U 3Si 2 temperature is expected to be below 1000 K,more » intragranular bubbles are dominant and fission gas is retained in those bubbles. The consequent gaseous swelling is low and associated degradation in the fuel thermal conductivity is also limited. Those predictions of U 3Si 2 performance during steady-state operations in LWRs suggest that this fuel material is an appropriate LWR candidate fuel material. Fission gas behavior models established based on this work are being coupled to the thermo-mechanical simulation of the fuel behavior using the BISON fuel performance multi-dimensional finite element code.« less

Gaseous swelling of U 3 Si 2 during steady-state LWR operation: A rate theory investigation

DOE PAGES

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

2017-07-25

Rate theory simulations of fission gas behavior in U 3Si 2 are reported for light water reactor (LWR) steady-state operation scenarios. We developed a model of U 3Si 2 and implemented into the GRASS-SST code based on available research reactor post-irradiation examination (PIE) data, and density functional theory (DFT) calculations of key material properties. Simplified peripheral models were also introduced to capture the fuel-cladding interaction. The simulations identified three regimes of U 3Si 2 swelling behavior between 390 K and 1190 K. Under typical steady-state LWR operating conditions where U 3Si 2 temperature is expected to be below 1000 K,more » intragranular bubbles are dominant and fission gas is retained in those bubbles. The consequent gaseous swelling is low and associated degradation in the fuel thermal conductivity is also limited. Those predictions of U 3Si 2 performance during steady-state operations in LWRs suggest that this fuel material is an appropriate LWR candidate fuel material. Fission gas behavior models established based on this work are being coupled to the thermo-mechanical simulation of the fuel behavior using the BISON fuel performance multi-dimensional finite element code.« less

Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

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

Miao, Yinbin; Mo, Kun; Yacout, Abdellatif

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U 3Si 2) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U 3Si 2 as an AFT for LWRs. Considering the high cost,more » long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U 3Si 2 fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U 3Si 2. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.« 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.

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

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.

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

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

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.

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

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

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

Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

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

Lippek, H.E.; Schuller, C.R.

1979-03-01

A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light mostmore » of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.« less

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

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

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

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.

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

NASA Astrophysics Data System (ADS)

Acır, Adem; Altunok, Taner

2010-10-01

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li20Sn80 coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.

NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL FROM PHWR'S IN A CLOSED THORIUM FUEL CYCLE

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

Sleaford, B W; Collins, B A; Ebbinghaus, B B

2010-04-26

This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that {sup 233}U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date needmore » 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.« less

Nuclear Material Attractiveness: An Assessment of Material from PHWR's in a Closed Thorium Fuel Cycle

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

Sleaford, Brad W.; Ebbinghaus, B. B.; Bradley, Keith S.

2010-06-11

This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies [ , ] that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that 233U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined tomore » date 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.« less

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

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

Gougar, Hans

This document outlines the development of a high fidelity, best estimate nuclear power plant severe transient simulation capability that will complement or enhance the integral system codes historically used for licensing and analysis of severe accidents. As with other tools in the Risk Informed Safety Margin Characterization (RISMC) Toolkit, the ultimate user of Enhanced Severe Transient Analysis and Prevention (ESTAP) capability is the plant decision-maker; the deliverable to that customer is a modern, simulation-based safety analysis capability, applicable to a much broader class of safety issues than is traditional Light Water Reactor (LWR) licensing analysis. Currently, the RISMC pathway’s majormore » emphasis is placed on developing RELAP-7, a next-generation safety analysis code, and on showing how to use RELAP-7 to analyze margin from a modern point of view: that is, by characterizing margin in terms of the probabilistic spectra of the “loads” applied to systems, structures, and components (SSCs), and the “capacity” of those SSCs to resist those loads without failing. The first objective of the ESTAP task, and the focus of one task of this effort, is to augment RELAP-7 analyses with user-selected multi-dimensional, multi-phase models of specific plant components to simulate complex phenomena that may lead to, or exacerbate, severe transients and core damage. Such phenomena include: coolant crossflow between PWR assemblies during a severe reactivity transient, stratified single or two-phase coolant flow in primary coolant piping, inhomogeneous mixing of emergency coolant water or boric acid with hot primary coolant, and water hammer. These are well-documented phenomena associated with plant transients but that are generally not captured in system codes. They are, however, generally limited to specific components, structures, and operating conditions. The second ESTAP task is to similarly augment a severe (post-core damage) accident integral analyses code with high fidelity simulations that would allow investigation of multi-dimensional, multi-phase containment phenomena that are only treated approximately in established codes.« less

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.

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

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

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.

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.

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

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)

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.

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

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

Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model

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

Baudron, Anne-Marie, E-mail: anne-marie.baudron@cea.fr; CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex; Lautard, Jean-Jacques, E-mail: jean-jacques.lautard@cea.fr

2014-12-15

In this paper we present a time-parallel algorithm for the 3D neutrons calculation of a transient model in a nuclear reactor core. The neutrons calculation consists in numerically solving the time dependent diffusion approximation equation, which is a simplified transport equation. The numerical resolution is done with finite elements method based on a tetrahedral meshing of the computational domain, representing the reactor core, and time discretization is achieved using a θ-scheme. The transient model presents moving control rods during the time of the reaction. Therefore, cross-sections (piecewise constants) are taken into account by interpolations with respect to the velocity ofmore » the control rods. The parallelism across the time is achieved by an adequate use of the parareal in time algorithm to the handled problem. This parallel method is a predictor corrector scheme that iteratively combines the use of two kinds of numerical propagators, one coarse and one fine. Our method is made efficient by means of a coarse solver defined with large time step and fixed position control rods model, while the fine propagator is assumed to be a high order numerical approximation of the full model. The parallel implementation of our method provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch–Maurer–Werner benchmark.« less

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

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

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.

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

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

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

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.

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.

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

Continuously improving safety of nuclear installations: An approach to be reinforced after the Fukushima accident

NASA Astrophysics Data System (ADS)

Repussard, Jacques; Schwarz, Michel

2012-05-01

After the Three Mile Island accident in 1979 and the Chernobyl accident in 1986, the Fukushima accident shows that the probability of a core meltdown accident in an LWR (Light Water Reactor) has been largely underestimated. The consequences of such an accident are unacceptable: except in the case of TMI2 (Three Mile Island 2) large areas around the damaged plants are contaminated for decades and populations have to be relocated for long periods. This article presents the French approach which consists in improving continuously the safety of the Nuclear Power Plants (NPP) on the basis of lessons learned from operating experience and from the progress in R&D (Research and Development). It details the key role played by IRSN (Institut de radioprotection et de sûreté nucléaire), the French TSO (Technical and scientific Safety Organization), and shows how the Fukushima accident contributes to this approach in improving NPP robustness. It concludes on the necessity to keep on networking TSOs, to share knowledge as well as R&D resources, with the ultimate goal of enhancing and harmonizing nuclear safety worldwide.

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.

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

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.

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

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.

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

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.

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.

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.

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

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

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

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.

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.

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.

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.

Energy loss due to eddy current in linear transformer driver cores

NASA Astrophysics Data System (ADS)

Kim, A. A.; Mazarakis, M. G.; Manylov, V. I.; Vizir, V. A.; Stygar, W. A.

2010-07-01

In linear transformer drivers [Phys. Rev. ST Accel. Beams 12, 050402 (2009)PRABFM1098-440210.1103/PhysRevSTAB.12.050402; Phys. Rev. ST Accel. Beams 12, 050401 (2009)PRABFM1098-440210.1103/PhysRevSTAB.12.050401] as well as any other linear induction accelerator cavities, ferromagnetic cores are used to prevent the current from flowing along the induction cavity walls which are in parallel with the load. But if the core is made of conductive material, the applied voltage pulse generates the eddy current in the core itself which heats the core and therefore also reduces the overall linear transformer driver (LTD) efficiency. The energy loss due to generation of the eddy current in the cores depends on the specific resistivity of the core material, the design of the core, as well as on the distribution of the eddy current in the core tape during the remagnetizing process. In this paper we investigate how the eddy current is distributed in a core tape with an arbitrary shape hysteresis loop. Our model is based on the textbook knowledge related to the eddy current generation in ferromagnetics with rectangular hysteresis loop, and in usual conductors. For the reader’s convenience, we reproduce some most important details of this knowledge in our paper. The model predicts that the same core would behave differently depending on how fast the applied voltage pulse is: in the high frequency limit, the equivalent resistance of the core reduces during the pulse whereas in the low frequency limit it is constant. An important inference is that the energy loss due to the eddy current generation can be reduced by increasing the cross section of the core over the minimum value which is required to avoid its saturation. The conclusions of the model are confirmed with experimental observations presented at the end of the paper.

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

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...

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

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

Performance of the NASA Digitizing Core-Loss Instrumentation

NASA Technical Reports Server (NTRS)

Schwarze, Gene E. (Technical Monitor); Niedra, Janis M.

2003-01-01

The standard method of magnetic core loss measurement was implemented on a high frequency digitizing oscilloscope in order to explore the limits to accuracy when characterizing high Q cores at frequencies up to 1 MHz. This method computes core loss from the cycle mean of the product of the exciting current in a primary winding and induced voltage in a separate flux sensing winding. It is pointed out that just 20 percent accuracy for a Q of 100 core material requires a phase angle accuracy of 0.1 between the voltage and current measurements. Experiment shows that at 1 MHz, even high quality, high frequency current sensing transformers can introduce phase errors of a degree or more. Due to the fact that the Q of some quasilinear core materials can exceed 300 at frequencies below 100 kHz, phase angle errors can be a problem even at 50 kHz. Hence great care is necessary with current sensing and ground loops when measuring high Q cores. Best high frequency current sensing accuracy was obtained from a fabricated 0.1-ohm coaxial resistor, differentially sensed. Sample high frequency core loss data taken with the setup for a permeability-14 MPP core is presented.

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

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.

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.

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.

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.

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.

Direct-Current Monitor With Flux-Reset Transformer Coupling

NASA Technical Reports Server (NTRS)

Canter, Stanley

1993-01-01

Circuit measures constant or slowly-varying unidirectional electrical current using flux-reset transformer coupling. Measurement nonintrusive in sense that no need for direct contact with wire that carries load current to be measured, and no need to install series resistive element in load-current path. Toroidal magnetic core wrapped with coil of wire placed around load-current-carrying wire, acts as transformer core, load-current-carrying wire acts as primary winding of transformer, and coil wrapped on core acts as secondary winding.

3D Magnetic Field Analysis of a Turbine Generator Stator Core-end Region

NASA Astrophysics Data System (ADS)

Wakui, Shinichi; Takahashi, Kazuhiko; Ide, Kazumasa; Takahashi, Miyoshi; Watanabe, Takashi

In this paper we calculated magnetic flux density and eddy current distributions of a 71MVA turbine generator stator core-end using three-dimensional numerical magnetic field analysis. Subsequently, the magnetic flux densities and eddy current densities in the stator core-end region on the no-load and three-phase short circuit conditions obtained by the analysis have good agreements with the measurements. Furthermore, the differences of eddy current and eddy current loss in the stator core-end region for various load conditions are shown numerically. As a result, the facing had an effect that decrease the eddy current loss of the end plate about 84%.

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

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.

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.

Transformer coupling for transmitting direct current through a barrier

DOEpatents

Brown, Ralph L.; Guilford, Richard P.; Stichman, John H.

1988-01-01

The transmission system for transmitting direct current from an energy source on one side of an electrical and mechanical barrier to a load on the other side of the barrier utilizes a transformer comprising a primary core on one side of the transformer and a secondary core on the other side of the transformer. The cores are magnetically coupled selectively by moving a magnetic ferrite coupler in and out of alignment with the poles of the cores. The direct current from the energy source is converted to a time varying current by an oscillating circuit, which oscillating circuit is optically coupled to a secondary winding on the secondary core to interrupt oscillations upon the voltage in the secondary winding exceeding a preselected level.

Transformer coupling for transmitting direct current through a barrier

DOEpatents

Brown, R.L.; Guilford, R.P.; Stichman, J.H.

1987-06-29

The transmission system for transmitting direct current from an energy source on one side of an electrical and mechanical barrier to a load on the other side of the barrier utilizes a transformer comprising a primary core on one side of the transformer and a secondary core on the other side of the transformer. The cores are magnetically coupled selectively by moving a magnetic ferrite coupler in and out of alignment with the poles of the cores. The direct current from the energy source is converted to a time varying current by an oscillating circuit, which oscillating circuit is optically coupled to a secondary winding on the secondary core to interrupt oscillations upon the voltage in the secondary winding exceeding a preselected level. 4 figs.

Demagnetization using a determined estimated magnetic state

DOEpatents

Denis, Ronald J; Makowski, Nathanael J

2015-01-13

A method for demagnetizing comprising positioning a core within the electromagnetic field generated by a first winding until the generated first electrical current is not substantially increasing, thereby determining a saturation current. A second voltage, having the opposite polarity, is then applied across the first winding until the generated second electrical current is approximately equal to the magnitude of the determined saturation current. The maximum magnetic flux within the core is then determined using the voltage across said first winding and the second current. A third voltage, having the opposite polarity, is then applied across the first winding until the core has a magnetic flux equal to approximately half of the determined maximum magnetic flux within the core.

Parallel Algorithms for Monte Carlo Particle Transport Simulation on Exascale Computing Architectures

NASA Astrophysics Data System (ADS)

Romano, Paul Kollath

Monte Carlo particle transport methods are being considered as a viable option for high-fidelity simulation of nuclear reactors. While Monte Carlo methods offer several potential advantages over deterministic methods, there are a number of algorithmic shortcomings that would prevent their immediate adoption for full-core analyses. In this thesis, algorithms are proposed both to ameliorate the degradation in parallel efficiency typically observed for large numbers of processors and to offer a means of decomposing large tally data that will be needed for reactor analysis. A nearest-neighbor fission bank algorithm was proposed and subsequently implemented in the OpenMC Monte Carlo code. A theoretical analysis of the communication pattern shows that the expected cost is O( N ) whereas traditional fission bank algorithms are O(N) at best. The algorithm was tested on two supercomputers, the Intrepid Blue Gene/P and the Titan Cray XK7, and demonstrated nearly linear parallel scaling up to 163,840 processor cores on a full-core benchmark problem. An algorithm for reducing network communication arising from tally reduction was analyzed and implemented in OpenMC. The proposed algorithm groups only particle histories on a single processor into batches for tally purposes---in doing so it prevents all network communication for tallies until the very end of the simulation. The algorithm was tested, again on a full-core benchmark, and shown to reduce network communication substantially. A model was developed to predict the impact of load imbalances on the performance of domain decomposed simulations. The analysis demonstrated that load imbalances in domain decomposed simulations arise from two distinct phenomena: non-uniform particle densities and non-uniform spatial leakage. The dominant performance penalty for domain decomposition was shown to come from these physical effects rather than insufficient network bandwidth or high latency. The model predictions were verified with measured data from simulations in OpenMC on a full-core benchmark problem. Finally, a novel algorithm for decomposing large tally data was proposed, analyzed, and implemented/tested in OpenMC. The algorithm relies on disjoint sets of compute processes and tally servers. The analysis showed that for a range of parameters relevant to LWR analysis, the tally server algorithm should perform with minimal overhead. Tests were performed on Intrepid and Titan and demonstrated that the algorithm did indeed perform well over a wide range of parameters. (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

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...

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)

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...

Comparison of predicted engine core noise with current and proposed aircraft noise certification requirements

NASA Technical Reports Server (NTRS)

Vonglahn, U. H.; Groesbeck, D. E.

1981-01-01

Predicted engine core noise levels are compared with measured total aircraft noise levels and with current and proposed federal noise certification requirements. Comparisons are made at the FAR-36 measuring stations and include consideration of both full- and cutback-power operation at takeoff. In general, core noise provides a barrier to achieving proposed EPA stage 5 noise levels for all types of aircraft. More specifically, core noise levels will limit further reductions in aircraft noise levels for current widebody commercial aircraft.

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.

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.

Evaluation of the Pool Critical Assembly Benchmark with Explicitly-Modeled Geometry using MCNP6

DOE PAGES

Kulesza, Joel A.; Martz, Roger Lee

2017-03-01

Despite being one of the most widely used benchmarks for qualifying light water reactor (LWR) radiation transport methods and data, no benchmark calculation of the Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA) pressure vessel wall benchmark facility (PVWBF) using MCNP6 with explicitly modeled core geometry exists. As such, this paper provides results for such an analysis. First, a criticality calculation is used to construct the fixed source term. Next, ADVANTG-generated variance reduction parameters are used within the final MCNP6 fixed source calculations. These calculations provide unadjusted dosimetry results using three sets of dosimetry reaction cross sections of varyingmore » ages (those packaged with MCNP6, from the IRDF-2002 multi-group library, and from the ACE-formatted IRDFF v1.05 library). These results are then compared to two different sets of measured reaction rates. The comparison agrees in an overall sense within 2% and on a specific reaction- and dosimetry location-basis within 5%. Except for the neptunium dosimetry, the individual foil raw calculation-to-experiment comparisons usually agree within 10% but is typically greater than unity. Finally, in the course of developing these calculations, geometry that has previously not been completely specified is provided herein for the convenience of future analysts.« less

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

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.

On the recovery of electric currents in the liquid core of the Earth

NASA Astrophysics Data System (ADS)

Kuslits, Lukács; Prácser, Ernő; Lemperger, István

2017-04-01

Inverse geodynamo modelling has become a standard method to get a more accurate image of the processes within the outer core. In this poster excerpts from the preliminary results of an other approach are presented. This comes around the possibility of recovering the currents within the liquid core directly, using Main Magnetic Field data. The approximation of different systems of the flow of charge is possible with various geometries. Based on previous geodynamo simulations, current coils can furnish a good initial geometry for such an estimation. The presentation introduces our preliminary test results and the study of reliability of the applied inversion algorithm for different numbers of coils, distributed in a grid simbolysing the domain between the inner-core and core-mantle boundaries. We shall also present inverted current structures using Main Field model data.

Core/corona modeling of diode-imploded annular loads

NASA Astrophysics Data System (ADS)

Terry, R. E.; Guillory, J. U.

1980-11-01

The effects of a tenuous exterior plasma corona with anomalous resistivity on the compression and heating of a hollow, collisional aluminum z-pinch plasma are predicted by a one-dimensional code. As the interior ("core") plasma is imploded by its axial current, the energy exchange between core and corona determines the current partition. Under the conditions of rapid core heating and compression, the increase in coronal current provides a trade-off between radial acceleration and compression, which reduces the implosion forces and softens the pitch. Combined with a heuristic account of energy and momentum transport in the strongly coupled core plasma and an approximate radiative loss calculation including Al line, recombination and Bremsstrahlung emission, the current model can provide a reasonably accurate description of imploding annular plasma loads that remain azimuthally symmetric. The implications for optimization of generator load coupling are examined.

Analysis and Design of ITER 1 MV Core Snubber

NASA Astrophysics Data System (ADS)

Wang, Haitian; Li, Ge

2012-11-01

The core snubber, as a passive protection device, can suppress arc current and absorb stored energy in stray capacitance during the electrical breakdown in accelerating electrodes of ITER NBI. In order to design the core snubber of ITER, the control parameters of the arc peak current have been firstly analyzed by the Fink-Baker-Owren (FBO) method, which are used for designing the DIIID 100 kV snubber. The B-H curve can be derived from the measured voltage and current waveforms, and the hysteresis loss of the core snubber can be derived using the revised parallelogram method. The core snubber can be a simplified representation as an equivalent parallel resistance and inductance, which has been neglected by the FBO method. A simulation code including the parallel equivalent resistance and inductance has been set up. The simulation and experiments result in dramatically large arc shorting currents due to the parallel inductance effect. The case shows that the core snubber utilizing the FBO method gives more compact design.

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...

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)

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...

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

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

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

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.

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.

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

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

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

Fault current limiter with shield and adjacent cores

DOEpatents

Darmann, Francis Anthony; Moriconi, Franco; Hodge, Eoin Patrick

2013-10-22

In a fault current limiter (FCL) of a saturated core type having at least one coil wound around a high permeability material, a method of suppressing the time derivative of the fault current at the zero current point includes the following step: utilizing an electromagnetic screen or shield around the AC coil to suppress the time derivative current levels during zero current conditions.

PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

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

Smith, P.; Sheetz, S.

Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply. The primary module in the model compares the consumption of uranium for various production reactors against the base case ofmore » Watts Bar I running a nominal load of 1,696 tritium producing burnable absorber rods (TPBARs) with an average refueling of 41,000 kg low enriched uranium (LEU) on an 18 month cycle. After inputting an initial year, starting inventory of unobligated uranium and tritium production forecast, the model will compare and contrast the depletion rate of the LEU between the entered alternatives. This is an annual tritium production rate of approximately 0.059 grams of tritium per kilogram of LEU (g-T/kg-LEU). To date, the Nuclear Regulatory Commission (NRC) license has not been amended to accept a full load of TPBARs so the nominal tritium production has not yet been achieved. The alternatives currently loaded into the model include the three light water SMRs evaluated in TAMU senior projects including, mPower, Holtec and NuScale designs. Initial evaluations of tritium production in light water reactor (LWR) based SMRs using optimized loads TPBARs is on the order 0.02-0.06 grams of tritium per kilogram of LEU used. The TAMU students also chose to model tritium production in the GE-Hitachi SPRISM, a pooltype sodium fast reactor (SFR) utilizing a modified TPBAR type target. The team was unable to complete their project so no data is available. In order to include results from a fast reactor, the SRNL Technical Advisory Committee (TAC) ran a Monte Carlo N-Particle (MCNP) model of a basic SFR for comparison. A 600MWth core surrounded by a lithium blanket produced approximately 1,000 grams of tritium annually with a 13% enriched, 6 year core. This is similar results to a mid-1990’s study where the Fast Flux Test Facility (FFTF), a 400 MWth reactor at the Idaho National Laboratory (INL), could produce about 1,000 grams with an external lithium target. Normalized to the LWRs values, comparative tritium production for an SFR could be approximately 0.31 g-T/kg LEU.« less

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

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

A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis

2015-04-01

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Currentmore » Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.« less

Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

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

Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov

One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example.more » Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the size constraints of the MITR water loop are described. The safety case for operation of the high pressure gas-driven bellows mechanism is also presented. Key issues are the design and response of systems to limit gas flow in the event of a high pressure gas leak in the in-core autoclave. Integrity of the autoclave must be maintained and reactivity effects due to voiding of the loop coolant must be shown to be within the reactor technical specifications. The technical development of the crack growth monitor for application in the INL Advanced Test Reactor or the MITR can act as a template for adaptation of this technology in other reactors. (authors)« less

Plasma current start-up by the outer ohmic heating coils in the Saskatchewan TORus Modified (STOR-M) iron core tokamak

DOE PAGES

Mitarai, O.; Xiao, C.; McColl, D.; ...

2015-03-24

A plasma current up to 15 kA has been driven with outer ohmic heating (OH) coils in the STOR-M iron core tokamak. Even when the inner OH coil is disconnected, the outer OH coils alone can induce the plasma current as primary windings and initial breakdown are even easier in this coil layout. Our results suggest a possibility to use an iron core in a spherical tokamak to start up the plasma current without a central solenoid. Finally, the effect of the iron core saturation on the extension of the discharge pulse length has been estimated for further experiments inmore » the STOR-M tokamak.« less

Plasma current start-up by the outer ohmic heating coils in the Saskatchewan TORus Modified (STOR-M) iron core tokamak

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

Mitarai, O.; Xiao, C.; McColl, D.

