Tritium handling experience at Atomic Energy of Canada Limited

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

Suppiah, S.; McCrimmon, K.; Lalonde, S.

2015-03-15

Canada has been a leader in tritium handling technologies as a result of the successful CANDU reactor technology used for power production. Over the last 50 to 60 years, capabilities have been established in tritium handling and tritium management in CANDU stations, tritium removal processes for heavy and light water, tritium measurement and monitoring, and understanding the effects of tritium on the environment. This paper outlines details of tritium-related work currently being carried out at Atomic Energy of Canada Limited (AECL). It concerns the CECE (Combined Electrolysis and Catalytic Exchange) process for detritiation, tritium-compatible electrolysers, tritium permeation studies, and tritiummore » powered batteries. It is worth noting that AECL offers a Tritium Safe-Handling Course to national and international participants, the course is a mixture of classroom sessions and hands-on practical exercises. The expertise and facilities available at AECL is ready to address technological needs of nuclear fusion and next-generation nuclear fission reactors related to tritium handling and related issues.« less

CANDU in-reactor quantitative visual-based inspection techniques

NASA Astrophysics Data System (ADS)

Rochefort, P. A.

2009-02-01

This paper describes two separate visual-based inspection procedures used at CANDU nuclear power generating stations. The techniques are quantitative in nature and are delivered and operated in highly radioactive environments with access that is restrictive, and in one case is submerged. Visual-based inspections at stations are typically qualitative in nature. For example a video system will be used to search for a missing component, inspect for a broken fixture, or locate areas of excessive corrosion in a pipe. In contrast, the methods described here are used to measure characteristic component dimensions that in one case ensure ongoing safe operation of the reactor and in the other support reactor refurbishment. CANDU reactors are Pressurized Heavy Water Reactors (PHWR). The reactor vessel is a horizontal cylindrical low-pressure calandria tank approximately 6 m in diameter and length, containing heavy water as a neutron moderator. Inside the calandria, 380 horizontal fuel channels (FC) are supported at each end by integral end-shields. Each FC holds 12 fuel bundles. The heavy water primary heat transport water flows through the FC pressure tube, removing the heat from the fuel bundles and delivering it to the steam generator. The general design of the reactor governs both the type of measurements that are required and the methods to perform the measurements. The first inspection procedure is a method to remotely measure the gap between FC and other in-core horizontal components. The technique involves delivering vertically a module with a high-radiation-resistant camera and lighting into the core of a shutdown but fuelled reactor. The measurement is done using a line-of-sight technique between the components. Compensation for image perspective and viewing elevation to the measurement is required. The second inspection procedure measures flaws within the reactor's end shield FC calandria tube rolled joint area. The FC calandria tube (the outer shell of the FC) is sealed by rolling its ends into the rolled joint area. During reactor refurbishment, the original FC calandria tubes are removed, potentially scratching the rolled joint area and, thereby, compromising the seal with the new FC calandria tube. The procedure involves delivering an inspection module having a radiation-resistant camera, standard lighting, and a structured lighting projector. The surface is inspected by rotating the module within the rolled joint area. If a flaw is detected, its depth and width are gauged from the profile variation of the structured lighting in a captured image. As well, the diameter profile of the area is measured from the analysis of a series of captured circumferential images of the structured lighting profiles on the surface.

Renewability and sustainability aspects of nuclear energy

NASA Astrophysics Data System (ADS)

Şahin, Sümer

2014-09-01

Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, 233U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO2/RG-PuO2) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG-PuO2 + 96 % ThO2; 6 % RG-PuO2 + 94 % ThO2; 10 % RG-PuO2 + 90 % ThO2; 20 % RG-PuO2 + 80 % ThO2; 30 % RG-PuO2 + 70 % ThO2, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ˜ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ˜ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG-PuO2 fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MWth has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ˜160 kg 233U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ˜1.3.

Intergranular fracture in irradiated Inconel X-750 containing very high concentrations of helium and hydrogen

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

Judge, Colin D.; Gauquelin, Nicolas; Walters, Lori

2015-02-01

In recent years, it has been determined that Inconel X-750 CANDU spacers have lost strength and material ductility following irradiation in reactor. The irradiated fracture behaviour of ex-service material was also found to be entirely intergranular. The heavily thermalized flux spectrum in a CANDU reactor results in transmutation of 58Ni to 59Ni. The 59Ni itself has unusually high thermal neutron reaction cross-sections of the type: (n, γ), (n, p), and (n,α). The latter two reactions, in particular, contribute to a significant enhancement of the atomic displacements in addition to creating high concentrations of hydrogen and helium within the material. Metallographicmore » examinations by transmission electron microscopy (TEM) have confirmed the presence of helium bubbles in the matrix and aligned along grain boundaries and matrix-precipitate interfaces. He bubble size and density are found to be highly dependent on the irradiation temperature and material microstructure; the bubbles are larger within grain boundary precipitates. TEM specimens extracted from fracture surfaces and crack tips give direct evidence linking crack propagation with grain boundary He bubbles.« less

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

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

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

Skutnik, Steven E.