A plasma current up to 15 kA has been driven with outer ohmic heating (OH) coils in the STOR-M iron core tokamak. Even when the inner OH coil is disconnected, the outer OH coils alone can induce the plasma current as primary windings and initial breakdown are even easier in this coil layout. Our results suggest a possibility to use an iron core in a spherical tokamak to start up the plasma current without a central solenoid. Finally, the effect of the iron core saturation on the extension of the discharge pulse length has been estimated for further experiments inmore » the STOR-M tokamak.« less

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...

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.

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...

Performance of ARCHITECT HCV core antigen test with specimens from US plasma donors and injecting drug users.

PubMed

Mixson-Hayden, Tonya; Dawson, George J; Teshale, Eyasu; Le, Thao; Cheng, Kevin; Drobeniuc, Jan; Ward, John; Kamili, Saleem

2015-05-01

Hepatitis C virus (HCV) core antigen is a serological marker of current HCV infection. The aim of this study was mainly to evaluate the performance characteristics of the ARCHITECT HCV core antigen assay with specimens from US plasma donors and injecting drug users. A total of 551 serum and plasma samples with known anti-HCV and HCV RNA status were tested for HCV core antigen using the Abbott ARCHITECT HCV core antigen test. HCV core antigen was detectable in 100% of US plasma donor samples collected during the pre-seroconversion phase of infection (anti-HCV negative/HCV RNA positive). Overall sensitivity of the HCV core antigen assay was 88.9-94.3% in samples collected after seroconversion. The correlation between HCV core antigen and HCV RNA titers was 0.959. HCV core antigen testing may be reliably used to identify current HCV infection. Published by Elsevier B.V.

Toroidal-Core Microinductors Biased by Permanent Magnets

NASA Technical Reports Server (NTRS)

Lieneweg, Udo; Blaes, Brent

2003-01-01

The designs of microscopic toroidal-core inductors in integrated circuits of DC-to-DC voltage converters would be modified, according to a proposal, by filling the gaps in the cores with permanent magnets that would apply bias fluxes (see figure). The magnitudes and polarities of the bias fluxes would be tailored to counteract the DC fluxes generated by the DC components of the currents in the inductor windings, such that it would be possible to either reduce the sizes of the cores or increase the AC components of the currents in the cores without incurring adverse effects. Reducing the sizes of the cores could save significant amounts of space on integrated circuits because relative to other integrated-circuit components, microinductors occupy large areas - of the order of a square millimeter each. An important consideration in the design of such an inductor is preventing magnetic saturation of the core at current levels up to the maximum anticipated operating current. The requirement to prevent saturation, as well as other requirements and constraints upon the design of the core are expressed by several equations based on the traditional magnetic-circuit approximation. The equations involve the core and gap dimensions and the magnetic-property parameters of the core and magnet materials. The equations show that, other things remaining equal, as the maximum current is increased, one must increase the size of the core to prevent the flux density from rising to the saturation level. By using a permanent bias flux to oppose the flux generated by the DC component of the current, one would reduce the net DC component of flux in the core, making it possible to reduce the core size needed to prevent the total flux density (sum of DC and AC components) from rising to the saturation level. Alternatively, one could take advantage of the reduction of the net DC component of flux by increasing the allowable AC component of flux and the corresponding AC component of current. In either case, permanent-magnet material and the slant (if any) and thickness of the gap must be chosen according to the equations to obtain the required bias flux. In modifying the design of the inductor, one must ensure that the inductance is not altered. The simplest way to preserve the original value of inductance would be to leave the gap dimensions unchanged and fill the gap with a permanent- magnet material that, fortuitously, would produce just the required bias flux. A more generally applicable alternative would be to partly fill either the original gap or a slightly enlarged gap with a suitable permanent-magnet material (thereby leaving a small residual gap) so that the reluctance of the resulting magnetic circuit would yield the desired inductance.

High Entropy Alloys: A Current Evaluation of Founding Ideas and Core Effects and Exploring Nonlinear Alloys (Postprint)

DTIC Science & Technology

2017-08-29

contain IM phases when using TEM diffraction.1,2 High -Entropy Alloys: A Current Evaluation of Founding Ideas and Core Effects and Exploring ‘‘Nonlinear...obvious outsider. Specifically, an alloy with a high Tm need not contain only elements with high Tm, and it can include one or two elements of moderate or...AFRL-RX-WP-JA-2017-0383 HIGH ENTROPY ALLOYS: A CURRENT EVALUATION OF FOUNDING IDEAS AND CORE EFFECTS AND EXPLORING "NONLINEAR ALLOYS

Parameter Analysis for Arc Snubber of EAST Neutral Beam Injector

NASA Astrophysics Data System (ADS)

Wang, Haitian; Li, Ge; Cao, Liang; Dang, Xiaoqiang; Fu, Peng

2010-08-01

According to the B-H curve and structural dimensions of the snubber by the Fink-Baker Method, the inductive voltage and the eddy current of any core tape with the thickness of the saturated regions are derived when the accelerator breakdown occurs. Using the Ampere's law, in each core tape, the eddy current of the core lamination is equal to the arc current, and the relation of the thickness of the saturated regions for different laminations can be deduced. The total equivalent resistance of the snubber can be obtained. The transient eddy current model based on the stray capacitance and the equivalent resistance is analyzed, and the solving process is given in detail. The exponential time constant and the arc current are obtained. Then, the maximum width of the lamination and the minimum thickness of the core tape are determined. The experimental time constant of the eddy current obtained, with or without the bias current, is approximately the same as that by the analytical method, which proves the accuracy of the adopted assumptions and the analysis method.

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

Health physics aspects of advanced reactor licensing reviews

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

Hinson, C.S.

1995-03-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovativemore » design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.« less

Discussion of examination of a cored hydraulic fracture in a deep gas well

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

Nolte, K.G.

Warpinski et al. document information found from a core through a formation after a hydraulic fracture treatment. As they indicate, the core provides the first detailed evaluation of an actual propped hydraulic fracture away from the well and at a significant depth, and this evaluation leads to findings that deviate substantially from the assumptions incorporated into current fracturing models. In this discussion, a defense of current fracture design assumptions is developed. The affirmation of current assumptions, for general industry applications, is based on an assessment of the global impact of the local complexity found in the core. The assessment leadsmore » to recommendations for the evolution of fracture design practice.« 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.

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...

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

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.

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

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

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.

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

A motional Stark effect diagnostic analysis routine for improved resolution of iota in the core of the large helical device.

PubMed

Dobbins, T J; Ida, K; Suzuki, C; Yoshinuma, M; Kobayashi, T; Suzuki, Y; Yoshida, M

2017-09-01

A new Motional Stark Effect (MSE) analysis routine has been developed for improved spatial resolution in the core of the Large Helical Device (LHD). The routine was developed to reduce the dependency of the analysis on the Pfirsch-Schlüter (PS) current in the core. The technique used the change in the polarization angle as a function of flux in order to find the value of diota/dflux at each measurement location. By integrating inwards from the edge, the iota profile can be recovered from this method. This reduces the results' dependency on the PS current because the effect of the PS current on the MSE measurement is almost constant as a function of flux in the core; therefore, the uncertainty in the PS current has a minimal effect on the calculation of the iota profile. In addition, the VMEC database was remapped from flux into r/a space by interpolating in mode space in order to improve the database core resolution. These changes resulted in a much smoother iota profile, conforming more to the physics expectations of standard discharge scenarios in the core of the LHD.

The first-principle coupled calculations using TMCC and CFX for the pin-wise simulation of LWR

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

Li, L.; Wang, K.

2012-07-01

The coupling of neutronics and thermal-hydraulics plays an important role in the reactor safety, core design and operation of nuclear power facilities. This paper introduces the research on the coupling of Monte Carlo method and CFD method, specifically using TMCC and CFX. The methods of the coupling including the coupling approach, data transfer, mesh mapping and transient coupling scheme are studied firstly. The coupling of TMCC and CFX for the steady state calculations is studied and described for the single rod model and the 3 x 3 Rod Bundle model. The calculation results prove that the coupling method is feasiblemore » and the coupled calculation can be used for steady state calculations. However, the oscillation which occurs during the coupled calculation indicates that this method still needs to be improved for the accuracy. Then the coupling for the transient calculations is also studied and tested by two cases of the steady state and the lost of heat sink. The preliminary results of the transient coupled calculations indicates that the transient coupling with TMCC and CFX is able to simulate the transients but instabilities are occurring. It is also concluded that the transient coupling of TMCC and CFX needs to be improved due to the limitation of computational resource and the difference of time scales. (authors)« less

DIODE STEERED MANGETIC-CORE MEMORY

DOEpatents

Melmed, A.S.; Shevlin, R.T.; Laupheimer, R.

1962-09-18

A word-arranged magnetic-core memory is designed for use in a digital computer utilizing the reverse or back current property of the semi-conductor diodes to restore the information in the memory after read-out. In order to ob tain a read-out signal from a magnetic core storage unit, it is necessary to change the states of some of the magnetic cores. In order to retain the information in the memory after read-out it is then necessary to provide a means to return the switched cores to their states before read-out. A rewrite driver passes a pulse back through each row of cores in which some switching has taken place. This pulse combines with the reverse current pulses of diodes for each column in which a core is switched during read-out to cause the particular cores to be switched back into their states prior to read-out. (AEC)

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.

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

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

Core Training in Low Back Disorders: Role of the Pilates Method.

PubMed

Joyce, Andrew A; Kotler, Dana H

The Pilates method is a system of exercises developed by Joseph Pilates, which emphasizes recruitment and strengthening of the core muscles, flexibility, and breathing, to promote stability and control of movement. Its focus bears similarity to current evidence-based exercise programs for low back disorders. Spinal stability is a function of three interdependent systems, osseoligamentous, muscular, and neural control; exercise addresses both the muscular and neural function. The "core" typically refers to the muscular control required to maintain functional stability. Prior research has highlighted the importance of muscular strength and recruitment, with debate over the importance of individual muscles in the wider context of core control. Though developed long before the current evidence, the Pilates method is relevant in this setting and clearly relates to current evidence-based exercise interventions. Current literature supports the Pilates method as a treatment for low back disorders, but its benefit when compared with other exercise is less clear.

Light Water Reactor Sustainability Program, U.S. Efforts in Support of Examinations at Fukushima Daiichi-2017 Evaluations

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

Farmer, Mitchell T.

Although 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 limited full scale prototypic data.more » 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, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include FY2017 information, summarizes results from U.S. efforts to use information obtained by TEPCO Holdings to enhance the safety of existing and future nuclear power plant designs. This effort, which was initiated in 2014 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 U.S. experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO Holdings information from Daiichi that address these needs. Each year, annual reports include examples demonstrating 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. Highlights in this FY2017 report include new insights with respect to the forces required to produce the observed Daiichi Unit 1 (1F1) shield plug endstate, the observed leakage from 1F1 components, and the amount of combustible gas generation required to produce the observed explosions in Daiichi Units 3 and 4 (1F3 and 1F4). This report contains an appendix with a list of examination needs that was updated after U.S. experts reviewed recently obtained information from examinations at Daiichi. Additional details for higher priority, near-term, examination activities are also provided. This report also includes an appendix with a description of an updated website that has been reformatted to better assist U.S. experts by providing information in an archived retrievable location, as well as an appendix summarizing U.S. Forensics activities to host the TMI-2 Knowledge Transfer and Relevance to Fukushima Meeting that was held in Idaho Falls, ID, on October 10-14, 2016.« less

Nanolaminated Permalloy Core for High-Flux, High-Frequency Ultracompact Power Conversion

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

Kim, J; Kim, M; Galle, P

2013-09-01

Metallic magnetic materials have desirable magnetic properties, including high permeability, and high saturation flux density, when compared with their ferrite counterparts. However, eddy-current losses preclude their use in many switching converter applications, due to the challenge of simultaneously achieving sufficiently thin laminations such that eddy currents are suppressed (e.g., 500 nm-1 mu m for megahertz frequencies), while simultaneously achieving overall core thicknesses such that substantial power can be handled. A CMOS-compatible fabrication process based on robot-assisted sequential electrodeposition followed by selective chemical etching has been developed for the realization of a core of substantial overall thickness (tens to hundreds ofmore » micrometers) comprised of multiple, stacked permalloy (Ni80Fe20) nanolaminations. Tests of toroidal inductors with nanolaminated cores showed negligible eddy-current loss relative to total core loss even at a peak flux density of 0.5 T in the megahertz frequency range. To illustrate the use of these cores, a buck power converter topology is implemented with switching frequencies of 1-2 MHz. Power conversion efficiency greater than 85% with peak operating flux density of 0.3-0.5 T in the core and converter output power level exceeding 5 W was achieved.« less

Core Today! Rationale and Implications. Revised Edition.

ERIC Educational Resources Information Center

Vars, Gordon, Ed.; Larson, Craig, Ed.

This pamphlet is designed to help educators apply the core concept to current problems and situations in educational settings. The preface establishes the position of the National Association for Core Curriculum. A definition of the core curriculum concept is stated in the introduction. Ten assumptions and beliefs on which the core concept is…

Phase Equilibrium Experiments on Potential Lunar Core Compositions: Extension of Current Knowledge to Multi-Component (Fe-Ni-Si-S-C) Systems

NASA Technical Reports Server (NTRS)

Righter, K.; Pando, K.; Danielson, L.

2014-01-01

Numerous geophysical and geochemical studies have suggested the existence of a small metallic lunar core, but the composition of that core is not known. Knowledge of the composition can have a large impact on the thermal evolution of the core, its possible early dynamo creation, and its overall size and fraction of solid and liquid. Thermal models predict that the current temperature at the core-mantle boundary of the Moon is near 1650 K. Re-evaluation of Apollo seismic data has highlighted the need for new data in a broader range of bulk core compositions in the PT range of the lunar core. Geochemical measurements have suggested a more volatile-rich Moon than previously thought. And GRAIL mission data may allow much better constraints on the physical nature of the lunar core. All of these factors have led us to determine new phase equilibria experimental studies in the Fe-Ni-S-C-Si system in the relevant PT range of the lunar core that will help constrain the composition of Moon's core.

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)

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

DC-Compensated Current Transformer.

PubMed

Ripka, Pavel; Draxler, Karel; Styblíková, Renata

2016-01-20

Instrument current transformers (CTs) measure AC currents. The DC component in the measured current can saturate the transformer and cause gross error. We use fluxgate detection and digital feedback compensation of the DC flux to suppress the overall error to 0.15%. This concept can be used not only for high-end CTs with a nanocrystalline core, but it also works for low-cost CTs with FeSi cores. The method described here allows simultaneous measurements of the DC current component.

Improved Thermoplastic/Iron-Particle Transformer Cores

NASA Technical Reports Server (NTRS)

Wincheski, Russell A.; Bryant, Robert G.; Namkung, Min

2004-01-01

A method of fabricating improved transformer cores from composites of thermoplastic matrices and iron-particles has been invented. Relative to commercially available laminated-iron-alloy transformer cores, the cores fabricated by this method weigh less and are less expensive. Relative to prior polymer-matrix/ iron-particle composite-material transformer cores, the cores fabricated by this method can be made mechanically stronger and more magnetically permeable. In addition, whereas some prior cores have exhibited significant eddy-current losses, the cores fabricated by this method exhibit very small eddy-current losses. The cores made by this method can be expected to be attractive for use in diverse applications, including high-signal-to-noise transformers, stepping motors, and high-frequency ignition coils. The present method is a product of an experimental study of the relationships among fabrication conditions, final densities of iron particles, and mechanical and electromagnetic properties of fabricated cores. Among the fabrication conditions investigated were molding pressures (83, 104, and 131 MPa), and molding temperatures (250, 300, and 350 C). Each block of core material was made by uniaxial-compression molding, at the applicable pressure/temperature combination, of a mixture of 2 weight percent of LaRC (or equivalent high-temperature soluble thermoplastic adhesive) with 98 weight percent of approximately spherical iron particles having diameters in the micron range. Each molded block was cut into square cross-section rods that were used as core specimens in mechanical and electromagnetic tests. Some of the core specimens were annealed at 900 C and cooled slowly before testing. For comparison, a low-carbon-steel core was also tested. The results of the tests showed that density, hardness, and rupture strength generally increased with molding pressure and temperature, though the correlation was rather weak. The weakness of the correlation was attributed to the pores in the specimens. The maximum relative permeabilities of cores made without annealing ranged from 30 to 110, while those of cores made with annealing ranged from 900 to 1,400. However, the greater permeabilities of the annealed specimens were not associated with noticeably greater densities. The major practical result of the investigation was the discovery of an optimum distribution of iron-particle sizes: It was found that eddy-current losses in the molded cores were minimized by using 100 mesh (corresponding to particles with diameters less than or equal to 100 m) iron particles. The effect of optimization of particle sizes on eddy-current losses is depicted in the figure.

Detecting the position of the moving-iron solenoid by non-displacement sensor based on parameter identification of flux linkage characteristics

NASA Astrophysics Data System (ADS)

Wang, Xuping; Quan, Long; Xiong, Guangyu

2013-11-01

Currently, most researches use signals, such as the coil current or voltage of solenoid, to identify parameters; typically, parameter identification method based on variation rate of coil current is applied for position estimation. The problem exists in these researches that the detected signals are prone to interference and difficult to obtain. This paper proposes a new method for detecting the core position by using flux characteristic quantity, which adds a new group of secondary winding to the coil of the ordinary switching electromagnet. On the basis of electromagnetic coupling theory analysis and simulation research of the magnetic field regarding the primary and secondary winding coils, and in accordance with the fact that under PWM control mode varying core position and operating current of windings produce different characteristic of flux increment of the secondary winding. The flux increment of the electromagnet winding can be obtained by conducting time domain integration for the induced voltage signal of the extracted secondary winding, and the core position from the two-dimensional fitting curve of the operating winding current and flux-linkage characteristic quantity of solenoid are calculated. The detecting and testing system of solenoid core position is developed based on the theoretical research. The testing results show that the flux characteristic quantity of switching electromagnet magnetic circuit is able to effectively show the core position and thus to accomplish the non-displacement transducer detection of the said core position of the switching electromagnet. This paper proposes a new method for detecting the core position by using flux characteristic quantity, which provides a new theory and method for switch solenoid to control the proportional valve.

DC-Compensated Current Transformer †

PubMed Central

Ripka, Pavel; Draxler, Karel; Styblíková, Renata

2016-01-01

Instrument current transformers (CTs) measure AC currents. The DC component in the measured current can saturate the transformer and cause gross error. We use fluxgate detection and digital feedback compensation of the DC flux to suppress the overall error to 0.15%. This concept can be used not only for high-end CTs with a nanocrystalline core, but it also works for low-cost CTs with FeSi cores. The method described here allows simultaneous measurements of the DC current component. PMID:26805830

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.

Nuclear Neutrino Spectra in Late Stellar Evolution

NASA Astrophysics Data System (ADS)

Misch, G. Wendell; Sun, Yang; Fuller, George

2018-05-01

Neutrinos are the principle carriers of energy in massive stars, beginning from core carbon burning and continuing through core collapse and after the core bounce. In fact, it may be possible to detect neutrinos from nearby pre-supernova stars. Therefore, it is of great interest to understand the neutrino energy spectra from these stars. Leading up to core collapse, beginning around core silicon burning, nuclei become dominant producers of neutrinos, particularly at high neutrino energy, so a systematic study of nuclear neutrino spectra is desirable. We have done such a study, and we present our sd-shell model calculations of nuclear neutrino energy spectra for nuclei in the mass number range A = 21 - 35. Our study includes neutrinos produced by charged lepton capture, charged lepton emission, and neutral current nuclear deexcitation. Previous authors have tabulated the rates of charged current nuclear weak interactions in astrophysical conditions, but the present work expands on this not only by providing neutrino energy spectra, but also by including the heretofore untabulated neutral current de-excitation neutrino pairs.

Core Journal Lists: Classic Tool, New Relevance

ERIC Educational Resources Information Center

Paynter, Robin A.; Jackson, Rose M.; Mullen, Laura Bowering

2010-01-01

Reviews the historical context of core journal lists, current uses in collection assessment, and existing methodologies for creating lists. Outlines two next generation core list projects developing new methodologies and integrating novel information/data sources to improve precision: a national-level core psychology list and the other a local…

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

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.

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

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

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

Transfluxor circuit amplifies sensing current for computer memories

NASA Technical Reports Server (NTRS)

Milligan, G. C.

1964-01-01

To transfer data from the magnetic memory core to an independent core, a reliable sensing amplifier has been developed. Later the data in the independent core is transferred to the arithmetical section of the computer.

Multi-Core Processor Memory Contention Benchmark Analysis Case Study

NASA Technical Reports Server (NTRS)

Simon, Tyler; McGalliard, James

2009-01-01

Multi-core processors dominate current mainframe, server, and high performance computing (HPC) systems. This paper provides synthetic kernel and natural benchmark results from an HPC system at the NASA Goddard Space Flight Center that illustrate the performance impacts of multi-core (dual- and quad-core) vs. single core processor systems. Analysis of processor design, application source code, and synthetic and natural test results all indicate that multi-core processors can suffer from significant memory subsystem contention compared to similar single-core processors.

Lithography-Free Fabrication of Core-Shell GaAs Nanowire Tunnel Diodes.

PubMed

Darbandi, A; Kavanagh, K L; Watkins, S P

2015-08-12

GaAs core-shell p-n junction tunnel diodes were demonstrated by combining vapor-liquid-solid growth with gallium oxide deposition by atomic layer deposition for electrical isolation. The characterization of an ensemble of core-shell structures was enabled by the use of a tungsten probe in a scanning electron microscope without the need for lithographic processing. Radial tunneling transport was observed, exhibiting negative differential resistance behavior with peak-to-valley current ratios of up to 3.1. Peak current densities of up to 2.1 kA/cm(2) point the way to applications in core-shell photovoltaics and tunnel field effect transistors.

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

Fast Heating of Imploded Core with Counterbeam Configuration.

PubMed

Mori, Y; Nishimura, Y; Hanayama, R; Nakayama, S; Ishii, K; Kitagawa, Y; Sekine, T; Sato, N; Kurita, T; Kawashima, T; Kan, H; Komeda, O; Nishi, T; Azuma, H; Hioki, T; Motohiro, T; Sunahara, A; Sentoku, Y; Miura, E

2016-07-29

A tailored-pulse-imploded core with a diameter of 70  μm is flashed by counterirradiating 110 fs, 7 TW laser pulses. Photon emission (>40  eV) from the core exceeds the emission from the imploded core by 6 times, even though the heating pulse energies are only one seventh of the implosion energy. The coupling efficiency from the heating laser to the core using counterirradiation is 14% from the enhancement of photon emission. Neutrons are also produced by counterpropagating fast deuterons accelerated by the photon pressure of the heating pulses. A collisional two-dimensional particle-in-cell simulation reveals that the collisionless two counterpropagating fast-electron currents induce mega-Gauss magnetic filaments in the center of the core due to the Weibel instability. The counterpropagating fast-electron currents are absolutely unstable and independent of the core density and resistivity. Fast electrons with energy below a few MeV are trapped by these filaments in the core region, inducing an additional coupling. This might lead to the observed bright photon emissions.

Chiral vortical effect generated by chiral anomaly in vortex-skyrmions

NASA Astrophysics Data System (ADS)

Volovik, G. E.

2017-03-01

We discuss the type of the general macroscopic parity-violating effects, when there is the current along the vortex, which is concentrated in the vortex core. We consider vortices in chiral superfluids with Weyl points. In the vortex core, the positions of the Weyl points form the skyrmion structure. We show that the mass current concentrated in such a core is provided by the spectral flow through the Weyl points according to the Adler-Bell-Jackiw equation for chiral anomaly.

Development and validation of a low-frequency modeling code for high-moment transmitter rod antennas

NASA Astrophysics Data System (ADS)

Jordan, Jared Williams; Sternberg, Ben K.; Dvorak, Steven L.

2009-12-01

The goal of this research is to develop and validate a low-frequency modeling code for high-moment transmitter rod antennas to aid in the design of future low-frequency TX antennas with high magnetic moments. To accomplish this goal, a quasi-static modeling algorithm was developed to simulate finite-length, permeable-core, rod antennas. This quasi-static analysis is applicable for low frequencies where eddy currents are negligible, and it can handle solid or hollow cores with winding insulation thickness between the antenna's windings and its core. The theory was programmed in Matlab, and the modeling code has the ability to predict the TX antenna's gain, maximum magnetic moment, saturation current, series inductance, and core series loss resistance, provided the user enters the corresponding complex permeability for the desired core magnetic flux density. In order to utilize the linear modeling code to model the effects of nonlinear core materials, it is necessary to use the correct complex permeability for a specific core magnetic flux density. In order to test the modeling code, we demonstrated that it can accurately predict changes in the electrical parameters associated with variations in the rod length and the core thickness for antennas made out of low carbon steel wire. These tests demonstrate that the modeling code was successful in predicting the changes in the rod antenna characteristics under high-current nonlinear conditions due to changes in the physical dimensions of the rod provided that the flux density in the core was held constant in order to keep the complex permeability from changing.

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

Core Characteristics Deterioration due to Plastic Deformation

NASA Astrophysics Data System (ADS)

Kaido, Chikara; Arai, Satoshi

This paper discusses the effect of plastic deformation at core manufacturing on the characteristics of cores where non-oriented electrical steel sheets are used as core material. Exciting field and iron loss increase proportionally to plastic deformation in the case of rP<10, where rP is a ratio of plastic deformation to that at yield point. In this region, anomalous eddy currents increase because plastic deformations of crystalline grains are distributed and then the flux distribution is induced. In the case of rP>20, the deterioration tend to saturate, and the increases in magnetic field and iron loss are 1000 to 1500A/m and 2 to 4W/kg. They are related to grain size, and high grade with larger grain may have lager field increase and smaller iron loss increase. Anomalous eddy current losses scarcely increase in this region. In actual motors, the plastic deformation affects iron loss increase although exciting current increases a little.

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

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

Human-In-The-Loop Simulation in Support of Long-Term Sustainability of Light Water Reactors

DOE PAGES

Hallbert, Bruce P

2015-01-01

Reliable instrumentation, information, and control 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. The NPP owners and operators realize that this analog technology represents a significant challenge to sustaining the operation of the current fleet of NPPs. Beyond control systems, new technologies are neededmore » to monitor and characterize the effects of aging and degradation in critical areas of key structures, systems, and components. The objective of the efforts sponsored by the U.S. Department of Energy is to develop, demonstrate, and deploy new digital technologies for II&C architectures and provide monitoring capabilities to ensure the continued safe, reliable, and economic operation of the nation’s NPPs.« less

Potential advantages associated with implementing a risk-based inspection program by a nuclear facility

NASA Astrophysics Data System (ADS)

McNeill, Alexander, III; Balkey, Kenneth R.

1995-05-01

The current inservice inspection activities at a U.S. nuclear facility are based upon the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI. The Code selects examination locations based upon a sampling criteria which includes component geometry, stress, and usage among other criteria. This can result in a significant number of required examinations. As a result of regulatory action each nuclear facility has conducted probabilistic risk assessments (PRA) or individual plant examinations (IPE), producing plant specific risk-based information. Several initiatives have been introduced to apply this new plant risk information. Among these initiatives is risk-based inservice inspection. A code case has been introduced for piping inspections based upon this new risk- based technology. This effort brought forward to the ASME Section XI Code committee, has been initiated and championed by the ASME Research Task Force on Risk-Based Inspection Guidelines -- LWR Nuclear Power Plant Application. Preliminary assessments associated with the code case have revealed that potential advantages exist in a risk-based inservice inspection program with regard to a number of exams, risk, personnel exposure, and cost.

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.

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

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

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).

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.

Evaluation of Analysis Techniques for Fluted-Core Sandwich Cylinders

NASA Technical Reports Server (NTRS)

Lovejoy, Andrew E.; Schultz, Marc R.

2012-01-01

Buckling-critical launch-vehicle structures require structural concepts that have high bending stiffness and low mass. Fluted-core, also known as truss-core, sandwich construction is one such concept. In an effort to identify an analysis method appropriate for the preliminary design of fluted-core cylinders, the current paper presents and compares results from several analysis techniques applied to a specific composite fluted-core test article. The analysis techniques are evaluated in terms of their ease of use and for their appropriateness at certain stages throughout a design analysis cycle (DAC). Current analysis techniques that provide accurate determination of the global buckling load are not readily applicable early in the DAC, such as during preliminary design, because they are too costly to run. An analytical approach that neglects transverse-shear deformation is easily applied during preliminary design, but the lack of transverse-shear deformation results in global buckling load predictions that are significantly higher than those from more detailed analysis methods. The current state of the art is either too complex to be applied for preliminary design, or is incapable of the accuracy required to determine global buckling loads for fluted-core cylinders. Therefore, it is necessary to develop an analytical method for calculating global buckling loads of fluted-core cylinders that includes transverse-shear deformations, and that can be easily incorporated in preliminary design.

Single coil bistable, bidirectional micromechanical actuator

DOEpatents

Tabat, Ned; Guckel, Henry

1998-09-15

Micromechanical actuators capable of bidirectional and bistable operation can be formed on substrates using lithographic processing techniques. Bistable operation of the microactuator is obtained using a single coil and a magnetic core with a gap. A plunger having two magnetic heads is supported for back and forth linear movement with respect to the gap in the magnetic core, and is spring biased to a neutral position in which the two heads are on each side of the gap in the core. The single electrical coil is coupled to the core and is provided with electrical current to attract one of the heads toward the core by reluctance action to drive the plunger to a limit of travel in one direction. The current is then cut off and the plunger returns by spring action toward the gap, whereafter the current is reapplied to the coil to attract the other head of the plunger by reluctance action to drive the plunger to its other limit of travel. This process can be repeated at a time when switching of the actuator is required.

Infrasonic acoustic waves generated by fast air heating in sprite cores

NASA Astrophysics Data System (ADS)

Silva, Caitano L.; Pasko, Victor P.

2014-03-01

Acceleration, expansion, and branching of sprite streamers can lead to concentration of high electrical currents in regions of space, that are observed in the form of bright sprite cores. Driven by this electrical current, a series of chemical processes take place in the sprite plasma. Excitation, followed by quenching of excited electronic states leads to energy transfer from charged to neutral species. The consequence is heating and expansion of air leading to emission of infrasonic acoustic waves. Results indicate that ≳0.01 Pa pressure perturbations on the ground, observed in association with sprites, can only be produced by exceptionally strong currents in sprite cores, exceeding 2 kA.

77 FR 47069 - Proposed Collection; Comment Request

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-07

... currently approved information collection known as ``Federal Home Loan Bank Acquired Member Assets, Core...; Comment Request: Federal Home Loan Bank Acquired Member Assets, Core Mission Activities, Investments and... Collection; Comment Request: Federal Home Loan Bank Acquired Member Assets, Core Mission Activities...

Eddy current position indicating apparatus for measuring displacements of core components of a liquid metal nuclear reactor

DOEpatents

Day, Clifford K.; Stringer, James L.

1977-01-01

Apparatus for measuring displacements of core components of a liquid metal fast breeder reactor by means of an eddy current probe. The active portion of the probe is located within a dry thimble which is supported on a stationary portion of the reactor core support structure. Split rings of metal, having a resistivity significantly different than sodium, are fixedly mounted on the core component to be monitored. The split rings are slidably positioned around, concentric with the probe and symmetrically situated along the axis of the probe so that motion of the ring along the axis of the probe produces a proportional change in the probes electrical output.

Making the case for high temperature low sag (htls) overhead transmission line conductors

NASA Astrophysics Data System (ADS)

Banerjee, Koustubh

The future grid will face challenges to meet an increased power demand by the consumers. Various solutions were studied to address this issue. One alternative to realize increased power flow in the grid is to use High Temperature Low Sag (HTLS) since it fulfills essential criteria of less sag and good material performance with temperature. HTLS conductors like Aluminum Conductor Composite Reinforced (ACCR) and Aluminum Conductor Carbon Composite (ACCC) are expected to face high operating temperatures of 150-200 degree Celsius in order to achieve the desired increased power flow. Therefore, it is imperative to characterize the material performance of these conductors with temperature. The work presented in this thesis addresses the characterization of carbon composite core based and metal matrix core based HTLS conductors. The thesis focuses on the study of variation of tensile strength of the carbon composite core with temperature and the level of temperature rise of the HTLS conductors due to fault currents cleared by backup protection. In this thesis, Dynamic Mechanical Analysis (DMA) was used to quantify the loss in storage modulus of carbon composite cores with temperature. It has been previously shown in literature that storage modulus is correlated to the tensile strength of the composite. Current temperature relationships of HTLS conductors were determined using the IEEE 738-2006 standard. Temperature rise of these conductors due to fault currents were also simulated. All simulations were performed using Microsoft Visual C++ suite. Tensile testing of metal matrix core was also performed. Results of DMA on carbon composite cores show that the storage modulus, hence tensile strength, decreases rapidly in the temperature range of intended use. DMA on composite cores subjected to heat treatment were conducted to investigate any changes in the variation of storage modulus curves. The experiments also indicates that carbon composites cores subjected to temperatures at or above 250 degree Celsius can cause permanent loss of mechanical properties including tensile strength. The fault current temperature analysis of carbon composite based conductors reveal that fault currents eventually cleared by backup protection in the event of primary protection failure can cause damage to fiber matrix interface.

Numerical simulation and experimental study of heat-fluid-solid coupling of double flapper-nozzle servo valve

NASA Astrophysics Data System (ADS)

Zhao, Jianhua; Zhou, Songlin; Lu, Xianghui; Gao, Dianrong

2015-09-01

The double flapper-nozzle servo valve is widely used to launch and guide the equipment. Due to the large instantaneous flow rate of servo valve working under specific operating conditions, the temperature of servo valve would reach 120°C and the valve core and valve sleeve deform in a short amount of time. So the control precision of servo valve significantly decreases and the clamping stagnation phenomenon of valve core appears. In order to solve the problem of degraded control accuracy and clamping stagnation of servo valve under large temperature difference circumstance, the numerical simulation of heat-fluid-solid coupling by using finite element method is done. The simulation result shows that zero position leakage of servo valve is basically impacted by oil temperature and change of fit clearance. The clamping stagnation is caused by warpage-deformation and fit clearance reduction of the valve core and valve sleeve. The distribution rules of the temperature and thermal-deformation of shell, valve core and valve sleeve and the pressure, velocity and temperature field of flow channel are also analyzed. Zero position leakage and electromagnet's current when valve core moves in full-stroke are tested using Electro-hydraulic Servo-valve Characteristic Test-bed of an aerospace sciences and technology corporation. The experimental results show that the change law of experimental current at different oil temperatures is roughly identical to simulation current. The current curve of the electromagnet is smooth when oil temperature is below 80°C, but the amplitude of current significantly increases and the hairy appears when oil temperature is above 80°C. The current becomes smooth again after the warped valve core and valve sleeve are reground. It indicates that clamping stagnation is caused by warpage-deformation and fit clearance reduction of valve core and valve sleeve. This paper simulates and tests the heat-fluid-solid coupling of double flapper-nozzle servo valve, and the obtained results provide the reference value for the design of double flapper-nozzle force feedback servo valve.

Hyperactivity in Boys with Attention-Deficit/Hyperactivity Disorder (ADHD): A Ubiquitous Core Symptom or Manifestation of Working Memory Deficits?

ERIC Educational Resources Information Center

Rapport, Mark D.; Bolden, Jennifer; Kofler, Michael J.; Sarver, Dustin E.; Raiker, Joseph S.; Alderson, R. Matt

2009-01-01

Hyperactivity is currently considered a core and ubiquitous feature of attention-deficit/hyperactivity disorder (ADHD); however, an alternative model challenges this premise and hypothesizes a functional relationship between working memory (WM) and activity level. The current study investigated whether children's activity level is functionally…

A Literature Review and Analysis of Mode Deactivation Therapy

ERIC Educational Resources Information Center

Apsche, Jack A.

2010-01-01

This article is a review of articles, chapters and current research examining Mode Deactivation Therapy. Current applications of MDT suggest that mindfulness is a core component of MDT, as well as acceptance, defusion and validation, clarification and redirection of the functional alternative beliefs. These components are the core of MDT and a…

Phase equilibria of a low S and C lunar core: Implications for an early lunar dynamo and physical state of the current core

NASA Astrophysics Data System (ADS)

Righter, K.; Go, B. M.; Pando, K. A.; Danielson, L.; Ross, D. K.; Rahman, Z.; Keller, L. P.

2017-04-01

Multiple lines of geochemical and geophysical evidence suggest the Moon has a small metallic core, yet the composition of the core is poorly constrained. The physical state of the core (now or in the past) depends on detailed knowledge of its composition, and unfortunately, there is little available data on relevant multicomponent systems (i.e., Fe-Ni-S-C) at lunar interior conditions. In particular, there is a dearth of phase equilibrium data to elucidate whether a specific core composition could help to explain an early lunar geodynamo and magnetic field intensities, or current solid inner core/liquid outer core states. We utilize geochemical information to estimate the Ni, S and C contents of the lunar core, and then carry out phase equilibria experiments on several possible core compositions at the pressure and temperature conditions relevant to the lunar interior. The first composition is 0.5 wt% S and 0.375 wt% C, based on S and C contents of Apollo glasses. A second composition contains 1 wt% each of S and C, and assumes that the lunar mantle experienced degassing of up to 50% of its S and C. Finally a third composition contains C as the dominant light element. Phase equilibrium experiments were completed at 1, 3 and 5 GPa, using piston cylinder and multi-anvil techniques. The first composition has a liquidus near 1550 °C and solidus near 1250 °C. The second composition has a narrower liquidus and solidus temperatures of 1400 and 1270 °C, respectively, while the third composition is molten down to 1150 °C. As the composition crystallizes, the residual liquid becomes enriched in S and C, but S enrichment is greater due to the incorporation of C (but not S) into solid metallic FeNi. Comparison of these results to thermal models for the Moon allow an evaluation of which composition is consistent with the geophysical data of an early dynamo and a currently solid inner and liquid outer core. Composition 1 has a high enough liquidus to start crystallizing early in lunar history (4.3 Ga), consistent with the possible core dynamo initiated by crystallization of a solid inner core. Composition 1 also stays partially molten throughout lunar history, and could easily explain the seismic data. Composition 2, on the other hand, can satisfy one or the other set of geophysical data, but not both and thus seems like a poor candidate for a lunar core composition. Composition 3 remains molten to temperatures that are lower than current estimates for the lunar core, thus ruling out the possibility of a C-rich (and S-poor) lunar core. The S- and C-poor core composition studied here (composition 1) is consistent with all available geochemical and geophysical data and provides a simple heat source and mechanism for a lunar core dynamo (core crystallization) that would obviate the need for other primary mechanisms such as impacts, core-mantle coupling, or unusual thermal histories.

Electrodeposited Nanolaminated CoNiFe Cores for Ultracompact DC-DC Power Conversion

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

Kim, J; Kim, M; Herrault, F

2015-09-01

Laminated metallic alloy cores (i.e., alternating layers of thin film metallic alloy and insulating material) of appropriate lamination thickness enable suppression of eddy current losses at high frequencies. Magnetic cores comprised of many such laminations yield substantial overall magnetic volume, thereby enabling high-power operation. Previously, we reported nanolaminated permalloy (Ni-80 Fe-20) cores based on a sequential electrodeposition technique, demonstrating negligible eddy current losses at peak flux densities up to 0.5 T and operating at megahertz frequencies. This paper demonstrates improved performance of nanolaminated cores comprising tens to hundreds of layers of 300-500-nm-thick CoNiFe films that exhibit superior magnetic properties (e.g.,more » higher saturation flux density and lower coercivity) than permalloy. Nanolaminated CoNiFe cores can be operated up to a peak flux density of 0.9 T, demonstrating improved power handling capacity and exhibiting 30% reduced volumetric core loss, attributed to lowered hysteresis losses compared to the nanolaminated permalloy core of the same geometry. Operating these cores in a buck dc-dc power converter at a switching frequency of 1 MHz, the nanolaminated CoNiFe cores achieved a conversion efficiency exceeding 90% at output power levels up to 7 W, compared to an achieved permalloy core conversion efficiency below 86% at 6 W.« less

Sedimentation of the mud belt along the coast of China from the mouth of the Yangtze (Changjiang) River to northern Taiwan Strait: An Source-to-Sink Perspective

NASA Astrophysics Data System (ADS)

Chien, C. C.; Liu, J. T.; Yang, R.; Huh, C. A.; Su, C. C.

2016-02-01

Sediments in the Taiwan Strait are originated from Mainland China and Taiwan. The China Coastal Current, influenced by the northeast monsoon in winter, becomes enhanced, which caries the sediments exported from the Yangtze River to the southern East China Sea and the Taiwan Strait along the Zhemin-Taiwan Strait mud belt. The sediment transport process is also influenced by tidal current and Kuroshio Branch Current and Taiwan Warm Current, making the seafloor sediment signals complex. This study used R/V Ocean Researcher V (Cruise 0032), to collect six box cores and three gravity cores along the Zhemin mud belt and the mud belt in northern Taiwan Strait in the winter of 2014. From the core samples, grain-size distribution, Multi-Sensor Core Logger, and 7Be activity were measured to investigate the sedimentation process along the mud belts. The box core taken at the mouth of the Changjiang- is composed of homogeneous clay and rich in shell fragments. The core off the mouth of Ou River is composed of homogeneous clay, but showing horizontal laminations. Near the Taishan Island off the coast of Zhejiang the core is consisted of a homogeneous sandy sediments that turned into clay. Off the mouth of the Min River the core consists of clay with shell fragments. Off the coast of the Wu River on the west coast of the Taiwan, the core is mainly composed of muddy sediments, which has the siltstone layers of oblique bedding. Off the mouth of Zhuoshui River in central Taiwan, the core is composed of sandy sediments. From the mouth of the Changhjiang, Zhemin mud belt, the northern Taiwan Strait mud belt, to the central Taiwan Strait, 7Be activity in the seafloor sediment indicates that the freshness of the terrigenous sediments decreased. The Mass Magnetic Susceptiblity (MSI) demonstrates that the terrigenous sediments decreased from north to south. The MSI signals in the core off the mouth of the Minjiang are different from those in the neighboring cores. This is suspected due to the convergence of sediments from the Changjiang and Taiwan. The particle sizes of the cores show that the sediment became coarser from the north to south. In the future the study will make use of 210Pbex, and other environmental and provenance such as water dynamic mechanism variables to explore the sediment source and sink patterns along with the Zhemin-Taiwan Strait mud belts.

High performance of PbSe/PbS core/shell quantum dot heterojunction solar cells: short circuit current enhancement without the loss of open circuit voltage by shell thickness control.

PubMed

Choi, Hyekyoung; Song, Jung Hoon; Jang, Jihoon; Mai, Xuan Dung; Kim, Sungwoo; Jeong, Sohee

2015-11-07

We fabricated heterojunction solar cells with PbSe/PbS core shell quantum dots and studied the precisely controlled PbS shell thickness dependency in terms of optical properties, electronic structure, and solar cell performances. When the PbS shell thickness increases, the short circuit current density (JSC) increases from 6.4 to 11.8 mA cm(-2) and the fill factor (FF) enhances from 30 to 49% while the open circuit voltage (VOC) remains unchanged at 0.46 V even with the decreased effective band gap. We found that the Fermi level and the valence band maximum level remain unchanged in both the PbSe core and PbSe/PbS core/shell with a less than 1 nm thick PbS shell as probed via ultraviolet photoelectron spectroscopy (UPS). The PbS shell reduces their surface trap density as confirmed by relative quantum yield measurements. Consequently, PbS shell formation on the PbSe core mitigates the trade-off relationship between the open circuit voltage and the short circuit current density. Finally, under the optimized conditions, the PbSe core with a 0.9 nm thick shell yielded a power conversion efficiency of 6.5% under AM 1.5.

Experimental Investigation of DC-Bias Related Core Losses in a Boost Inductor (Postprint)

DTIC Science & Technology

2014-08-01

dc bias-flux conditions. These dc bias conditions result in distorted hysteresis loops , increased core losses, and have been shown to be independent...These dc bias conditions result in dis- torted hysteresis loops , increased core losses, and have been shown to be independent of core material. The...controllable converter load currents, this topology is ideal to study dc-related losses. Inductor core hysteresis loop characterization was accomplished

Synthesis of bimetallic Pt-Pd core-shell nanocrystals and their high electrocatalytic activity modulated by Pd shell thickness

NASA Astrophysics Data System (ADS)

Li, Yujing; Wang, Zhi Wei; Chiu, Chin-Yi; Ruan, Lingyan; Yang, Wenbing; Yang, Yang; Palmer, Richard E.; Huang, Yu

2012-01-01

Bimetallic Pt-Pd core-shell nanocrystals (NCs) are synthesized through a two-step process with controlled Pd thickness from sub-monolayer to multiple atomic layers. The oxygen reduction reaction (ORR) catalytic activity and methanol oxidation reactivity of the core-shell NCs for fuel cell applications in alkaline solution are systematically studied and compared based on different Pd thickness. It is found that the Pd shell helps to reduce the over-potential of ORR by up to 50mV when compared to commercial Pd black, while generating up to 3-fold higher kinetic current density. The carbon monoxide poisoning test shows that the bimetallic NCs are more resistant to the CO poisoning than Pt NCs and Pt black. It is also demonstrated that the bimetallic Pt-Pd core-shell NCs can enhance the current density of the methanol oxidation reaction, lowering the over-potential by 35 mV with respect to the Pt core NCs. Further investigation reveals that the Pd/Pt ratio of 1/3, which corresponds to nearly monolayer Pd deposition on Pt core NCs, gives the highest oxidation current density and lowest over-potential. This study shows for the first time the systematic investigation of effects of Pd atomic shells on Pt-Pd bimetallic nanocatalysts, providing valuable guidelines for designing high-performance catalysts for fuel cell applications.Bimetallic Pt-Pd core-shell nanocrystals (NCs) are synthesized through a two-step process with controlled Pd thickness from sub-monolayer to multiple atomic layers. The oxygen reduction reaction (ORR) catalytic activity and methanol oxidation reactivity of the core-shell NCs for fuel cell applications in alkaline solution are systematically studied and compared based on different Pd thickness. It is found that the Pd shell helps to reduce the over-potential of ORR by up to 50mV when compared to commercial Pd black, while generating up to 3-fold higher kinetic current density. The carbon monoxide poisoning test shows that the bimetallic NCs are more resistant to the CO poisoning than Pt NCs and Pt black. It is also demonstrated that the bimetallic Pt-Pd core-shell NCs can enhance the current density of the methanol oxidation reaction, lowering the over-potential by 35 mV with respect to the Pt core NCs. Further investigation reveals that the Pd/Pt ratio of 1/3, which corresponds to nearly monolayer Pd deposition on Pt core NCs, gives the highest oxidation current density and lowest over-potential. This study shows for the first time the systematic investigation of effects of Pd atomic shells on Pt-Pd bimetallic nanocatalysts, providing valuable guidelines for designing high-performance catalysts for fuel cell applications. Electronic supplementary information (ESI) available: Supplementary TEM, EELS, EDS, Electro-chemical measurement data can be found. See DOI: 10.1039/c1nr11374g

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

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

NASA Astrophysics Data System (ADS)

Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

2017-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

Assessment of Current Inservice Inspection and Leak Monitoring Practices for Detecting Materials Degradation in Light Water Reactors

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

Anderson, Michael T.; Simonen, Fredric A.; Muscara, Joseph

2016-09-01

An assessment was performed to determine the effectiveness of existing inservice inspection (ISI) and leak monitoring techniques, and recommend improvements, as necessary, to the programs as currently performed for light water reactor (LWR) components. Information from nuclear power plant (NPP) aging studies and from the U. S. Nuclear Regulatory Commission’s Generic Aging Lessons Learned (GALL) report (NUREG-1801) was used to identify components that have already experienced, or are expected to experience, degradation. This report provides a discussion of the key aspects and parameters that constitute an effective ISI program and a discussion of the basis and background against which themore » effectiveness of the ISI and leak monitoring programs for timely detection of degradation was evaluated. Tables based on the GALL components were used to systematically guide the process, and table columns were included that contained the ISI requirements and effectiveness assessment. The information in the tables was analyzed using histograms to reduce the data and help identify any trends. The analysis shows that the overall effectiveness of the ISI programs is very similar for both boiling water reactors (BWRs) and pressurized water reactors (PWRs). The evaluations conducted as part of this research showed that many ISI programs are not effective at detecting degradation before its extent reached 75% of the component wall thickness. This work should be considered as an assessment of NDE practices at this time; however, industry and regulatory activities are currently underway that will impact future effectiveness assessments. A number of actions have been identified to improve the current ISI programs so that degradation can be more reliably detected.« 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

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.

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

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

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.

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

A novel concept of fault current limiter based on saturable core in high voltage DC transmission system

NASA Astrophysics Data System (ADS)

Yuan, Jiaxin; Zhou, Hang; Gan, Pengcheng; Zhong, Yongheng; Gao, Yanhui; Muramatsu, Kazuhiro; Du, Zhiye; Chen, Baichao

2018-05-01

To develop mechanical circuit breaker in high voltage direct current (HVDC) system, a fault current limiter is required. Traditional method to limit DC fault current is to use superconducting technology or power electronic devices, which is quite difficult to be brought to practical use under high voltage circumstances. In this paper, a novel concept of high voltage DC transmission system fault current limiter (DCSFCL) based on saturable core was proposed. In the DCSFCL, the permanent magnets (PM) are added on both up and down side of the core to generate reverse magnetic flux that offset the magnetic flux generated by DC current and make the DC winding present a variable inductance to the DC system. In normal state, DCSFCL works as a smoothing reactor and its inductance is within the scope of the design requirements. When a fault occurs, the inductance of DCSFCL rises immediately and limits the steepness of the fault current. Magnetic field simulations were carried out, showing that compared with conventional smoothing reactor, DCSFCL can decrease the high steepness of DC fault current by 17% in less than 10ms, which verifies the feasibility and effectiveness of this method.

Pharmacotherapy for the Core Symptoms in Autistic Disorder: Current Status of the Research

PubMed Central

Farmer, Cristan; Thurm, Audrey; Grant, Paul

2013-01-01

The current review covers extant literature on pharmacotherapy for core symptoms of autism. The core symptoms of autism include impairments in social interaction and communication, as well as the presence of restricted and repetitive behaviors. There are no known efficacious treatments for the core social symptoms, although effects on repetitive behaviors are indicated with some data. While studies of fenfluramine, secretin, opiates, and mood stabilizers generally find no effect, mixed results suggest more research is needed on antidepressants and atypical antipsychotics. Newer lines of research, including cholinergic and glutamatergic agents and oxytocin, will be of considerable interest in the future. However, research on the treatment of core symptoms is plagued by limitations in study design, statistical power and other issues inherent to the study of treatments for autism (e.g., heterogeneity of the disorder) that continue to prevent the elucidation of efficacious treatments. PMID:23504356

Power flow control using distributed saturable reactors

DOEpatents

Dimitrovski, Aleksandar D.

2016-02-13

A magnetic amplifier includes a saturable core having a plurality of legs. Control windings wound around separate legs are spaced apart from each other and connected in series in an anti-symmetric relation. The control windings are configured in such a way that a biasing magnetic flux arising from a control current flowing through one of the plurality of control windings is substantially equal to the biasing magnetic flux flowing into a second of the plurality of control windings. The flow of the control current through each of the plurality of control windings changes the reactance of the saturable core reactor by driving those portions of the saturable core that convey the biasing magnetic flux in the saturable core into saturation. The phasing of the control winding limits a voltage induced in the plurality of control windings caused by a magnetic flux passing around a portion of the saturable core.

Pharmacotherapy for the core symptoms in autistic disorder: current status of the research.

PubMed

Farmer, Cristan; Thurm, Audrey; Grant, Paul

2013-03-01

The current review covers extant literature on pharmacotherapy for core symptoms of autism. The core symptoms of autism include impairments in social interaction and communication, as well as the presence of restricted and repetitive behaviors. There are no known efficacious treatments for the core social symptoms, although effects on repetitive behaviors are indicated with some data. While studies of fenfluramine, secretin, opiates, and mood stabilizers generally find no effect, mixed results suggest more research is needed on antidepressants and atypical antipsychotics. Newer lines of research, including cholinergic and glutamatergic agents and oxytocin, will be of considerable interest in the future. However, research on the treatment of core symptoms is plagued by limitations in study design, statistical power, and other issues inherent to the study of treatments for autism (e.g., heterogeneity of the disorder) that continue to prevent the elucidation of efficacious treatments.

Core-shell homojunction silicon vertical nanowire tunneling field-effect transistors.

PubMed

Yoon, Jun-Sik; Kim, Kihyun; Baek, Chang-Ki

2017-01-23

We propose three-terminal core-shell (CS) silicon vertical nanowire tunneling field-effect transistors (TFETs), which can be fabricated by conventional CMOS technology. CS TFETs show lower subthreshold swing (SS) and higher on-state current than conventional TFETs through their high surface-to-volume ratio, which increases carrier-tunneling region with no additional device area. The on-state current can be enhanced by increasing the nanowire height, decreasing equivalent oxide thickness (EOT) or creating a nanowire array. The off-state current is also manageable for power saving through selective epitaxial growth at the top-side nanowire region. CS TFETs with an EOT of 0.8 nm and an aspect ratio of 20 for the core nanowire region provide the largest drain current ranges with point SS values below 60 mV/dec and superior on/off current ratio under all operation voltages of 0.5, 0.7, and 1.0 V. These devices are promising for low-power applications at low fabrication cost and high device density.

Tokamak reactor for treating fertile material or waste nuclear by-products

DOEpatents

Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

2012-10-02

Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

Core outcome sets in women's and newborn health: a systematic review.

PubMed

Duffy, Jmn; Rolph, R; Gale, C; Hirsch, M; Khan, K S; Ziebland, S; McManus, R J

2017-09-01

Variation in outcome collection and reporting is a serious hindrance to progress in our specialty; therefore, over 80 journals have come together to support the development, dissemination, and implementation of core outcome sets. This study systematically reviewed and characterised registered, progressing, or completed core outcome sets relevant to women's and newborn health. Systematic search using the Core Outcome Measures in Effectiveness Trial initiative and the Core Outcomes in Women's and Newborn Health initiative databases. Registry entries, protocols, systematic reviews, and core outcome sets. Descriptive statistics to describe characteristics and results. There were 49 core outcome sets registered in maternal and newborn health, with the majority registered in 2015 (n = 22; 48%) or 2016 (n = 16; 32%). Benign gynaecology (n = 8; 16%) and newborn health (n = 3; 6%) are currently under-represented. Twenty-four (52%) core outcome sets were funded by international (n = 1; <1%), national (n = 18; 38%), and regional (n = 4; 8%) bodies. Seven protocols were published. Twenty systematic reviews have characterised the inconsistency in outcome reporting across a broad range of relevant healthcare conditions. Four core outcome sets were completed: reconstructive breast surgery (11 outcomes), preterm birth (13 outcomes), epilepsy in pregnancy (29 outcomes), and maternity care (48 outcomes). The quantitative, qualitative, and consensus methods used to develop core outcome sets varied considerably. Core outcome sets are currently being developed across women's and newborn health, although coverage of topics is variable. Development of further infrastructure to develop, disseminate, and implement core outcome sets is urgently required. Forty-nine women's and newborn core outcome sets registered. 50% funded. 7 protocols, 20 systematic reviews, and 4 core outcome sets published. @coreoutcomes @jamesmnduffy. © 2017 Royal College of Obstetricians and Gynaecologists.

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

Common Core in the Real World

ERIC Educational Resources Information Center

Hess, Frederick M.; McShane, Michael Q.

2013-01-01

There are at least four key places where the Common Core intersects with current efforts to improve education in the United States--testing, professional development, expectations, and accountability. Understanding them can help educators, parents, and policymakers maximize the chance that the Common Core is helpful to these efforts and, perhaps…

Inferences from the dynamical history of Mercury's rotation

NASA Technical Reports Server (NTRS)

Peale, S. J.

1976-01-01

The history of Mercury's spin angular momentum is reviewed. It is shown that the current nonsynchronous but resonant spin and the nearly zero obliquity place almost no restrictions on the primordial spin state. The only exception comes about from a liquid core-solid mantle interaction which excludes a slow primordial spin concurrent with a large obliquity. The current occupancy of a final evolutionary spin state leads to the description of a scheme by which we can determine the extent of a currently liquid Mercurian core.

The Effects of Visual Art Integration on Reading at the Elementary Level. A Review of Literature

ERIC Educational Resources Information Center

McCarty, Kristine A.

2007-01-01

Although visual art is considered a subject deemed by the federal government as part of the core curriculum, many elementary schools do not include this subject into the current core curriculum of studies. This review of literature provides insight through current qualitative and quantitative studies on the effectiveness of including visual art…

Models of the Earth's Core.

PubMed

Stevenson, D J

1981-11-06

Combined inferences from seismology, high-pressure experiment and theory, geomagnetism, fluid dynamics, and current views of terrestrial planetary evolution lead to models of the earth's core with the following properties. Core formation was contemporaneous with earth accretion; the core is not in chemical equilibrium with the mantle; the outer core is a fluid iron alloy containing significant quantities of lighter elements and is probably almost adiabatic and compositionally uniform; the more iron-rich inner solid core is a consequence of partial freezing of the outer core, and the energy release from this process sustains the earth's magnetic field; and the thermodynamic properties of the core are well constrained by the application of liquid-state theory to seismic and laboratory data.

Magnetic characterization of the stator core of a high-speed motor made of an ultrathin electrical steel sheet using the magnetic property evaluation system

NASA Astrophysics Data System (ADS)

Oka, Mohachiro; Enokizono, Masato; Mori, Yuji; Yamazaki, Kazumasa

2018-04-01

Recently, the application areas for electric motors have been expanding. For instance, electric motors are used in new technologies such as rovers, drones, cars, and robots. The motor used in such machinery should be small, high-powered, highly-efficient, and high-speed. In such motors, loss at high-speed rotation must be especially minimal. Eddy-current loss in the stator core is known to increase greatly during loss at high-speed rotation of the motor. To produce an efficient high-speed motor, we are developing a stator core for a motor using an ultrathin electrical steel sheet with only a small amount of eddy-current loss. Furthermore, the magnetic property evaluation for efficient, high-speed motor stator cores that use conventional commercial frequency is insufficient. Thus, we made a new high-speed magnetic property evaluation system to evaluate the magnetic properties of the efficient high-speed motor stator core. This system was composed of high-speed A/D converters, D/A converters, and a high-speed power amplifier. In experiments, the ultrathin electrical steel sheet dramatically suppressed iron loss and, in particular, eddy-current loss. In addition, a new high-speed magnetic property evaluation system accurately evaluated the magnetic properties of the efficient high-speed motor stator core.

Evidence for Updating the Core Domain Set of Outcome Measures for Juvenile Idiopathic Arthritis: Report from a Special Interest Group at OMERACT 2016.

PubMed

Morgan, Esi M; Riebschleger, Meredith P; Horonjeff, Jennifer; Consolaro, Alessandro; Munro, Jane E; Thornhill, Susan; Beukelman, Timothy; Brunner, Hermine I; Creek, Emily L; Harris, Julia G; Horton, Daniel B; Lovell, Daniel J; Mannion, Melissa L; Olson, Judyann C; Rahimi, Homaira; Gallo, Maria Chiara; Calandra, Serena; Ravelli, Angelo; Ringold, Sarah; Shenoi, Susan; Stinson, Jennifer; Toupin-April, Karine; Strand, Vibeke; Bingham, Clifton O

2017-12-01

The current Juvenile Idiopathic Arthritis (JIA) Core Set was developed in 1997 to identify the outcome measures to be used in JIA clinical trials using statistical and consensus-based techniques, but without patient involvement. The importance of patient/parent input into the research process has increasingly been recognized over the years. An Outcome Measures in Rheumatology (OMERACT) JIA Core Set Working Group was formed to determine whether the outcome domains of the current core set are relevant to those involved or whether the core set domains should be revised. Twenty-four people from the United States, Canada, Australia, and Europe, including patient partners, formed the working group. Guided by the OMERACT Filter 2.0 process, we performed (1) a systematic literature review of outcome domains, (2) a Web-based survey (142 patients, 343 parents), (3) an idea-generation study (120 parents), (4) 4 online discussion boards (24 patients, 20 parents), and (5) a Special Interest Group (SIG) activity at the OMERACT 13 (2016) meeting. A MEDLINE search of outcome domains used in studies of JIA yielded 5956 citations, of which 729 citations underwent full-text review, and identified additional domains to those included in the current JIA Core Set. Qualitative studies on the effect of JIA identified multiple additional domains, including pain and participation. Twenty-one participants in the SIG achieved consensus on the need to revise the entire JIA Core Set. The results of qualitative studies and literature review support the need to expand the JIA Core Set, considering, among other things, additional patient/parent-centered outcomes, clinical data, and imaging data.

Eddy-Current Monitoring Of Composite Layups

NASA Technical Reports Server (NTRS)

Fox, Robert L.; Buckley, John D.

1993-01-01

Eddy-current-probe apparatus used to determine predominant orientations of fibers in fiber/matrix composite materials. Apparatus nondestructive, noninvasive means for monitoring composite prepregs and layups during fabrication to ensure predictable and repeatable mechanical properties of finished composite panels. Consists essentially of electromagnet coil wrapped around horseshoe-shaped powdered-iron or ferrite ore. Optionally, capacitor included in series or parallel with coil to form resonant circuit. Impedance monitor excites radio-frequency current in coil and measures impedance of probe circuit. Affected by whatever material placed near ends of core, where material intercepts alternating magnetic field excited in core by current in coil.

The Thermal Conductivity of Earth's Core: A Key Geophysical Parameter's Constraints and Uncertainties

NASA Astrophysics Data System (ADS)

Williams, Q.

2018-05-01

The thermal conductivity of iron alloys at high pressures and temperatures is a critical parameter in governing ( a) the present-day heat flow out of Earth's core, ( b) the inferred age of Earth's inner core, and ( c) the thermal evolution of Earth's core and lowermost mantle. It is, however, one of the least well-constrained important geophysical parameters, with current estimates for end-member iron under core-mantle boundary conditions varying by about a factor of 6. Here, the current state of calculations, measurements, and inferences that constrain thermal conductivity at core conditions are reviewed. The applicability of the Wiedemann-Franz law, commonly used to convert electrical resistivity data to thermal conductivity data, is probed: Here, whether the constant of proportionality, the Lorenz number, is constant at extreme conditions is of vital importance. Electron-electron inelastic scattering and increases in Fermi-liquid-like behavior may cause uncertainties in thermal conductivities derived from both first-principles-associated calculations and electrical conductivity measurements. Additional uncertainties include the role of alloying constituents and local magnetic moments of iron in modulating the thermal conductivity. Thus, uncertainties in thermal conductivity remain pervasive, and hence a broad range of core heat flows and inner core ages appear to remain plausible.

The International Classification of Functioning (ICF) core set for breast cancer from the perspective of women with the condition.

PubMed

Cooney, Marese; Galvin, Rose; Connolly, Elizabeth; Stokes, Emma

2013-05-01

The ICF Core Set for breast cancer was generated by international experts for women who have had surgery and radiation but it has not yet been validated. The objective of the study was to validate the ICF Core Set from the perspective of women with breast cancer. A qualitative focus group methodology was used. The sessions were transcribed verbatim. Meaning units were identified by two independent researchers. The agreed list was subsequently linked to ICF categories by two independent researchers according to pre-defined linking rules. Data saturation determined the number of focus groups conducted. Quality of the data analyses was assured by multiple coding and peer review. Thirty-four women participated in seven focus groups. A total of 1621 meaning units were identified which were linked to 74 of the existing 80 Core Set categories. Additional ICF categories not currently included in the Core Set were identified by the women. The validity of the Core Set was largely supported. However, some categories currently not covered by the ICF Core Set for Breast Cancer will need to be considered for inclusion if the Core Set is to reflect all women who have had treatment for breast cancer

Progress on core outcome sets for critical care research.

PubMed

Blackwood, Bronagh; Marshall, John; Rose, Louise

2015-10-01

Appropriate selection and definition of outcome measures are essential for clinical trials to be maximally informative. Core outcome sets (an agreed, standardized collection of outcomes measured and reported in all trials for a specific clinical area) were developed due to established inconsistencies in trial outcome selection. This review discusses the rationale for, and methods of, core outcome set development, as well as current initiatives in critical care. Recent systematic reviews of reported outcomes and measurement instruments relevant to the critically ill highlight inconsistencies in outcome selection, definition, and measurement, thus establishing the need for core outcome sets. Current critical care initiatives include development of core outcome sets for trials aimed at reducing mechanical ventilation duration; rehabilitation following critical illness; long-term outcomes in acute respiratory failure; and epidemic and pandemic studies of severe acute respiratory infection. Development and utilization of core outcome sets for studies relevant to the critically ill is in its infancy compared to other specialties. Notwithstanding, core outcome set development frameworks and guidelines are available, several sets are in various stages of development, and there is strong support from international investigator-led collaborations including the International Forum for Acute Care Trialists.

Current advances in precious metal core-shell catalyst design.

PubMed

Wang, Xiaohong; He, Beibei; Hu, Zhiyu; Zeng, Zhigang; Han, Sheng

2014-08-01

Precious metal nanoparticles are commonly used as the main active components of various catalysts. Given their high cost, limited quantity, and easy loss of catalytic activity under severe conditions, precious metals should be used in catalysts at low volumes and be protected from damaging environments. Accordingly, reducing the amount of precious metals without compromising their catalytic performance is difficult, particularly under challenging conditions. As multifunctional materials, core-shell nanoparticles are highly important owing to their wide range of applications in chemistry, physics, biology, and environmental areas. Compared with their single-component counterparts and other composites, core-shell nanoparticles offer a new active interface and a potential synergistic effect between the core and shell, making these materials highly attractive in catalytic application. On one hand, when a precious metal is used as the shell material, the catalytic activity can be greatly improved because of the increased surface area and the closed interfacial interaction between the core and the shell. On the other hand, when a precious metal is applied as the core material, the catalytic stability can be remarkably improved because of the protection conferred by the shell material. Therefore, a reasonable design of the core-shell catalyst for target applications must be developed. We summarize the latest advances in the fabrications, properties, and applications of core-shell nanoparticles in this paper. The current research trends of these core-shell catalysts are also highlighted.

Ultrafast Photodetection in the Quantum Wells of Single AlGaAs/GaAs-Based Nanowires.

PubMed

Erhard, N; Zenger, S; Morkötter, S; Rudolph, D; Weiss, M; Krenner, H J; Karl, H; Abstreiter, G; Finley, J J; Koblmüller, G; Holleitner, A W

2015-10-14

We investigate the ultrafast optoelectronic properties of single Al0.3Ga0.7As/GaAs core-shell nanowires. The nanowires contain GaAs-based quantum wells. For a resonant excitation of the quantum wells, we find a picosecond photocurrent which is consistent with an ultrafast lateral expansion of the photogenerated charge carriers. This Dember-effect does not occur for an excitation of the GaAs-based core of the nanowires. Instead, the core exhibits an ultrafast displacement current and a photothermoelectric current at the metal Schottky contacts. Our results uncover the optoelectronic dynamics in semiconductor core-shell nanowires comprising quantum wells, and they demonstrate the possibility to use the low-dimensional quantum well states therein for ultrafast photoswitches and photodetectors.

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.

Adaptive control method for core power control in TRIGA Mark II reactor

NASA Astrophysics Data System (ADS)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Training in Vocational Assessment: Preparing Rehabilitation Counselors and Meeting the Requirements of the CORE Standards

ERIC Educational Resources Information Center

Tansey, Timothy N.

2008-01-01

Assessment represents a foundational component of rehabilitation counseling services. The revised Council on Rehabilitation Education (CORE) standards implemented in 2004 resulted in the redesign of the knowledge and outcomes under the Assessment standard. The author reviews the current CORE standard for training in assessment within the context…

20 CFR 641.700 - What performance measures/indicators apply to SCSEP grantees?

Code of Federal Regulations, 2010 CFR

2010-04-01

... performance. There are currently eight performance measures, of which six are core indicators and two are additional indicators. Core indicators (defined in § 641.710) are subject to goal-setting and corrective action (described in § 641.720); that is, performance level goals for each core indicator must be agreed...

The Trouble with Triplets in Biodiversity Informatics: A Data-Driven Case against Current Identifier Practices

PubMed Central

Guralnick, Robert; Conlin, Tom; Deck, John; Stucky, Brian J.; Cellinese, Nico

2014-01-01

The biodiversity informatics community has discussed aspirations and approaches for assigning globally unique identifiers (GUIDs) to biocollections for nearly a decade. During that time, and despite misgivings, the de facto standard identifier has become the “Darwin Core Triplet”, which is a concatenation of values for institution code, collection code, and catalog number associated with biocollections material. Our aim is not to rehash the challenging discussions regarding which GUID system in theory best supports the biodiversity informatics use case of discovering and linking digital data across the Internet, but how well we can link those data together at this moment, utilizing the current identifier schemes that have already been deployed. We gathered Darwin Core Triplets from a subset of VertNet records, along with vertebrate records from GenBank and the Barcode of Life Data System, in order to determine how Darwin Core Triplets are deployed “in the wild”. We asked if those triplets follow the recommended structure and whether they provide an easy and unambiguous means to track from specimen records to genetic sequence records. We show that Darwin Core Triplets are often riddled with semantic and syntactic errors when deployed and curated in practice, despite specifications about how to construct them. Our results strongly suggest that Darwin Core Triplets that have not been carefully curated are not currently serving a useful role for relinking data. We briefly consider needed next steps to overcome current limitations. PMID:25470125

Models of the earth's core

NASA Technical Reports Server (NTRS)

Stevenson, D. J.

1981-01-01

Combined inferences from seismology, high-pressure experiment and theory, geomagnetism, fluid dynamics, and current views of terrestrial planetary evolution lead to models of the earth's core with five basic properties. These are that core formation was contemporaneous with earth accretion; the core is not in chemical equilibrium with the mantle; the outer core is a fluid iron alloy containing significant quantities of lighter elements and is probably almost adiabatic and compositionally uniform; the more iron-rich inner solid core is a consequence of partial freezing of the outer core, and the energy release from this process sustains the earth's magnetic field; and the thermodynamic properties of the core are well constrained by the application of liquid-state theory to seismic and labroatory data.

Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

NASA Astrophysics Data System (ADS)

Huddar, Lakshana Ravindranath

With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by underlining key distortions between the experimental and the prototypical conditions. This dissertation is broadly split into four parts. Firstly, the heat transfer phenomenology in the PB-FHR core was outlined. Although the viscous dissipation term and the thermal diffusion term (including thermal dispersion) were similar in magnitude, they were overshadowed by the advection term which was about 104 times bigger during normal operation and 105 times bigger during accident transients in which natural circulation becomes the main mode of fluid flow. Thus it is safe to neglect the viscous dissipation and the thermal diffusion terms in the PB-FHR core without a significant loss of accuracy. Secondly, separate effects tests (SET) were performed using simulant oils, and the results were compared to the prototypical conditions using flinak as the fluoride salt. The main purpose of these experiments was to study natural convection heat transfer and identify any distortions between the two cases. An isolated copper sphere was immersed in flinak and a parallel experiment was performed using simulant oil. A large discrepancy between the flinak and the oil was noted, due to distortions from assuming quasi-steady state conditions. A steady state experiment using a cylindrical heater immersed in oil was also performed, and the results compared to a similar experiment done at Oak Ridge National Laboratory (ORNL) using flinak. The Nusselt numbers matched within 10% for laminar flows. This supports the conclusion that natural convection similitude does exist for oils used in scaled experiments, allowing natural convection data to be used for for FHR and MSR modeling. This is important, due to the lack of significant experimental data showing natural convection in fluoride salts, so these SETs add to the overall understanding of their heat transfer properties. With the knowledge of the distortions between the oil and the salt, an experiment to measure heat transfer coefficients within a pebble-bed test section was designed, constructed and performed. Oil was pumped through a test section filled with randomly packed copper spheres. The temperature of the oil was pulsed at a constant frequency, which caused a temperature difference between the pebbles and the oil. An excellent match was found between the measured heat transfer coefficients and the literature. This data provides an essential closure parameter for multiphysics modeling of the PB-FHR. Using frequency response techniques in scaled experiments is an innovative approach for extracting dynamic responses to coolant-structure interactions. Finally, an integrated model of the passive decay heat removal system was presented using Flownex and the simulations compared to experimental data. A good match was found with the data, which was within 14%. The work presented in this dissertation shows fundamental details on heat transfer in the PB-FHR core using experimental data and simulations, leading us closer to developing advanced nuclear reactors that can later be commercialized. Advanced nuclear reactors such as the PB-FHR have immense potential in reducing greenhouse gas emissions and combating climate change while being exceedingly safe and providing reliable electricity.

Pattern optimizing verification of self-align quadruple patterning

NASA Astrophysics Data System (ADS)

Yamato, Masatoshi; Yamada, Kazuki; Oyama, Kenichi; Hara, Arisa; Natori, Sakurako; Yamauchi, Shouhei; Koike, Kyohei; Yaegashi, Hidetami

2017-03-01

Lithographic scaling continues to advance by extending the life of 193nm immersion technology, and spacer-type multi-patterning is undeniably the driving force behind this trend. Multi-patterning techniques such as self-aligned double patterning (SADP) and self-aligned quadruple patterning (SAQP) have come to be used in memory devices, and they have also been adopted in logic devices to create constituent patterns in the formation of 1D layout designs. Multi-patterning has consequently become an indispensible technology in the fabrication of all advanced devices. In general, items that must be managed when using multi-patterning include critical dimension uniformity (CDU), line edge roughness (LER), and line width roughness (LWR). Recently, moreover, there has been increasing focus on judging and managing pattern resolution performance from a more detailed perspective and on making a right/wrong judgment from the perspective of edge placement error (EPE). To begin with, pattern resolution performance in spacer-type multi-patterning is affected by the process accuracy of the core (mandrel) pattern. Improving the controllability of CD and LER of the mandrel is most important, and to reduce LER, an appropriate smoothing technique should be carefully selected. In addition, the atomic layer deposition (ALD) technique is generally used to meet the need for high accuracy in forming the spacer film. Advances in scaling are accompanied by stricter requirements in the controllability of fine processing. In this paper, we first describe our efforts in improving controllability by selecting the most appropriate materials for the mandrel pattern and spacer film. Then, based on the materials selected, we present experimental results on a technique for improving etching selectivity.

To enhance the efficiency of a power supply circuit by the use of Fe-P-B-Nb-type ultralow loss glassy metal core

NASA Astrophysics Data System (ADS)

Matsumoto, H.; Urata, A.; Yamada, Y.; Makino, A.

2009-04-01

The inductor in a power supply is required to be capable of dealing satisfactorily with the high-current supply and to improve the power loss characteristic. A novel glassy metal powder with a chemical composition Fe77P7B13Nb3 features both a high saturated magnetic flux density of 1.3 T and a low coercive force of 2.0 A/m, which has a stable amorphous structure suitable for glassy metal composite cores. Hence there is no magnetic saturation even under a high-current supply, and it is confirmed to have significantly low magnetic loss resulting from the low coercive force. As a result of using the glassy metal alloy Fe77P7B13Nb3 powder in an inductor core, we have achieved improvement in power supply efficiency by up to roughly 2.0%. Moreover, the reduction in the standby power requirement by the improvement in the power supply efficiency in the low load current case, where the core loss occupies a high ratio in the entire loss, can be expected. Additionally, heat generation in a core is suppressed by using the low loss powder, and it becomes easy to design a temperature rise in the entire power supply circuit.

Current limiting behavior in three-phase transformer-type SFCLs using an iron core according to variety of fault

NASA Astrophysics Data System (ADS)

Cho, Yong-Sun; Jung, Byung-Ik; Ha, Kyoung-Hun; Choi, Soo-Geun; Park, Hyoung-Min; Choi, Hyo-Sang

To apply the superconducting fault current limiter (SFCL) to the power system, the reliability of the fault-current-limiting operation must be ensured in diverse fault conditions. The SFCL must also be linked to the operation of the high-speed recloser in the power system. In this study, a three-phase transformer-type SFCL, which has a neutral line to improve the simultaneous quench characteristics of superconducting elements, was manufactured to analyze the fault-current-limiting characteristic according to the single, double, and triple line-to-ground faults. The transformer-type SFCL, wherein three-phase windings are connected to one iron core, reduced the burden on the superconducting element as the superconducting element on the sound phase was also quenched in the case of the single line-to-ground fault. In the case of double or triple line-to-ground faults, the flux from the faulted phase winding was interlinked with other faulted or sound phase windings, and the fault-current-limiting rate decreased because the windings of three phases were inductively connected by one iron core.

Electrically operated magnetic switch designed to display reduced leakage inductance

DOEpatents

Cook, Edward G.

1994-01-01

An electrically operated magnetic switch is disclosed herein for use in opening and closing a circuit between two terminals depending upon the voltage across these terminals. The switch so disclosed is comprised of a ferrite core in the shape of a toroid having opposing ends and opposite inner and outer sides and an arrangement of electrically conductive components defining at least one current flow path which makes a number of turns around the core. This arrangement of components includes a first plurality of electrically conducive rigid rods parallel with and located outside the outer side of the core and a second plurality of electrically conductive rigid rods parallel with and located inside the inner side of the core. The arrangement also includes means for electrically connecting these rods together so that the define the current flow path. In one embodiment, this latter means uses rigid cross-tab means. In another, preferred embodiment, printed circuits on rigid dielectric substrates located on opposite ends of the core are utilized to interconnect the rods together.

Superconducting shielded core reactor with reduced AC losses

DOEpatents

Cha, Yung S.; Hull, John R.

2006-04-04

A superconducting shielded core reactor (SSCR) operates as a passive device for limiting excessive AC current in a circuit operating at a high power level under a fault condition such as shorting. The SSCR includes a ferromagnetic core which may be either closed or open (with an air gap) and extends into and through a superconducting tube or superconducting rings arranged in a stacked array. First and second series connected copper coils each disposed about a portion of the iron core are connected to the circuit to be protected and are respectively wound inside and outside of the superconducting tube or rings. A large impedance is inserted into the circuit by the core when the shielding capability of the superconducting arrangement is exceeded by the applied magnetic field generated by the two coils under a fault condition to limit the AC current in the circuit. The proposed SSCR also affords reduced AC loss compared to conventional SSCRs under continuous normal operation.

Radial tunnel diodes based on InP/InGaAs core-shell nanowires

NASA Astrophysics Data System (ADS)

Tizno, Ofogh; Ganjipour, Bahram; Heurlin, Magnus; Thelander, Claes; Borgström, Magnus T.; Samuelson, Lars

2017-03-01

We report on the fabrication and characterization of radial tunnel diodes based on InP(n+)/InGaAs(p+) core-shell nanowires, where the effect of Zn-dopant precursor flow on the electrical properties of the devices is evaluated. Selective and local etching of the InGaAs shell is employed to access the nanowire core in the contact process. Devices with an n+-p doping profile show normal diode rectification, whereas n+-p+ junctions exhibit typical tunnel diode characteristics with peak-to-valley current ratios up to 14 at room temperature and 100 at 4.2 K. A maximum peak current density of 28 A/cm2 and a reverse current density of 7.3 kA/cm2 at VSD = -0.5 V are extracted at room temperature after normalization with the effective junction area.

Analytical Estimation of the Scale of Earth-Like Planetary Magnetic Fields

NASA Astrophysics Data System (ADS)

Bologna, Mauro; Tellini, Bernardo

2014-10-01

In this paper we analytically estimate the magnetic field scale of planets with physical core conditions similar to that of Earth from a statistical physics point of view. We evaluate the magnetic field on the basis of the physical parameters of the center of the planet, such as density, temperature, and core size. We look at the contribution of the Seebeck effect on the magnetic field, showing that a thermally induced electrical current can exist in a rotating fluid sphere. We apply our calculations to Earth, where the currents would be driven by the temperature difference at the outer-inner core boundary, Jupiter and the Jupiter's satellite Ganymede. In each case we show that the thermal generation of currents leads to a magnetic field scale comparable to the observed fields of the considered celestial bodies.

Air core poloidal magnetic field system for a toroidal plasma producing device

DOEpatents

Marcus, Frederick B.

1978-01-01

A poloidal magnetics system for a plasma producing device of toroidal configuration is provided that reduces both the total volt-seconds requirement and the magnitude of the field change at the toroidal field coils. The system utilizes an air core transformer wound between the toroidal field (TF) coils and the major axis outside the TF coils. Electric current in the primary windings of this transformer is distributed and the magnetic flux returned by air core windings wrapped outside the toroidal field coils. A shield winding that is closely coupled to the plasma carries a current equal and opposite to the plasma current. This winding provides the shielding function and in addition serves in a fashion similar to a driven conducting shell to provide the equilibrium vertical field for the plasma. The shield winding is in series with a power supply and a decoupling coil located outside the TF coil at the primary winding locations. The present invention requires much less energy than the usual air core transformer and is capable of substantially shielding the toroidal field coils from poloidal field flux.

Exploring Marriage and Family Therapy Supervisees' Perspectives about Postgraduate Supervision and the Acquisition of Core Competencies

ERIC Educational Resources Information Center

Steele, Stephanie J.

2013-01-01

The topic of core competencies has been a central focus in the marriage and family therapy field since 2003. There are currently no published studies from the supervisees' perspective about the role of supervision in the acquisition of core competencies. This qualitative study used transcendental phenomenology to explore supervisees' perspectives…

Core Self-Evaluations, Worry, Life Satisfaction, and Psychological Well-Being: An Investigation in the Asian Context

ERIC Educational Resources Information Center

Rathi, Neerpal; Lee, Kidong

2018-01-01

The concept of core self-evaluations has been extensively investigated in Western and European countries, nonetheless its implications in Asian countries remains relatively unexplored. To void this gap, the current study investigated the association of core self-evaluations with worry, life satisfaction, and psychological well-being among South…

Design of air-gapped magnetic-core inductors for superimposed direct and alternating currents

NASA Technical Reports Server (NTRS)

Ohri, A. K.; Wilson, T. G.; Owen, H. A., Jr.

1976-01-01

Using data on standard magnetic-material properties and standard core sizes for air-gap-type cores, an algorithm designed for a computer solution is developed which optimally determines the air-gap length and locates the quiescent point on the normal magnetization curve so as to yield an inductor design with the minimum number of turns for a given ac voltage and frequency and with a given dc bias current superimposed in the same winding. Magnetic-material data used in the design are the normal magnetization curve and a family of incremental permeability curves. A second procedure, which requires a simpler set of calculations, starts from an assigned quiescent point on the normal magnetization curve and first screens candidate core sizes for suitability, then determines the required turns and air-gap length.

High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

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

McDowell, Bruce K.; Nickolaus, James R.; Mitchell, Mark R.

2011-10-31

Current interest expressed by industry in HTGR plants, particularly modular plants with power up to about 600 MW(e) per unit, has prompted NRC to task PNNL with assessing the currently available literature related to codes and standards applicable to HTGR plants, the operating history of past and present HTGR plants, and with evaluating the proposed designs of RPV and associated piping for future plants. Considering these topics in the order they are arranged in the text, first the operational histories of five shut-down and two currently operating HTGR plants are reviewed, leading the authors to conclude that while small, simplemore » prototype HTGR plants operated reliably, some of the larger plants, particularly Fort St. Vrain, had poor availability. Safety and radiological performance of these plants has been considerably better than LWR plants. Petroleum processing plants provide some applicable experience with materials similar to those proposed for HTGR piping and vessels. At least one currently operating plant - HTR-10 - has performed and documented a leak before break analysis that appears to be applicable to proposed future US HTGR designs. Current codes and standards cover some HTGR materials, but not all materials are covered to the high temperatures envisioned for HTGR use. Codes and standards, particularly ASME Codes, are under development for proposed future US HTGR designs. A 'roadmap' document has been prepared for ASME Code development; a new subsection to section III of the ASME Code, ASME BPVC III-5, is scheduled to be published in October 2011. The question of terminology for the cross-duct structure between the RPV and power conversion vessel is discussed, considering the differences in regulatory requirements that apply depending on whether this structure is designated as a 'vessel' or as a 'pipe'. We conclude that designing this component as a 'pipe' is the more appropriate choice, but that the ASME BPVC allows the owner of the facility to select the preferred designation, and that either designation can be acceptable.« less

Performance Comparison of Finemet and Metglas Tape Cores Under Non-Sinusoidal Waveforms with DC Bias (POSTPRINT)

DTIC Science & Technology

2017-06-01

dc converter-based test system was built to intentionally introduce inductor current harmonics by varying the filter capacitance and parasitic...the inclusion of distorted waveforms obtained by varying filter capacitance. At higher frequencies, the Metglas cores were found to exhibit greater...was built to intentionally introduce inductor current harmonics by varying the filter capacitance and parasitic inductance of the test system. Both

Multicore Architectures for Multiple Independent Levels of Security Applications

DTIC Science & Technology

2012-09-01

to bolster the MILS effort. However, current MILS operating systems are not designed for multi-core platforms. They do not have the hardware support...current MILS operating systems are not designed for multi‐core platforms. They do not have the hardware support to ensure that the separation...the availability of information at different security classification levels while increasing the overall security of the computing system . Due to the

Reconciling Ecological Educational Planning with Access to the Common Core: Putting the Cart before the Horse?--A Response to Hunt and McDonnell

ERIC Educational Resources Information Center

Ayres, Kevin Michael

2012-01-01

Hunt, McDonnell, and Crocket (2012) highlight the current curriculum debate occurring in the area of severe disabilities and suggest that that a middle ground exists between these competing views: one emphasizing the general curriculum (e.g., Common Core) for all students and the other one stressing an ecological approach focused on current and…

Transcatheter Aortic Valve Implantation: Experience with the CoreValve Device.

PubMed

Asgar, Anita W; Bonan, Raoul

2012-01-01

The field of transcatheter aortic valve implantation has been rapidly evolving. The Medtronic CoreValve first emerged on the landscape in 2004 with initial first human studies, and it is currently being studied in the Pivotal US trial. This article details the current experience with the self-expanding aortic valve with a focus on clinical results and ongoing challenges. Copyright © 2012 Elsevier Inc. All rights reserved.

Incidence of tissue coring with the 25-gauge Quincke and Whitacre spinal needles.

PubMed

Campbell, D C; Douglas, M J; Taylor, G

1996-01-01

Tissue cores, implanted into the subarachnoid space during subarachnoid injections, can develop into intraspinal lumbar epidermoid tumors. The availability of smaller needles has made spinal anesthesia more popular. Therefore, this prospective, randomized, blinded study was undertaken to determine whether tissue coring occurs with two of the currently used 25-gauge spinal needles. Fifteen 25-gauge Quincke and seventeen 25-gauge Whitacre spinal needles, in which cerebrospinal fluid (CSF) was not identified and the local anesthetic solution not injected, were obtained from adult male patients undergoing spinal anesthesia. The needles were then evaluated by a pathologist following randomization with similar sterile, unused spinal needles. Twenty additional needles, ten of each type, in which CSF was identified and through which local anesthetic was injected, were also randomized with similar sterile, unused spinal needles and examined. Tissue cores were identified in 12 of the 15 Quincke and 7 of the 17 Whitacre spinal needles in which CSF was not identified (P < .05). Of the 20 needles in which CSF was identified and local anesthetic injected, no tissue cores were identified in the 10 Whitacre needles and only one small tissue core was identified in the 10 Quincke needles. All the tissue cores were identified as fat tissue. The 25-gauge Quincke and 25-gauge Whitacre spinal needles currently used in anesthesia can produce tissue coring.

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

Neutron radiation embrittlement studies in support of continued operation, and validation by sampling of Magnox reactor steel pressure vessels and components

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

Jones, R.B.; Bolton, C.J.

1997-02-01

Magnox steel reactor pressure vessels differ significantly from US LWR vessels in terms of the type of steel used, as well as their operating environment (dose level, exposure temperature range, and neutron spectra). The large diameter ferritic steel vessels are constructed from C-Mn steel plates and forgings joined together with manual metal and submerged-arc welds which are stress-relieved. All Magnox vessels are now at least thirty years old and their continued operation is being vigorously pursued. Vessel surveillance and other programmes are summarized which support this objective. The current understanding of the roles of matrix irradiation damage, irradiation-enhanced copper impuritymore » precipitation and intergranular embrittlement effects is described in so far as these influence the form of the embrittlement and hardening trend curves for each material. An update is given on the influence of high temperature exposure, and on the role of differing neutron spectra. Finally, the validation offered by the results of an initial vessel sampling exercise is summarized together with the objectives of a more extensive future sampling programme.« less

Station Blackout: A case study in the interaction of mechanistic and probabilistic safety analysis

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

Curtis Smith; Diego Mandelli; Cristian Rabiti

2013-11-01

The ability to better characterize and quantify safety margins is important to improved decision making about nuclear power plant design, operation, and plant life extension. As research and development (R&D) in the light-water reactor (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. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway R&D is to support plant decisions for risk-informed margin management with the aim tomore » improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a “station blackout” wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario.« 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.

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

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

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.

Validation of a coupled core-transport, pedestal-structure, current-profile and equilibrium model

NASA Astrophysics Data System (ADS)

Meneghini, O.

2015-11-01

The first workflow capable of predicting the self-consistent solution to the coupled core-transport, pedestal structure, and equilibrium problems from first-principles and its experimental tests are presented. Validation with DIII-D discharges in high confinement regimes shows that the workflow is capable of robustly predicting the kinetic profiles from on axis to the separatrix and matching the experimental measurements to within their uncertainty, with no prior knowledge of the pedestal height nor of any measurement of the temperature or pressure. Self-consistent coupling has proven to be essential to match the experimental results, and capture the non-linear physics that governs the core and pedestal solutions. In particular, clear stabilization of the pedestal peeling ballooning instabilities by the global Shafranov shift and destabilization by additional edge bootstrap current, and subsequent effect on the core plasma profiles, have been clearly observed and documented. In our model, self-consistency is achieved by iterating between the TGYRO core transport solver (with NEO and TGLF for neoclassical and turbulent flux), and the pedestal structure predicted by the EPED model. A self-consistent equilibrium is calculated by EFIT, while the ONETWO transport package evolves the current profile and calculates the particle and energy sources. The capabilities of such workflow are shown to be critical for the design of future experiments such as ITER and FNSF, which operate in a regime where the equilibrium, the pedestal, and the core transport problems are strongly coupled, and for which none of these quantities can be assumed to be known. Self-consistent core-pedestal predictions for ITER, as well as initial optimizations, will be presented. Supported by the US Department of Energy under DE-FC02-04ER54698, DE-SC0012652.

Paralleling power MOSFETs in their active region: Extended range of passively forced current sharing

NASA Technical Reports Server (NTRS)

Niedra, Janis M.

1989-01-01

A simple passive circuit that improves current balance in parallelled power MOSFETs that are not precisely matched and that are operated in their active region from a common gate drive are exhibited. A nonlinear circuit consisting of diodes and resistors generates the differential gate potential required to correct for unbalance while maintaining low losses over a range of current. Also application of a thin tape wound magnetic core to effect dynamic current balance is reviewed, and a simple theory is presented showing that for operation in the active region the branch currents tend to revert to their normal unbalanced values even if the core is not driven into saturation. Results of several comparative experiments are given.

Large Scale Gas Mixing and Stratification Triggered by a Buoyant Plume With and Without Occurrence of Condensation

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

Paladino, Domenico; Auban, Olivier; Zboray, Robert

The benefits of using codes with 3-D capabilities to address safety issues of LWRs will be applicable to both the current generation of nuclear reactors as well to future ALWRs. The phenomena governing the containment response in case of some postulated severe accident scenarios include gas (air, hydrogen, steam) stratification in the containment, gas distribution between containment compartments, wall condensation, etc. These phenomena are driven by buoyant high momentum injection (jets) and/or low momentum injection (plumes). For instance, mixing in the immediate vicinity of the postulated line break is mainly dominated by very high velocity efflux, while low-momentum flows aremore » responsible for most of the transport processes within the containment. A project named SETH is currently in progress under the auspices of 15 OECD countries, with the aim of creating an experimental database suitable to assess the 3-D code capabilities in analyzing key-physical phenomena relevant for LWR safety analysis. This paper describes some results of two SETH tests, performed in the PANDA facility (located at PSI in Switzerland), focusing on plumes flowing near a containment wall. The plumes are generated by injecting a constant amount of steam in one of two interconnected vessels initially filled with air. In one of the two tests the temperature of the injected steam and the initial containment wall and fluid temperatures allowed for condensation during the test. (authors)« 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

Reviewing Core Kindergarten and First-Grade Reading Programs in Light of No Child Left Behind: An Exploratory Study

ERIC Educational Resources Information Center

Al Otaiba, Stephanie; Kosanovich-Grek, Marcia L.; Torgesen, Joseph K.; Hassler, Laura; Wahl, Michelle

2005-01-01

This article describes the findings of our review process for core reading programs and provides a preliminary rubric emanating from this process for rating core reading programs. To our knowledge, this is the first published review of the current "Reading First" guidelines and includes all five components of scientifically based reading…

Core structure of two-dimensional Fermi gas vortices in the BEC-BCS crossover region

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

Madeira, Lucas; Gandolfi, Stefano; Schmidt, Kevin E.

2017-05-02

We report T = 0 diffusion Monte Carlo results for the ground-state and vortex excitation of unpolarized spin-1/2 fermions in a two-dimensional disk. We investigate how vortex core structure properties behave over the BEC-BCS crossover. We calculate the vortex excitation energy, density pro les, and vortex core properties related to the current. We nd a density suppression at the vortex core on the BCS side of the crossover and a depleted core on the BEC limit. Size-effect dependencies in the disk geometry were carefully studied.

Remnant field detector

DOEpatents

Visser, Age T.

1988-05-03

A method apparatus for qualitatively detecting remnant magnetic fields in matched pairs of magnet cores. Equal magnitude and oppositely oriented magnetic flux is induced in the magnet cores by oppositely wound primary windings and current source. Identically wound secondary windings generate output voltages in response to the induced flux. The output voltages generated should be of equal magnitude and opposite polarity if there is no remnant field in the cores. The output voltages will be unequal which is detected if either core has a remnant field.

Remnant field detector

DOEpatents

Visser, Age T.

1988-01-01

A method apparatus for qualitatively detecting remnant magnetic fields in matched pairs of magnet cores. Equal magnitude and oppositely oriented magnetic flux is induced in the magnet cores by oppositely wound primary windings and current source. Identically wound secondary windings generate output voltages in response to the induced flux. The output voltages generated should be of equal magnitude and opposite polarity if there is no remnant field in the cores. The output voltages will be unequal which is detected if either core has a remnant field.

Neutrino probe comparisons of supernovae as a function of redshift

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

Fryer, Christopher Lee

2009-01-01

We compare aspects of supernova explosions produced in the current epoch against those produced in the first round of star formation. Although the total final mass of stars can change dramatically between these two epochs due to different mass-loss rates from winds, their cores remam very similar. The core structure is more sensitive to the stellar evolution code than it is to the amount of metals. As such, current stellar models produce supernovae from first stars that look very similar to that of stars produced in the current epoch. The neutrino signal, a powerful probe of the inner core, ismore » identical to the few percent level for both star formation epochs. A change in the neutrino signal in the supernova population between these two star formation epochs will only arise if the initial mass function is altered.« less

Kinetic equilibrium reconstruction for the NBI- and ICRH-heated H-mode plasma on EAST tokamak

NASA Astrophysics Data System (ADS)

Zhen, ZHENG; Nong, XIANG; Jiale, CHEN; Siye, DING; Hongfei, DU; Guoqiang, LI; Yifeng, WANG; Haiqing, LIU; Yingying, LI; Bo, LYU; Qing, ZANG

2018-04-01

The equilibrium reconstruction is important to study the tokamak plasma physical processes. To analyze the contribution of fast ions to the equilibrium, the kinetic equilibria at two time-slices in a typical H-mode discharge with different auxiliary heatings are reconstructed by using magnetic diagnostics, kinetic diagnostics and TRANSP code. It is found that the fast-ion pressure might be up to one-third of the plasma pressure and the contribution is mainly in the core plasma due to the neutral beam injection power is primarily deposited in the core region. The fast-ion current contributes mainly in the core region while contributes little to the pedestal current. A steep pressure gradient in the pedestal is observed which gives rise to a strong edge current. It is proved that the fast ion effects cannot be ignored and should be considered in the future study of EAST.

Double-diffusive translation of Earth's inner core

NASA Astrophysics Data System (ADS)

Deguen, R.; Alboussiére, T.; Labrosse, S.

2018-03-01

The hemispherical asymmetry of the inner core has been interpreted as resulting form a high-viscosity mode of inner core convection, consisting in a translation of the inner core. A thermally driven translation, as originally proposed, is unlikely if the currently favoured high values of the thermal conductivity of iron at core conditions are correct. We consider here the possibility that inner core translation results from an unstable compositional gradient, which would develop either because the light elements present in the core become increasingly incompatible as the inner core grows, or because of a possibly positive feedback of the development of the F-layer on inner core convection. Though the magnitude of the destabilising effect of the compositional field is predicted to be similar to or smaller than the stabilising effect of the thermal field, the huge difference between thermal and chemical diffusivities implies that double-diffusive instabilities can still arise even if the net buoyancy increases upward. Using linear stability analysis and numerical simulations, we demonstrate that a translation mode can indeed exist if the compositional field is destabilising, even if the temperature profile is subadiabatic, and irrespectively of the relative magnitudes of the composition and potential temperature gradients. The existence of this double diffusive mode of translation requires that the following conditions are met: (i) the compositional profile within the inner core is destabilising, and remains so for a duration longer than the destabilisation timescale (on the order of 200 My, but strongly dependent on the magnitude of the initial perturbation); and (ii) the inner core viscosity is sufficiently large, the required value being a strongly increasing function of the inner core size (e.g. 1017 Pa.s when the inner core was 200 km in radius, and ≃ 3 × 1021 Pa.s at the current inner core size). If these conditions are met, the predicted inner core translation rate is found to be similar to the inner core growth rate, which is more consistent with inferences from the geomagnetic field morphology and secular variation than the higher translation rate predicted for a thermally driven translation.

A technique to identify core journals for neurosurgery using citation scatter analysis and the Bradford distribution across neurosurgery journals.

PubMed

Madhugiri, Venkatesh S; Ambekar, Sudheer; Strom, Shane F; Nanda, Anil

2013-11-01

The volume of scientific literature doubles approximately every 7 years. The coverage of this literature provided by online compendia is variable and incomplete. It would hence be useful to identify "core" journals in any field and validate whether the h index and impact factor truly identify the core journals in every subject. The core journals in every medical specialty would be those that provide a current and comprehensive coverage of the science in that specialty. Identifying these journals would make it possible for individual physicians to keep abreast of research and clinical progress. The top 10 neurosurgical journals (on the basis of impact factor and h index) were selected. A database of all articles cited in the reference lists of papers published in issues of these journals published in the first quarter of 2012 was generated. The journals were ranked based on the number of papers cited from each. This citation rank list was compared with the h index and impact factor rank lists. The rank list was also examined to see if the concept of core journals could be validated for neurosurgical literature using Bradford's law. A total of 22,850 papers spread across 2522 journals were cited in neurosurgical literature over 3 months. Although the top 10 journals were the same, irrespective of ranking criterion (h index, impact factor, citation ranking), the 3 rank lists were not congruent. The top 25% of cited articles obeyed the Bradford distribution; beyond this, there was a zone of increased scatter. Six core journals were identified for neurosurgery. The core journals for neurosurgery were identified to be Journal of Neurosurgery, Neurosurgery, Spine, Acta Neurochirurgica, Stroke, and Journal of Neurotrauma. A list of core journals could similarly be generated for every subject. This would facilitate a focused reading to keep abreast of current knowledge. Collated across specialties, these journals could depict the current status of medical science.

Intensity of geomagnetic field in the Precambrian and evolution of the Earth's deep interior

NASA Astrophysics Data System (ADS)

Smirnov, A. V.

2017-09-01

Reliable data on the paleointensity of the geomagnetic field can become an important source of information both about the mechanisms of generation of the field at present and in the past, and about the internal structure of the Earth, especially the structure and evolution of its core. Unfortunately, the reliability of these data remains a serious problem of paleomagnetic research because of the limitations of experimental methods, and the complexity and diversity of rocks and their magnetic carriers. This is true even for relatively "young" Phanerozoic rocks, but investigation of Precambrian rocks is associated with many additional difficulties. As a consequence, our current knowledge of paleointensity, especially in the Precambrian period, is still very limited. The data limitations do not preclude attempts to use the currently available paleointensity results to analyze the evolution and characteristics of the Earth's internal structure, such as the age of the Earth's solid inner core or thermal conductivity in the liquid core. However, such attempts require considerable caution in handling data. In particular, it has now been reliably established that some results on the Precambrian paleointensity overestimate the true paleofield strength. When the paleointensity overestimates are excluded from consideration, the range of the field strength changes in the Precambrian does not exceed the range of its variation in the Phanerozoic. This result calls into question recent assertions that the Earth's inner core formed in the Mesoproterozoic, about 1.3 billion years ago, triggering a statistically significant increase in the long-term average field strength. Instead, our analysis has shown that the quantity and quality of the currently available data on the Precambrian paleointensity are insufficient to estimate the age of the solid inner core and, therefore, cannot be useful for solving the problem of the thermal conductivity of the Earth's core. The data are consistent with very young or very "old" inner core ages and, correspondingly, with high or low values of core thermal conductivity.

Designing nursing excellence through a National Quality Forum nurse scholar program.

PubMed

Neumann, Julie A; Brady-Schluttner, Katherine A; Attlesey-Pries, Jacqueline M; Twedell, Diane M

2010-01-01

Closing the knowledge gap for current practicing nurses in the Institute of Medicine (IOM) core competencies is critical to providing safe patient care. The National Quality Forum (NQF) nurse scholar program is one organization's journey to close the gap in the IOM core competencies in a large teaching organization. The NQF nurse scholar program is positioned to provide a plan to assist current nurses to accelerate their learning about quality improvement, evidence-based practice, and informatics, 3 of the core competencies identified by the IOM, and focus on application of skills to NQF nurse-sensitive measures. Curriculum outline, educational methodologies, administrative processes, and aims of the project are discussed.

Inductive Position Sensor

NASA Technical Reports Server (NTRS)

Youngquist, Robert C. (Inventor); Simmons, Stephen M. (Inventor)

2015-01-01

An inductive position sensor uses three independent inductors inductively coupled by a common medium such as air. First and second inductors are separated by a fixed distance with the first inductor's axial core and second inductor's axial core maintained parallel to one another. A third inductor is disposed between the first and second inductors with the third inductor's axial core being maintained parallel to those of the first and second inductors. The combination of the first and second inductors are configured for relative movement with the third inductor's axial core remaining parallel to those of the first and second inductors as distance changes from the third inductor to each of the first inductor and second inductor. An oscillating current can be supplied to at least one of the three inductors, while voltage induced in at least one of the three inductors not supplied with the oscillating current is measured.

Field analysis & eddy current losses calculation in five-phase tubular actuator

NASA Astrophysics Data System (ADS)

Waindok, Andrzej; Tomczuk, Bronislaw

2017-12-01

Field analysis including eddy currents in the magnetic core of five-phase permanent magnet tubular linear actuator (TLA) has been carried out. The eddy currents induced in the magnetic core cause the losses which have been calculated. The results from 2D finite element (FE) analysis have been compared with those from 3D calculations. The losses in the mover of the five-phase actuator are much lower than the losses in its stator. That is why the former ones can be neglected in the computer aided designing. The calculation results have been verified experimentally

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

The Driving Magnetic Field and Reconnection in CME/Flare Eruptions and Coronal Jets

NASA Technical Reports Server (NTRS)

Moore, Ronald L.

2010-01-01

Signatures of reconnection in major CME (coronal mass ejection)/flare eruptions and in coronal X-ray jets are illustrated and interpreted. The signatures are magnetic field lines and their feet that brighten in flare emission. CME/flare eruptions are magnetic explosions in which: 1. The field that erupts is initially a closed arcade. 2. At eruption onset, most of the free magnetic energy to be released is not stored in field bracketing a current sheet, but in sheared field in the core of the arcade. 3. The sheared core field erupts by a process that from its start or soon after involves fast "tether-cutting" reconnection at an initially small current sheet low in the sheared core field. If the arcade has oppositely-directed field over it, the eruption process from its start or soon after also involves fast "breakout" reconnection at an initially small current sheet between the arcade and the overarching field. These aspects are shown by the small area of the bright field lines and foot-point flare ribbons in the onset of the eruption. 4. At either small current sheet, the fast reconnection progressively unleashes the erupting core field to erupt with progressively greater force. In turn, the erupting core field drives the current sheet to become progressively larger and to undergo progressively greater fast reconnection in the explosive phase of the eruption, and the flare arcade and ribbons grow to become comparable to the pre-eruption arcade in lateral extent. In coronal X-ray jets: 1. The magnetic energy released in the jet is built up by the emergence of a magnetic arcade into surrounding unipolar "open" field. 2. A simple jet is produced when a burst of reconnection occurs at the current sheet between the arcade and the open field. This produces a bright reconnection jet and a bright reconnection arcade that are both much smaller in diameter that the driving arcade. 3. A more complex jet is produced when the arcade has a sheared core field and undergoes an ejective eruption in the manner of a miniature CME/flare eruption. The jet is then a combination of a miniature CME and the products of more widely distributed reconnection of the erupting arcade with the open field than in simple jets.

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.

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.

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.

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.

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.

The use of modified scaling factors in the design of high-power, non-linear, transmitting rod-core antennas

NASA Astrophysics Data System (ADS)

Jordan, Jared Williams; Dvorak, Steven L.; Sternberg, Ben K.

2010-10-01

In this paper, we develop a technique for designing high-power, non-linear, transmitting rod-core antennas by using simple modified scale factors rather than running labor-intensive numerical models. By using modified scale factors, a designer can predict changes in magnetic moment, inductance, core series loss resistance, etc. We define modified scale factors as the case when all physical dimensions of the rod antenna are scaled by p, except for the cross-sectional area of the individual wires or strips that are used to construct the core. This allows one to make measurements on a scaled-down version of the rod antenna using the same core material that will be used in the final antenna design. The modified scale factors were derived from prolate spheroidal analytical expressions for a finite-length rod antenna and were verified with experimental results. The modified scaling factors can only be used if the magnetic flux densities within the two scaled cores are the same. With the magnetic flux density constant, the two scaled cores will operate with the same complex permeability, thus changing the non-linear problem to a quasi-linear problem. We also demonstrate that by holding the number of turns times the drive current constant, while changing the number of turns, the inductance and core series loss resistance change by the number of turns squared. Experimental measurements were made on rod cores made from varying diameters of black oxide, low carbon steel wires and different widths of Metglas foil. Furthermore, we demonstrate that the modified scale factors work even in the presence of eddy currents within the core material.

Optimize out-of-core thermionic energy conversion for nuclear electric propulsion

NASA Technical Reports Server (NTRS)

Morris, J. F.

1977-01-01

Current designs for out of core thermionic energy conversion (TEC) to power nuclear electric propulsion (NEP) were evaluated. Approaches to improve out of core TEC are emphasized and probabilities for success are indicated. TEC gains are available with higher emitter temperatures and greater power densities. Good potentialities for accommodating external high temperature, high power density TEC with heat pipe cooled reactors exist.

Electrically operated magnetic switch designed to display reduced leakage inductance

DOEpatents

Cook, E.G.

1994-05-10

An electrically operated magnetic switch is disclosed herein for use in opening and closing a circuit between two terminals depending upon the voltage across these terminals. The switch so disclosed is comprised of a ferrite core in the shape of a toroid having opposing ends and opposite inner and outer sides and an arrangement of electrically conductive components defining at least one current flow path which makes a number of turns around the core. This arrangement of components includes a first plurality of electrically conducive rigid rods parallel with and located outside the outer side of the core and a second plurality of electrically conductive rigid rods parallel with and located inside the inner side of the core. The arrangement also includes means for electrically connecting these rods together so that the define the current flow path. In one embodiment, this latter means uses rigid cross-tab means. In another, preferred embodiment, printed circuits on rigid dielectric substrates located on opposite ends of the core are utilized to interconnect the rods together. 10 figures.

Characterization of a CMOS sensing core for ultra-miniature wireless implantable temperature sensors with application to cryomedicine.

PubMed

Khairi, Ahmad; Thaokar, Chandrajit; Fedder, Gary; Paramesh, Jeyanandh; Rabin, Yoed

2014-09-01

In effort to improve thermal control in minimally invasive cryosurgery, the concept of a miniature, wireless, implantable sensing unit has been developed recently. The sensing unit integrates a wireless power delivery mechanism, wireless communication means, and a sensing core-the subject matter of the current study. The current study presents a CMOS ultra-miniature PTAT temperature sensing core and focuses on design principles, fabrication of a proof-of-concept, and characterization in a cryogenic environment. For this purpose, a 100 μm × 400 μm sensing core prototype has been fabricated using a 130 nm CMOS process. The senor has shown to operate between -180°C and room temperature, to consume power of less than 1 μW, and to have an uncertainty range of 1.4°C and non-linearity of 1.1%. Results of this study suggest that the sensing core is ready to be integrated in the sensing unit, where system integration is the subject matter of a parallel effort. Copyright © 2014 IPEM. Published by Elsevier Ltd. All rights reserved.

Measurement method for determining the magnetic hysteresis effects of reluctance actuators by evaluation of the force and flux variation.

PubMed

Vrijsen, N H; Jansen, J W; Compter, J C; Lomonova, E A

2013-07-01

A measurement method is presented which identifies the magnetic hysteresis effects present in the force of linear reluctance actuators. The measurement method is applied to determine the magnetic hysteresis in the force of an E-core reluctance actuator, with and without pre-biasing permanent magnet. The force measurements are conducted with a piezoelectric load cell (Kistler type 9272). This high-bandwidth force measurement instrument is identified in the frequency domain using a voice-coil actuator that has negligible magnetic hysteresis and eddy currents. Specifically, the phase delay between the current and force of the voice-coil actuator is used for the calibration of the measurement instrument. This phase delay is also obtained by evaluation of the measured force and flux variation in the E-core actuator, both with and without permanent magnet on the middle tooth. The measured magnetic flux variation is used to distinguish the phase delay due to magnetic hysteresis from the measured phase delay between the current and the force of the E-core actuator. Finally, an open loop steady-state ac model is presented that predicts the magnetic hysteresis effects in the force of the E-core actuator.

When core competence is not enough: functional interplay of the DEAD-box helicase core with ancillary domains and auxiliary factors in RNA binding and unwinding.

PubMed

Rudolph, Markus G; Klostermeier, Dagmar

2015-08-01

DEAD-box helicases catalyze RNA duplex unwinding in an ATP-dependent reaction. Members of the DEAD-box helicase family consist of a common helicase core formed by two RecA-like domains. According to the current mechanistic model for DEAD-box mediated RNA unwinding, binding of RNA and ATP triggers a conformational change of the helicase core, and leads to formation of a compact, closed state. In the closed conformation, the two parts of the active site for ATP hydrolysis and of the RNA binding site, residing on the two RecA domains, become aligned. Closing of the helicase core is coupled to a deformation of the RNA backbone and destabilization of the RNA duplex, allowing for dissociation of one of the strands. The second strand remains bound to the helicase core until ATP hydrolysis and product release lead to re-opening of the core. The concomitant disruption of the RNA binding site causes dissociation of the second strand. The activity of the helicase core can be modulated by interaction partners, and by flanking N- and C-terminal domains. A number of C-terminal flanking regions have been implicated in RNA binding: RNA recognition motifs (RRM) typically mediate sequence-specific RNA binding, whereas positively charged, unstructured regions provide binding sites for structured RNA, without sequence-specificity. Interaction partners modulate RNA binding to the core, or bind to RNA regions emanating from the core. The functional interplay of the helicase core and ancillary domains or interaction partners in RNA binding and unwinding is not entirely understood. This review summarizes our current knowledge on RNA binding to the DEAD-box helicase core and the roles of ancillary domains and interaction partners in RNA binding and unwinding by DEAD-box proteins.

The Relationship of Core Strength and Activation and Performance on Three Functional Movement Screens.

PubMed

Johnson, Caleb D; Whitehead, Paul N; Pletcher, Erin R; Faherty, Mallory S; Lovalekar, Mita T; Eagle, Shawn R; Keenan, Karen A

2018-04-01

Johnson, CD, Whitehead, PN, Pletcher, ER, Faherty, MS, Lovalekar, MT, Eagle, SR, and Keenan, KA. The relationship of core strength and activation and performance on three functional movement screens. J Strength Cond Res 32(4): 1166-1173, 2018-Current measures of core stability used by clinicians and researchers suffer from several shortcomings. Three functional movement screens appear, at face-value, to be dependent on the ability to activate and control core musculature. These 3 screens may present a viable alternative to current measures of core stability. Thirty-nine subjects completed a deep squat, trunk stability push-up, and rotary stability screen. Scores on the 3 screens were summed to calculate a composite score (COMP). During the screens, muscle activity was collected to determine the length of time that the bilateral erector spinae, rectus abdominis, external oblique, and gluteus medius muscles were active. Strength was assessed for core muscles (trunk flexion and extension, trunk rotation, and hip abduction and adduction) and accessory muscles (knee flexion and extension and pectoralis major). Two ordinal logistic regression equations were calculated with COMP as the outcome variable, and: (a) core strength and accessory strength, (b) only core strength. The first model was significant in predicting COMP (p = 0.004) (Pearson's Chi-Square = 149.132, p = 0.435; Nagelkerke's R-Squared = 0.369). The second model was significant in predicting COMP (p = 0.001) (Pearson's Chi-Square = 148.837, p = 0.488; Nagelkerke's R-Squared = 0.362). The core muscles were found to be active for most screens, with percentages of "time active" for each muscle ranging from 54-86%. In conclusion, performance on the 3 screens is predicted by core strength, even when accounting for "accessory" strength variables. Furthermore, it seems the screens elicit wide-ranging activation of core muscles. Although more investigation is needed, these screens, collectively, seem to be a good assessment of core strength.

Engineering of high performance supercapacitor electrode based on Fe-Ni/Fe{sub 2}O{sub 3}-NiO core/shell hybrid nanostructures

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

Singh, Ashutosh K., E-mail: ashuvishen@gmail.com, E-mail: aksingh@bose.res.in; Mandal, Kalyan

The present work reports on fabrication and supercapacitor applications of a core/shell Fe-Ni/Fe{sub 2}O{sub 3}-NiO hybrid nanostructures (HNs) electrode. The core/shell Fe-Ni/Fe{sub 2}O{sub 3}-NiO hybrid nanostructures have been fabricated through a two step method (nanowire fabrication and their controlled oxidation). The 1D hybrid nanostructure consists of highly porous shell layer (redox active materials NiO and Fe{sub 2}O{sub 3}) and the conductive core (FeNi nanowire). Thus, the highly porous shell layer allows facile electrolyte diffusion as well as faster redox reaction kinetics; whereas the conductive FeNi nanowire core provides the proficient express way for electrons to travel to the current collector,more » which helps in the superior electrochemical performance. The core/shell Fe-Ni/Fe{sub 2}O{sub 3}-NiO hybrid nanostructures electrode based supercapacitor shows very good electrochemical performances in terms of high specific capacitance nearly 1415 F g{sup −1} at a current density of 2.5 A g{sup −1}, excellent cycling stability and rate capability. The high quality electrochemical performance of core/shell hybrid nanostructures electrode shows its potential as an alternative electrode for forthcoming supercapacitor devices.« less

St. Petersburg Coastal and Marine Science Center's Core Archive Portal

USGS Publications Warehouse

Reich, Chris; Streubert, Matt; Dwyer, Brendan; Godbout, Meg; Muslic, Adis; Umberger, Dan

2012-01-01

This Web site contains information on rock cores archived at the U.S. Geological Survey (USGS) St. Petersburg Coastal and Marine Science Center (SPCMSC). Archived cores consist of 3- to 4-inch-diameter coral cores, 1- to 2-inch-diameter rock cores, and a few unlabeled loose coral and rock samples. This document - and specifically the archive Web site portal - is intended to be a 'living' document that will be updated continually as additional cores are collected and archived. This document may also contain future references and links to a catalog of sediment cores. Sediment cores will include vibracores, pushcores, and other loose sediment samples collected for research purposes. This document will: (1) serve as a database for locating core material currently archived at the USGS SPCMSC facility; (2) provide a protocol for entry of new core material into the archive system; and, (3) set the procedures necessary for checking out core material for scientific purposes. Core material may be loaned to other governmental agencies, academia, or non-governmental organizations at the discretion of the USGS SPCMSC curator.

Current status of the EPOS WG4 - GNSS and Other Geodetic Data

NASA Astrophysics Data System (ADS)

Fernandes, Rui; Bastos, Luísa; Bruyninx, Carine; D'Agostino, Nicola; Dousa, Jan; Ganas, Athanassios; Lidberg, Martin; Nocquet, Jean-Mathieu

2013-04-01

WG4 - "EPOS Geodetic Data and Other Geodetic Data" is the Working Group of the EPOS project in charge of defining and preparing the integration of the existing Pan-European Geodetic Infrastructures that will support the European Geosciences, which is the ultimate goal of the EPOS project. The WG4 is formed by representatives of the participating EPOS countries (23) but it is also open to the entire geodetic community. In fact, WG4 also includes members from countries that formally are not part of the current phase of EPOS. In an ongoing effort, the majority of existing GNSS Research Infrastructures in Europe were identified. The current database, available at http://epos-couch.cloudant.com/epos-couch/_design/epos-couch/, lists a total of 50 Research Infrastructures managing a total of 1534 GNSS CORS sites. This presentation intends to detail the work being produced within the working group WG4 related with the definition of strategies towards the implementation of the best solutions that will permit to the end-users, and in particular geo-scientists, to access the geodetic data, derived solutions, and associated metadata using transparent and uniform processes. The first step toward the design of an implementation and business plan is the definition of the core services for geodetic data within EPOS. In this talk, we will present the current status of the discussion about the content of core services. Three levels of core services could be distinguished, for which their content need to be defined. The 3 levels are: (1) the core services associated to data (diffusion, archive, long-term preservation, quality check, rapid analysis) (2) core services associated to geodetic products (analysis, products definition like position time series, velocity field and Zenithal Total Delay) (3) User oriented services (reference frames, real-time solutions for early warning systems, strain rate maps, meteorology, space weather, …). Current propositions and remaining open questions will be discussed.

A current review of core decompression in the treatment of osteonecrosis of the femoral head.

PubMed

Pierce, Todd P; Jauregui, Julio J; Elmallah, Randa K; Lavernia, Carlos J; Mont, Michael A; Nace, James

2015-09-01

The review describes the following: (1) how traditional core decompression is performed, (2) adjunctive treatments, (3) multiple percutaneous drilling technique, and (4) the overall outcomes of these procedures. Core decompression has optimal outcomes when used in the earliest, precollapse disease stages. More recent studies have reported excellent outcomes with percutaneous drilling. Furthermore, adjunct treatment methods combining core decompression with growth factors, bone morphogenic proteins, stem cells, and bone grafting have demonstrated positive results; however, larger randomized trial is needed to evaluate their overall efficacy.

Advanced Wireless Integrated Navy Network - AWINN

DTIC Science & Technology

2005-09-30

progress report No. 3 on AWINN hardware and software configurations of smart , wideband, multi-function antennas, secure configurable platform, close-in...results to the host PC via a UART soft core. The UART core used is a proprietary Xilinx core which incorporates features described in National...current software uses wheel odometry and visual landmarks to create a map and estimate position on an internal x, y grid . The wheel odometry provides a

The hyacinth project

NASA Astrophysics Data System (ADS)

Francis, T.

2003-04-01

HYACINTH is the acronym for "Development of HYACE tools in new tests on Hydrates". The project is being carried out by a consortium of six companies and academic institutions from Germany, The Netherlands and the United Kingdom. It is a European Framework Five project whose objective is to bring the pressure corers developed in the earlier HYACE project, together with new core handling technology developed in the HYACINTH project, to the operational stage. Our philosophy is that if all one does with a pressure core is to bleed off the gas it contains, a major scientific opportunity has been missed. The current system enables pressure cores to be acquired, then transferred, without loss of pressure, into laboratory chambers so that they can be geophysically logged. The suite of equipment - HYACE Rotary Corer (HRC), Fugro Pressure Corer (FPC), Shear Transfer Chamber (STC), Logging Chamber (LC), Storage Chamber (SC) and Vertical Multi-Sensor Core Logger (V-MSCL) - will be briefly described. Other developments currently in progress to extend the capabilities of the system will be summarised: - to allow electrical resistivity logging of the pressure cores - to enable pressurised sub-samples to be taken from the cores - to facilitate microbiological experiments on pressurised sub-samples The first scientific results obtained with the HYACE/HYACINTH technology were achieved on ODP Leg 204 and are the subject of another talk at this meeting.

Magnetic vortex racetrack memory

NASA Astrophysics Data System (ADS)

Geng, Liwei D.; Jin, Yongmei M.

2017-02-01

We report a new type of racetrack memory based on current-controlled movement of magnetic vortices in magnetic nanowires with rectangular cross-section and weak perpendicular anisotropy. Data are stored through the core polarity of vortices and each vortex carries a data bit. Besides high density, non-volatility, fast data access, and low power as offered by domain wall racetrack memory, magnetic vortex racetrack memory has additional advantages of no need for constrictions to define data bits, changeable information density, adjustable current magnitude for data propagation, and versatile means of ultrafast vortex core switching. By using micromagnetic simulations, current-controlled motion of magnetic vortices in cobalt nanowire is demonstrated for racetrack memory applications.

GaAs/AlGaAs core multishell nanowire-based light-emitting diodes on Si.

PubMed

Tomioka, Katsuhiro; Motohisa, Junichi; Hara, Shinjiroh; Hiruma, Kenji; Fukui, Takashi

2010-05-12

We report on integration of GaAs nanowire-based light-emitting-diodes (NW-LEDs) on Si substrate by selective-area metalorganic vapor phase epitaxy. The vertically aligned GaAs/AlGaAs core-multishell nanowires with radial p-n junction and NW-LED array were directly fabricated on Si. The threshold current for electroluminescence (EL) was 0.5 mA (current density was approximately 0.4 A/cm(2)), and the EL intensity superlinearly increased with increasing current injections indicating superluminescence behavior. The technology described in this letter could help open new possibilities for monolithic- and on-chip integration of III-V NWs on Si.

Fault current limiter

DOEpatents

Darmann, Francis Anthony

2013-10-08

A fault current limiter (FCL) includes a series of high permeability posts for collectively define a core for the FCL. A DC coil, for the purposes of saturating a portion of the high permeability posts, surrounds the complete structure outside of an enclosure in the form of a vessel. The vessel contains a dielectric insulation medium. AC coils, for transporting AC current, are wound on insulating formers and electrically interconnected to each other in a manner such that the senses of the magnetic field produced by each AC coil in the corresponding high permeability core are opposing. There are insulation barriers between phases to improve dielectric withstand properties of the dielectric medium.

Implementation of kernels on the Maestro processor

NASA Astrophysics Data System (ADS)

Suh, Jinwoo; Kang, D. I. D.; Crago, S. P.

Currently, most microprocessors use multiple cores to increase performance while limiting power usage. Some processors use not just a few cores, but tens of cores or even 100 cores. One such many-core microprocessor is the Maestro processor, which is based on Tilera's TILE64 processor. The Maestro chip is a 49-core, general-purpose, radiation-hardened processor designed for space applications. The Maestro processor, unlike the TILE64, has a floating point unit (FPU) in each core for improved floating point performance. The Maestro processor runs at 342 MHz clock frequency. On the Maestro processor, we implemented several widely used kernels: matrix multiplication, vector add, FIR filter, and FFT. We measured and analyzed the performance of these kernels. The achieved performance was up to 5.7 GFLOPS, and the speedup compared to single tile was up to 49 using 49 tiles.

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

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

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

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.

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.

Monitoring of NMR porosity changes in the full-size core salvage through the drying process

NASA Astrophysics Data System (ADS)

Fattakhov, Artur; Kosarev, Victor; Doroginitskii, Mikhail; Skirda, Vladimir

2015-04-01

Currently the principle of nuclear magnetic resonance (NMR) is one of the most popular technologies in the field of borehole geophysics and core analysis. Results of NMR studies allow to calculate the values of the porosity and permeability of sedimentary rocks with sufficient reliability. All standard tools for the study of core salvage on the basis of NMR have significant limitations: there is considered only long relaxation times corresponding to the mobile formation fluid. Current trends in energy obligate to move away from conventional oil to various alternative sources of energy. One of these sources are deposits of bitumen and high-viscosity oil. In Kazan (Volga Region) Federal University (Russia) there was developed a mobile unit for the study of the full-length core salvage by the NMR method ("NMR-Core") together with specialists of "TNG-Group" (a company providing maintenance services to oil companies). This unit is designed for the study of core material directly on the well, after removing it from the core receiver. The maximum diameter of the core sample may be up to 116 mm, its length (or length of the set of samples) may be up to 1000 mm. Positional precision of the core sample relative to the measurement system is 1 mm, and the spatial resolution along the axis of the core is 10 mm. Acquisition time of the 1 m core salvage varies depending on the mode of research and is at least 20 minutes. Furthermore, there is implemented a special investigation mode of the core samples with super small relaxation times (for example, heavy oil) is in the tool. The aim of this work is tracking of the NMR porosity changes in the full-size core salvage in time. There was used a water-saturated core salvage from the shallow educational well as a sample. The diameter of the studied core samples is 93 mm. There was selected several sections length of 1m from the 200-meter coring interval. The studied core samples are being measured several times. The time interval between the measurements is from 1 hour to 48 hours. Making the measurements it possible to draw conclusions about that the processes of NMR porosity changes in time as a result of evaporation of the part of fluid from the surface layer of the core salvage and suggest a core analysis technique directly on the well. This work is supported by the grant of Ministry of Education and Science of the Russian Federation (project No. 02.G25.31.0029).

Examining the importance of incorporating emergency preparedness and disaster training core competencies into allied health curricula.

PubMed

Curtis, Tammy

2015-01-01

Preparation for responding to emergency events that does not warrant outside help beyond the local community resources or responding to disaster events that is beyond the capabilities of the local community both require first responders and healthcare professionals to have interdisciplinary skills needed to function as a team for saving lives. To date, there is no core emergency preparedness and disaster planning competencies that have been standardized at all levels across the various allied health curricula disciplines. To identify if emergency preparedness and disaster training content are currently being taught in allied health program courses, to identify possible gaps within allied health curricula, and to explore the perceptions of allied health college educators for implementing emergency preparedness and disaster training core competencies into their existing curricula, if not already included. A quantitative Internet-based survey was conducted in 2013. Convenient sample. Fifty-one allied health college educators completed the survey. Descriptive statistics indicated that the majority of allied health college instructors do not currently teach emergency preparedness and disaster training core competency content within their current allied health discipline; however, their perceived level of importance for inclusion of the competencies was high. The results of this study supported the need for developing and establishing a basic national set of standardized core emergency preparedness and disaster planning competencies at all levels across various allied health curricula disciplines to ensure victims receive the best patient care and have the best possible chance of survival.

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

Embedded binaries and their dense cores

NASA Astrophysics Data System (ADS)

Sadavoy, Sarah I.; Stahler, Steven W.

2017-08-01

We explore the relationship between young, embedded binaries and their parent cores, using observations within the Perseus Molecular Cloud. We combine recently published Very Large Array observations of young stars with core properties obtained from Submillimetre Common-User Bolometer Array 2 observations at 850 μm. Most embedded binary systems are found towards the centres of their parent cores, although several systems have components closer to the core edge. Wide binaries, defined as those systems with physical separations greater than 500 au, show a tendency to be aligned with the long axes of their parent cores, whereas tight binaries show no preferred orientation. We test a number of simple, evolutionary models to account for the observed populations of Class 0 and I sources, both single and binary. In the model that best explains the observations, all stars form initially as wide binaries. These binaries either break up into separate stars or else shrink into tighter orbits. Under the assumption that both stars remain embedded following binary break-up, we find a total star formation rate of 168 Myr-1. Alternatively, one star may be ejected from the dense core due to binary break-up. This latter assumption results in a star formation rate of 247 Myr-1. Both production rates are in satisfactory agreement with current estimates from other studies of Perseus. Future observations should be able to distinguish between these two possibilities. If our model continues to provide a good fit to other star-forming regions, then the mass fraction of dense cores that becomes stars is double what is currently believed.

Electrical Current Leakage and Open-Core Threading Dislocations in AlGaN-Based Deep Ultraviolet Light-Emitting Diodes.

DOE PAGES

Moseley, Michael William; Allerman, Andrew A.; Crawford, Mary H.; ...

2014-08-04

Electrical current transport through leakage paths in AlGaN-based deep ultraviolet (DUV) lightemitting diodes (LEDs) and their effect on LED performance are investigated. Open-core threading dislocations, or nanopipes, are found to conduct current through nominally insulating Al0.7Ga0.3N layers and limit the performance of DUV-LEDs. A defect-sensitive phosphoric acid etch reveals these opencore threading dislocations in the form of large, micron-scale hexagonal etch pits visible with optical microscopy, while closed-core screw-, edge-, and mixed-type threading dislocations are represented by smaller and more numerous nanometer-scale pits visible by atomic-force microscopy. The electrical and optical performances of DUV-LEDs fabricated on similar Si-doped Al0.7Ga0.3N templatesmore » are found to have a strong correlation to the density of these nanopipes, despite their small fraction (<0.1% in this study) of the total density of threading dislocations.« less

What Should Common Core Assessments Measure?

ERIC Educational Resources Information Center

Chandler, Kayla; Fortune, Nicholas; Lovett, Jennifer N.; Scherrer, Jimmy

2016-01-01

The Common Core State Standards for mathematics promote ideals about learning mathematics by providing specific standards focused on conceptual understanding and incorporating practices in which students must participate to develop conceptual understanding. Thus, how we define learning is pivotal because our current definition isn't aligned with…

Modeling of grain-oriented Si-steel and amorphous alloy iron core under ferroresonance using Jiles-Atherton hysteresis method

NASA Astrophysics Data System (ADS)

Sima, Wenxia; Zou, Mi; Yang, Ming; Yang, Qing; Peng, Daixiao

2018-05-01

Amorphous alloy is increasingly widely used in the iron core of power transformer due to its excellent low loss performance. However, its potential harm to the power system is not fully studied during the electromagnetic transients of the transformer. This study develops a simulation model to analyze the effect of transformer iron core materials on ferroresonance. The model is based on the transformer π equivalent circuit. The flux linkage-current (ψ-i) Jiles-Atherton reactor is developed in an Electromagnetic Transients Program-Alternative Transients Program and is used to represent the magnetizing branches of the transformer model. Two ferroresonance cases are studied to compare the performance of grain-oriented Si-steel and amorphous alloy cores. The ferroresonance overvoltage and overcurrent are discussed under different system parameters. Results show that amorphous alloy transformer generates higher voltage and current than those of grain-oriented Si-steel transformer and significantly harms the power system safety.

Current OCT Approaches Do Not Reliably Identify TCFAs

PubMed Central

Brezinski, Mark E.; Harjai, Kishore J

2017-01-01

It is now clearly established that Thin-Capped Fibroatheromas (TCFAs) lead to most Acute Coronary Syndromes (ACSs). The ability to selectively intervene on TCFAs predisposed to rupture and ACSs would dramatically alter the practice of cardiology. While the ability of OCT to identify thin walled plaques at micron scale resolutions has represented a major advance, it is a misconception that it can reliably identify TCFAs. One major reason is that the ‘diffuse border’ criteria currently used to determine ‘lipid plaque’ is almost undoubtedly from high scattering in the intima and not because of core composition (necrotic core). A second reason is that, rather than looking at lipid collections, studies need to be focused on identifying necrotic cores with OCT. Necrotic cores are characteristic of TCFAs and not lipid collections. Numerous other OCT approaches are available which can potentially accurately assess TCFAs, but these have not been aggressively pursed which we believe likely stems in part from the misconceptions over the efficacy of ‘diffuse borders’. PMID:29250457

JPRS Report, Science & Technology, China: Energy

DTIC Science & Technology

1988-06-29

capacity. There are currently two types of HTGR reactor designs: the particle-bed core , which uses spherical fuel elements, and the rod type core , in...and trial operating experience with the HTGR reactor. Its main design features are as follows. 1. A particle-bed core , continuous fueling and...Favorable for Development of Small-Scale HTGR (Xu Jiming; HE DONGLI GONGCHENG, Feb 88) 47 ERRATUM: In JPRS-CEN-88-003 of 25 April 1988 in article

Phase relations in iron-rich systems and implications for the earth's core

NASA Technical Reports Server (NTRS)

Anderson, William W.; Svendsen, Bob; Ahrens, Thomas J.

1987-01-01

Recent experimental data concerning the properties of iron, iron sulfide, and iron oxide at high pressures are combined with theoretical arguments to constrain the probable behavior of the Fe-rich portions of the Fe-O and Fe-S phase diagrams. Phase diagrams are constructed for the Fe-S-O system at core pressures and temperatures. These properties are used to evaluate the current temperature distribution and composition of the core.

Dynamical Upheaval in Ice Giant Formation: A Solution to the Fine-tuning Problem in the Formation Story

NASA Astrophysics Data System (ADS)

Frelikh, Renata; Murray-Clay, Ruth

2018-04-01

We report on our recent theoretical work, where we suggest that a protoplanetary disk dynamical instability may have played a crucial role in determining the atmospheric size of the solar system’s ice giants. In contrast to the gas giants, the intermediate-size ice giants never underwent runaway gas accretion in a full gas disk. However, as their substantial core masses are comparable to those of the gas giants, they would have gone runaway, given enough time. In the standard scenario, the ice giants stay at roughly their current size for most of the disk lifetime, undergoing period of slow gas accretion onto ~full-sized cores that formed early-on. The gas disk dissipates before the ice giants accumulate too much gas, but we believe this is fine tuned. A considerable amount of solids is observed in outer disks in mm-to-cm sized particles (pebbles). Assisted by gas drag, these pebbles rapidly accrete onto cores. This would cause the growing ice giants to exceed their current core masses, and quickly turn into gas giants. To resolve this problem, we propose that Uranus and Neptune stayed small for the bulk of the disk lifetime. They only finished their core and atmospheric growth in a short timeframe just as the disk gas dissipated, accreting most of their gas from a disk depleted to ~1% of its original mass. The ice giants have atmospheric mass fractions comparable to the disk gas-to-solid ratio of this depleted disk. This coincides with a disk dynamical upheaval onset by the depletion of gas. We propose that the cores started growing closer-in, where they were kept small by proximity to Jupiter and Saturn. As the gas cleared, the cores were kicked out by the gas giants. Then, they finished their core growth and accreted their atmospheres from the remaining, sparse gas at their current locations. We predict that the gas giants may play a key role in forming intermediate-size atmospheres in the outer disk.

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

Hierarchical core-shell structure of ZnO nanorod@NiO/MoO₂ composite nanosheet arrays for high-performance supercapacitors.

PubMed

Hou, Sucheng; Zhang, Guanhua; Zeng, Wei; Zhu, Jian; Gong, Feilong; Li, Feng; Duan, Huigao

2014-08-27

A hierarchical core-shell structure of ZnO nanorod@NiO/MoO2 composite nanosheet arrays on nickel foam substrate for high-performance supercapacitors was constructed by a two-step solution-based method involving two hydrothermal processes followed by a calcination treatment. Compared to one composed of pure NiO/MoO2 composite nanosheets, the hierarchical core-shell structure electrode displays better pseudocapacitive behaviors in 2 M KOH, including high areal specific capacitance values of 1.18 F cm(-2) at 5 mA cm(-2) and 0.6 F cm(-2) at 30 mA cm(-2) as well as relatively good rate capability at high current densities. Furthermore, it also shows remarkable cycle stability, remaining at 91.7% of the initial value even after 4000 cycles at a current density of 10 mA cm(-2). The enhanced pseudocapacitive behaviors are mainly due to the unique hierarchical core-shell structure and the synergistic effect of combining ZnO nanorod arrays and NiO/MoO2 composite nanosheets. This novel hierarchical core-shell structure shows promise for use in next-generation supercapacitors.

Electronic properties of core-shell nanowire resonant tunneling diodes

PubMed Central

2014-01-01

The electronic sub-band structure of InAs/InP/InAs/InP/InAs core-shell nanowire resonant tunneling diodes has been investigated in the effective mass approximation by varying the core radius and the thickness of the InP barriers and InAs shells. A top-hat, double-barrier potential profile and optimal energy configuration are obtained for core radii and surface shells >10 nm, InAs middle shells <10 nm, and 5 nm InP barriers. In this case, two sub-bands exist above the Fermi level in the InAs middle shell which belongs to the m = 0 and m = 1 ladder of states that have similar wave functions and energies. On the other hand, the lowest m = 0 sub-band in the core falls below the Fermi level but the m = 1 states do not contribute to the current transport since they reside energetically well above the Fermi level. We compare the case of GaAs/AlGaAs/GaAs/AlGaAs/GaAs which may conduct current with smaller applied voltages due to the larger effective mass of electrons in GaAs and discuss the need for doping. PMID:25288912

Electronic properties of core-shell nanowire resonant tunneling diodes.

PubMed

Zervos, Matthew

2014-01-01

The electronic sub-band structure of InAs/InP/InAs/InP/InAs core-shell nanowire resonant tunneling diodes has been investigated in the effective mass approximation by varying the core radius and the thickness of the InP barriers and InAs shells. A top-hat, double-barrier potential profile and optimal energy configuration are obtained for core radii and surface shells >10 nm, InAs middle shells <10 nm, and 5 nm InP barriers. In this case, two sub-bands exist above the Fermi level in the InAs middle shell which belongs to the m = 0 and m = 1 ladder of states that have similar wave functions and energies. On the other hand, the lowest m = 0 sub-band in the core falls below the Fermi level but the m = 1 states do not contribute to the current transport since they reside energetically well above the Fermi level. We compare the case of GaAs/AlGaAs/GaAs/AlGaAs/GaAs which may conduct current with smaller applied voltages due to the larger effective mass of electrons in GaAs and discuss the need for doping.

Spheromak reactor with poloidal flux-amplifying transformer

DOEpatents

Furth, Harold P.; Janos, Alan C.; Uyama, Tadao; Yamada, Masaaki

1987-01-01

An inductive transformer in the form of a solenoidal coils aligned along the major axis of a flux core induces poloidal flux along the flux core's axis. The current in the solenoidal coil is then reversed resulting in a poloidal flux swing and the conversion of a portion of the poloidal flux to a toroidal flux in generating a spheromak plasma wherein equilibrium approaches a force-free, minimum Taylor state during plasma formation, independent of the initial conditions or details of the formation. The spheromak plasma is sustained with the Taylor state maintained by oscillating the currents in the poloidal and toroidal field coils within the plasma-forming flux core. The poloidal flux transformer may be used either as an amplifier stage in a moving plasma reactor scenario for initial production of a spheromak plasma or as a method for sustaining a stationary plasma and further heating it. The solenoidal coil embodiment of the poloidal flux transformer can alternately be used in combination with a center conductive cylinder aligned along the length and outside of the solenoidal coil. This poloidal flux-amplifying inductive transformer approach allows for a relaxation of demanding current carrying requirements on the spheromak reactor's flux core, reduces plasma contamination arising from high voltage electrode discharge, and improves the efficiency of poloidal flux injection.

Low temperature nano-spin filtering using a diluted magnetic semiconductor core-shell quantum dot

NASA Astrophysics Data System (ADS)

Chattopadhyay, Saikat; Sen, Pratima; Andrews, Joshep Thomas; Sen, Pranay Kumar

2014-07-01

The spin polarized electron transport properties and spin polarized tunneling current have been investigated analytically in a diluted magnetic semiconductor core-shell quantum dot in the presence of applied electric and magnetic fields. Assuming the electron wave function to satisfy WKB approximation, the electron energy eigenvalues have been calculated. The spin polarized tunneling current and the spin dependent tunneling coefficient are obtained by taking into account the exchange interaction and Zeeman splitting. Numerical estimates made for a specific diluted magnetic semiconductor, viz., Zn1-xMnxSe/ZnS core-shell quantum dot establishes the possibility of a nano-spin filter for a particular biasing voltage and applied magnetic field. Influence of applied voltage on spin polarized electron transport has been investigated in a CSQD.

Network Coding on Heterogeneous Multi-Core Processors for Wireless Sensor Networks

PubMed Central

Kim, Deokho; Park, Karam; Ro, Won W.

2011-01-01

While network coding is well known for its efficiency and usefulness in wireless sensor networks, the excessive costs associated with decoding computation and complexity still hinder its adoption into practical use. On the other hand, high-performance microprocessors with heterogeneous multi-cores would be used as processing nodes of the wireless sensor networks in the near future. To this end, this paper introduces an efficient network coding algorithm developed for the heterogenous multi-core processors. The proposed idea is fully tested on one of the currently available heterogeneous multi-core processors referred to as the Cell Broadband Engine. PMID:22164053

Resubmission of Gap Analysis Workshop for Training for Reintegration of Surgical Skills

DTIC Science & Technology

2011-10-01

the ABS, many other organizations do not have current reentry requirements but work with physicians on a case-by-case basis. Global competency...facs.org/education/ • Animal Labs • American Urological Association (AUA) Core Curriculum - http://www.auanet.org/eforms/ elearning /core

Performing an allreduce operation on a plurality of compute nodes of a parallel computer

DOEpatents

Faraj, Ahmad

2013-07-09

Methods, apparatus, and products are disclosed for performing an allreduce operation on a plurality of compute nodes of a parallel computer, each node including at least two processing cores, that include: establishing, for each node, a plurality of logical rings, each ring including a different set of at least one core on that node, each ring including the cores on at least two of the nodes; iteratively for each node: assigning each core of that node to one of the rings established for that node to which the core has not previously been assigned, and performing, for each ring for that node, a global allreduce operation using contribution data for the cores assigned to that ring or any global allreduce results from previous global allreduce operations, yielding current global allreduce results for each core; and performing, for each node, a local allreduce operation using the global allreduce results.

Thermal interaction of the core and the mantle and long-term behavior of the geomagnetic field

NASA Technical Reports Server (NTRS)

Jones, G. M.

1977-01-01

The effects of temperature changes at the earth's core-mantle boundary on the velocity field of the core are analyzed. It is assumed that the geomagnetic field is maintained by thermal convection in the outer core. A model for the thermal interaction of the core and the mantle is presented which is consistent with current views on the presence of heat sources in the core and the properties of the lower mantle. Significant long-term variations in the frequency of geomagnetic reversals may be the result of fluctuating temperatures at the core-mantle boundary, caused by intermittent convection in the lower mantle. The thermal structure of the lower mantle region D double prime, extending from 2700 to 2900 km in depth, constitutes an important test of this hypothesis and offers a means of deciding whether the geomagnetic dynamo is thermally driven.

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

Static and dynamic parasitic magnetizations and their control in superconducting accelerator dipoles

NASA Astrophysics Data System (ADS)

Collings, E. W.; Sumption, M. D.

2001-05-01

Long dipole magnets guide the particle beams in synchrotron-type high energy accelerators. In principal Cu-wound DC-excited dipoles could be designed to deliver a very uniform transverse bore field, i.e. with small or negligible harmonic (multipolar) distortion. But if the Cu is replaced by (a) superconducting strand that is (b) wound into a Rutherford cable carrying a time-varying transport current, extra magnetizations present within the windings cause distortions of the otherwise uniform field. The static (persistent-current) strand magnetization can be reduced by reducing the filament diameter, and the residue compensated or corrected by strategically placed active or passive components. The cable’s interstrand coupling currents can be controlled by increasing the interstrand contact resistance by: adjusting the level of native oxidation of the strand, coating it, or by inserting a ribbon-like core into the cable itself. Methods of locally compensating the magnetization of NbTi and Nb 3Sn strand and cable are discussed, progress in coupling-current suppression through the use of coatings and cores is reviewed, and a method of simultaneously reducing both the static and dynamic magnetizations of a NbTi cable by means of a thin Ni core is suggested.

Core-Noise Research

NASA Technical Reports Server (NTRS)

Hultgren, Lennart S.

2012-01-01

This presentation is a technical summary of and outlook for NASA-internal and NASA-sponsored external research on core noise funded by the Fundamental Aeronautics Program Subsonic Fixed Wing (SFW) Project. Sections of the presentation cover: the SFW system-level noise metrics for the 2015 (N+1), 2020 (N+2), and 2025 (N+3) timeframes; SFW strategic thrusts and technical challenges; SFW advanced subsystems that are broadly applicable to N+3 vehicle concepts, with an indication where further noise research is needed; the components of core noise (compressor, combustor and turbine noise) and a rationale for NASA's current emphasis on the combustor-noise component; the increase in the relative importance of core noise due to turbofan design trends; the need to understand and mitigate core-noise sources for high-efficiency small gas generators; and the current research activities in the core-noise area, with additional details given about forthcoming updates to NASA's Aircraft Noise Prediction Program (ANOPP) core-noise prediction capabilities, two NRA efforts (Honeywell International, Phoenix, AZ and University of Illinois at Urbana-Champaign, respectively) to improve the understanding of core-noise sources and noise propagation through the engine core, and an effort to develop oxide/oxide ceramic-matrix-composite (CMC) liners for broadband noise attenuation suitable for turbofan-core application. Core noise must be addressed to ensure that the N+3 noise goals are met. Focused, but long-term, core-noise research is carried out to enable the advanced high-efficiency small gas-generator subsystem, common to several N+3 conceptual designs, needed to meet NASA's technical challenges. Intermediate updates to prediction tools are implemented as the understanding of the source structure and engine-internal propagation effects is improved. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The SFW Quiet-Aircraft Subproject aims to develop concepts and technologies to reduce perceived community noise attributable to aircraft with minimal impact on weight and performance. This reduction of aircraft noise is critical to enabling the anticipated large increase in future air traffic.

The French initiative for scientific cores virtual curating : a user-oriented integrated approach

NASA Astrophysics Data System (ADS)

Pignol, Cécile; Godinho, Elodie; Galabertier, Bruno; Caillo, Arnaud; Bernardet, Karim; Augustin, Laurent; Crouzet, Christian; Billy, Isabelle; Teste, Gregory; Moreno, Eva; Tosello, Vanessa; Crosta, Xavier; Chappellaz, Jérome; Calzas, Michel; Rousseau, Denis-Didier; Arnaud, Fabien

2016-04-01

Managing scientific data is probably one the most challenging issue in modern science. The question is made even more sensitive with the need of preserving and managing high value fragile geological sam-ples: cores. Large international scientific programs, such as IODP or ICDP are leading an intense effort to solve this problem and propose detailed high standard work- and dataflows thorough core handling and curating. However most results derived from rather small-scale research programs in which data and sample management is generally managed only locally - when it is … The national excellence equipment program (Equipex) CLIMCOR aims at developing French facilities for coring and drilling investigations. It concerns indiscriminately ice, marine and continental samples. As part of this initiative, we initiated a reflexion about core curating and associated coring-data management. The aim of the project is to conserve all metadata from fieldwork in an integrated cyber-environment which will evolve toward laboratory-acquired data storage in a near future. In that aim, our demarche was conducted through an close relationship with field operators as well laboratory core curators in order to propose user-oriented solutions. The national core curating initiative currently proposes a single web portal in which all scientifics teams can store their field data. For legacy samples, this will requires the establishment of a dedicated core lists with associated metadata. For forthcoming samples, we propose a mobile application, under Android environment to capture technical and scientific metadata on the field. This application is linked with a unique coring tools library and is adapted to most coring devices (gravity, drilling, percussion, etc...) including multiple sections and holes coring operations. Those field data can be uploaded automatically to the national portal, but also referenced through international standards or persistent identifiers (IGSN, ORCID and INSPIRE) and displayed in international portals (currently, NOAA's IMLGS). In this paper, we present the architecture of the integrated system, future perspectives and the approach we adopted to reach our goals. We will also present in front of our poster, one of the three mobile applications, dedicated more particularly to the operations of continental drillings.

ACCEPT 2: A public library of cluster properties

NASA Astrophysics Data System (ADS)

Donahue, Megan

2012-09-01

The current public ACCEPT database of cluster properties includes radial profiles of Tx, n_elec, entropy, and cooling time. We propose to more than double the current number of clusters in ACCEPT and to expand the current suite of properties to include uniformly measured profiles of gas mass and hydrostatic equilibrium mass along with signatures of dynamical relaxation (centroid shift, power ratios, surface brightness concentration, temperature ratios) and global quantities such as core-excised Tx, Lx, and metallicities. We will explore the relationship between cool cores and dynamical relaxation, the reliability of hydrostatic mass profiles, and the dependence of the gas mass fraction on halo mass, redshift, and the degree of relaxation. ACCEPT2 will enable further community science.

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.

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

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.

Influence of arc current and pressure on non-chemical equilibrium air arc behavior

NASA Astrophysics Data System (ADS)

Yi, WU; Yufei, CUI; Jiawei, DUAN; Hao, SUN; Chunlin, WANG; Chunping, NIU

2018-01-01

The influence of arc current and pressure on the non-chemical equilibrium (non-CE) air arc behavior of a nozzle structure was investigated based on the self-consistent non-chemical equilibrium model. The arc behavior during both the arc burning and arc decay phases were discussed at different currents and different pressures. We also devised the concept of a non-equilibrium parameter for a better understanding of non-CE effects. During the arc burning phase, the increasing current leads to a decrease of the non-equilibrium parameter of the particles in the arc core, while the increasing pressure leads to an increase of the non-equilibrium parameter of the particles in the arc core. During the arc decay phase, the non-CE effect will decrease by increasing the arc burning current and the nozzle pressure. Three factors together—convection, diffusion and chemical reactions—influence non-CE behavior.

Automated Defect and Correlation Length Analysis of Block Copolymer Thin Film Nanopatterns

PubMed Central

Murphy, Jeffrey N.; Harris, Kenneth D.; Buriak, Jillian M.

2015-01-01

Line patterns produced by lamellae- and cylinder-forming block copolymer (BCP) thin films are of widespread interest for their potential to enable nanoscale patterning over large areas. In order for such patterning methods to effectively integrate with current technologies, the resulting patterns need to have low defect densities, and be produced in a short timescale. To understand whether a given polymer or annealing method might potentially meet such challenges, it is necessary to examine the evolution of defects. Unfortunately, few tools are readily available to researchers, particularly those engaged in the synthesis and design of new polymeric systems with the potential for patterning, to measure defects in such line patterns. To this end, we present an image analysis tool, which we have developed and made available, to measure the characteristics of such patterns in an automated fashion. Additionally we apply the tool to six cylinder-forming polystyrene-block-poly(2-vinylpyridine) polymers thermally annealed to explore the relationship between the size of each polymer and measured characteristics including line period, line-width, defect density, line-edge roughness (LER), line-width roughness (LWR), and correlation length. Finally, we explore the line-edge roughness, line-width roughness, defect density, and correlation length as a function of the image area sampled to determine each in a more rigorous fashion. PMID:26207990

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

Farmer, M. T.

The overall objective of the current work is to carry out a scoping analysis to determine the impact of ATF on late phase accident progression; in particular, the molten core-concrete interaction portion of the sequence that occurs after the core debris fails the reactor vessel and relocates into containment. This additional study augments previous work by including kinetic effects that govern chemical reaction rates during core-concrete interaction. The specific ATF considered as part of this study is SiC-clad UO 2.

Electrophoretic extraction of proteins from two-dimensional electrophoresis gel spots

DOEpatents

Zhang, Jian-Shi; Giometti, C.S.; Tollaksen, S.L.

1987-09-04

After two-dimensional electrophoresis of proteins or the like, resulting in a polyacrylamide gel slab having a pattern of protein gel spots thereon, an individual protein gel spot is cored out from the slab, to form a gel spot core which is placed in an extraction tube, with a dialysis membrane across the lower end of the tube. Replicate gel spots can be cored out from replicate gel slabs and placed in the extraction tube. Molten agarose gel is poured into the extraction tube where the agarose gel hardens to form an immobilizing gel, covering the gel spot cores. The upper end portion of the extraction tube is filled with a volume of buffer solution, and the upper end is closed by another dialysis membrane. Upper and lower bodies of a buffer solution are brought into contact with the upper and lower membranes and are provided with electrodes connected to the positive and negative terminals of a dc power supply, thereby producing an electrical current which flows through the upper membrane, the volume of buffer solution, the agarose, the gel spot cores and the lower membrane. The current causes the proteins to be extracted electrophoretically from the gel spot cores, so that the extracted proteins accumulate and are contained in the space between the agarose gel and the upper membrane. 8 figs.

Fossil diatom assemblages as paleoecological indicators of paleo-water environmental change in the Ulleung Basin, East Sea, Republic of Korea

NASA Astrophysics Data System (ADS)

Yun, Suk Min; Lee, Taehee; Jung, Seung Won; Park, Joon Sang; Lee, Jin Hwan

2017-09-01

The fossil diatom assemblage record from two sediment cores obtained from the Ulleung Basin, East Sea, Republic of Korea, revealed changes in the diatom assemblage zones in PG1 and PD3 core samples. The two sediment cores were δC14 dated and approximately represented the late Pleistocene-Holocene. The analysis of age zones in the PG1 core and PD3 core was assessed based on the frequency of variations, and occurrences of biostratigraphical fossil diatom species. During the Last Glacial Maximum (LGM), the sea level was lower than that at present and the Ulleung Basin became isolated from the Pacific Ocean. As a result, there would have been a limited Tsushima Warm Current (TWC) influence, and salinity would have decreased resulting in increased freshwater and coastal diatoms. The distribution pattern of diatoms presented in the cores was associated with changes in water temperature and salinity and the adding of terrigenous material brought about by the input of freshwater. Changes in the abundance of a tychopelagic diatom, Paralia sulcata, reflected the effect of the water currents. Diatom temperature (Td) values and the ratio of centric/pennate diatoms provided evidence of limited influences of the TWC and freshwater inflow. It is thought that all assemblage zones were influenced by the TWC, which had an important effect on the distribution and composition of fossil diatoms.

Broadband absorption and enhanced photothermal conversion property of octopod-like Ag@Ag2S core@shell structures with gradually varying shell thickness.

PubMed

Jiang, Qian; Zeng, Wenxia; Zhang, Canying; Meng, Zhaoguo; Wu, Jiawei; Zhu, Qunzhi; Wu, Daxiong; Zhu, Haitao

2017-12-19

Photothermal conversion materials have promising applications in many fields and therefore they have attracted tremendous attention. However, the multi-functionalization of a single nanostructure to meet the requirements of multiple photothermal applications is still a challenge. The difficulty is that most nanostructures have specific absoprtion band and are not flexible to different demands. In the current work, we reported the synthesis and multi-band photothermal conversion of Ag@Ag 2 S core@shell structures with gradually varying shell thickness. We synthesized the core@shell structures through the sulfidation of Ag nanocubes by taking the advantage of their spatially different reactivity. The resulting core@shell structures show an octopod-like mopgorlogy with a Ag 2 S bulge sitting at each corner of the Ag nanocubes. The thickness of the Ag 2 S shell gradually increases from the central surface towards the corners of the structure. The synthesized core@shell structures show a broad band absorption spectrum from 300 to 1100 nm. Enhanced photothermal conversion effect is observed under the illuminations of 635, 808, and 1064 nm lasers. The results indicate that the octopod-like Ag@Ag 2 S core@shell structures have characteristics of multi-band photothermal conversion. The current work might provide a guidance for the design and synthesis of multifunctional photothermal conversion materials.

Dynamical Core in Atmospheric Model Does Matter in the Simulation of Arctic Climate

NASA Astrophysics Data System (ADS)

Jun, Sang-Yoon; Choi, Suk-Jin; Kim, Baek-Min

2018-03-01

Climate models using different dynamical cores can simulate significantly different winter Arctic climates even if equipped with virtually the same physics schemes. Current climate simulated by the global climate model using cubed-sphere grid with spectral element method (SE core) exhibited significantly warmer Arctic surface air temperature compared to that using latitude-longitude grid with finite volume method core. Compared to the finite volume method core, SE core simulated additional adiabatic warming in the Arctic lower atmosphere, and this was consistent with the eddy-forced secondary circulation. Downward longwave radiation further enhanced Arctic near-surface warming with a higher surface air temperature of about 1.9 K. Furthermore, in the atmospheric response to the reduced sea ice conditions with the same physical settings, only the SE core showed a robust cooling response over North America. We emphasize that special attention is needed in selecting the dynamical core of climate models in the simulation of the Arctic climate and associated teleconnection patterns.

Circular current loops, magnetic dipoles and spherical harmonic analysis.

USGS Publications Warehouse

Alldredge, L.R.

1980-01-01

Spherical harmonic analysis (SHA) is the most used method of describing the Earth's magnetic field, even though spherical harmonic coefficients (SHC) almost completely defy interpretation in terms of real sources. Some moderately successful efforts have been made to represent the field in terms of dipoles placed in the core in an effort to have the model come closer to representing real sources. Dipole sources are only a first approximation to the real sources which are thought to be a very complicated network of electrical currents in the core of the Earth. -Author

Wire inhomogeneity detector having a core with opposing pole pieces and guide pieces adjacent the opposing pole pieces

DOEpatents

Gibson, George H.; Smits, Robert G.; Eberhard, Philippe H.

1989-01-01

A device for uncovering imperfections in electrical conducting wire, particularly superconducting wire, by detecting variations in eddy currents. Eddy currents effect the magnetic field in a gap of an inductor, contained in a modified commercial ferrite core, through which the wire being tested is passed. A small increase or decrease in the amount of conductive material, such as copper, in a fixed cross section of wire will unbalance a bridge used to measure the impedance of the inductor, tripping a detector and sounding an alarm.

Genome-wide computational prediction and analysis of core promoter elements across plant monocots and dicots

USDA-ARS?s Scientific Manuscript database

Transcription initiation, essential to gene expression regulation, involves recruitment of basal transcription factors to the core promoter elements (CPEs). The distribution of currently known CPEs across plant genomes is largely unknown. This is the first large scale genome-wide report on the compu...

Qualifications and Assignments of Alternatively Certified Teachers: Testing Core Assumptions

ERIC Educational Resources Information Center

Cohen-Vogel, Lora; Smith, Thomas M.

2007-01-01

By analyzing data from the Schools and Staffing Survey, the authors empirically test four of the core assumptions embedded in current arguments for expanding alternative teacher certification (AC): AC attracts experienced candidates from fields outside of education; AC attracts top-quality, well-trained teachers; AC disproportionately trains…

Identification of a Core Curriculum in Gerontology for Allied Health Professionals. Final Report.

ERIC Educational Resources Information Center

Hedl, John J.; And Others

The overall goal of this project was to identify a core curriculum in gerontology for seven allied health professions (radiologic technologist, radiation therapist, respiratory therapist, dental hygienist, dental assistant, physical therapy assistant, and occupational therapy assistant). The project also identified the current state of gerontology…

Building a Case for the Core Curriculum in Agriculture.

ERIC Educational Resources Information Center

Hemp, Paul E.

1980-01-01

Changes in the types of students enrolled in vocational agriculture and their interests, background, and needs suggest that agricultural educators should rethink the approaches currently used in curriculum development. The advantages of the core curriculum and the traditional approach to curriculum development need to be compared and weighed…

Design and evaluation of 66 kV-class HTS power cable using REBCO wires

NASA Astrophysics Data System (ADS)

Ohya, M.; Yumura, H.; Masuda, T.; Amemiya, N.; Ishiyama, A.; Ohkuma, T.

2011-11-01

Sumitomo Electric (SEI) has been involved in the development of 66 kV-class HTS cables using REBCO wires. One of the technical targets in this project is to reduce the AC loss to less than 2 W/m/phase at 5 kA. SEI has developed a clad-type of textured metal substrate with lower magnetization loss compared with a conventional NiW substrate. In addition, 30 mm-wide REBCO tapes were slit into 4 mm-wide strips, and these strips were wound spirally on a former with small gaps. The AC loss of a manufactured 4-layer cable conductor was 1.5 W/m at 5 kA at 64 K. Given that the AC loss in a shield layer is supposed to be one-fourth of a whole cable core loss, our cables are expected to achieve the AC loss target of less than 2 W/m/phase at 5 kA. Another important target is to manage a fault current. A cable core was designed and fabricated based on the simulation findings, and over-current tests (max. 31.5 kA, 2 s) were conducted to check its performance. The critical current value of the cable cores were measured before and after the over-current tests and verified its soundness. A 5 kA-class current lead for the cable terminations was also developed. The current loading tests were conducted for the developed current leads. The temperature distribution of the current leads reached to the steady-state within less than 12 h, and it was confirmed that the developed current lead has enough capacity of 5 kA loading.