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

Renewability and sustainability aspects of nuclear energy

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

Şahin, Sümer, E-mail: ssahin@atilim.edit.tr

Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, {sup 233}U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO{sub 2}/RG‐PuO{sub 2}) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG‐PuO{sub 2} + 96 % ThO{sub 2}; 6 % RG‐PuO{sub 2} + 94 % ThO{sub 2};more » 10 % RG‐PuO{sub 2} + 90 % ThO{sub 2}; 20 % RG‐PuO{sub 2} + 80 % ThO{sub 2}; 30 % RG‐PuO{sub 2} + 70 % ThO{sub 2}, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ∼ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG‐PuO{sub 2} fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MW{sub th} has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ∼160 kg {sup 233}U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ∼1.3.« less

Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

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

Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.

2013-07-01

REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

Development of a Korean reference HLW disposal system under the Korean representative geologic conditions

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

Choi, Heui-Joo; Lee, Jong Youl; Choi, Jongwon

2007-07-01

The development of a Korean Reference disposal System for the spent fuels from PWR and CANDU reactors is outlined in this paper. Around 36,000 tU of spent fuels are being projected based on the lifetimes of 28 nuclear power reactors in Korea. Since the site for the geological disposal has not yet been decided, a hypothetical site with representative Korean geologic conditions is proposed for the conceptual design of the repository. The disposal rates of the spent fuels are determined according to the total operation time of 55 years. The canisters are optimized by considering natural Korean conditions, and themore » buffer is designed with domestic Ca-bentonite. The depth of the repository is determined to be 500 m below the ground's surface. The canister separation distances are determined through a thermal analysis. The main features of the repository are presented from the layout to the closure. A computer program has been developed to calculate and analyze the volume and the area of the disposal system to help in the cost analysis. The final output of the design is presented as a unit disposal cost, US $315 /kgU. (authors)« less

Feasible variants for intermediate storage of the spent fuel obtained at NPP Cernavoda, Romania

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

Radu, M.; Popescu, G.

1993-12-31

The 5 CANDU-PHW Reactors of 600 Standard type of Cernavoda Nuclear Power Plant are under construction and the first unit is expected to be commissioned in 1995, group 2 following after 2 years, and then groups 3, 4 and 5 one each year. In this study there are presented feasible variants for intermediate storage of spent fuel, obtained during 30 years of operation from the stations at Cernavoda. From the solutions applied worldwide, both dry and wet storage have been taken into account. In any of the two variants, a unique intermediate storage will be provided and the storage buildingmore » was proposed to be built in 4 different stages. As a first estimation, considering the fact that, by now Romania has only one nuclear plant of CANDU fuel type the dry variant seems to be the best.« less

Conceptual design of a thermalhydraulic loop for multiple test geometries at supercritical conditions named Supercritical Phenomena Experimental Test Apparatus (SPETA)

NASA Astrophysics Data System (ADS)

Adenariwo, Adepoju

The efficiency of nuclear reactors can be improved by increasing the operating pressure of current nuclear reactors. Current CANDU-type nuclear reactors use heavy water as coolant at an outlet pressure of up to 11.5 MPa. Conceptual SuperCritical Water Reactors (SCWRs) will operate at a higher coolant outlet pressure of 25 MPa. Supercritical water technology has been used in advanced coal plants and its application proves promising to be employed in nuclear reactors. To better understand how supercritical water technology can be applied in nuclear power plants, supercritical water loops are used to study the heat transfer phenomena as it applies to CANDU-type reactors. A conceptual design of a loop known as the Supercritical Phenomena Experimental Apparatus (SPETA) has been done. This loop has been designed to fit in a 9 m by 2 m by 2.8 m enclosure that will be installed at the University of Ontario Institute of Technology Energy Research Laboratory. The loop include components to safely start up and shut down various test sections, produce a heat source to the test section, and to remove reject heat. It is expected that loop will be able to investigate the behaviour of supercritical water in various geometries including bare tubes, annulus tubes, and multi-element-type bundles. The experimental geometries are designed to match the fluid properties of Canadian SCWR fuel channel designs so that they are representative of a practical application of supercritical water technology in nuclear plants. This loop will investigate various test section orientations which are the horizontal, vertical, and inclined to investigate buoyancy effects. Frictional pressure drop effects and satisfactory methods of estimating hydraulic resistances in supercritical fluid shall also be estimated with the loop. Operating limits for SPETA have been established to be able to capture the important heat transfer phenomena at supercritical conditions. Heat balance and flow calculations have been done to appropriately size components in the loop. Sensitivity analysis has been done to find the optimum design for the loop.

View of Pakistan Atomic Energy Commission towards SMPR's in the light of KANUPP performance

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

Huseini, S.D.

1985-01-01

The developing countries in general do not have grid capacities adequate enough to incorporate standard size, economic but rather large nuclear power plants for maximum advantage. Therefore, small and medium size reactors (SMPR) have been and still are, of particular interest to the developing countries in spite of certain known problems with these reactors. Pakistan Atomic Energy Commission (PAEC) has been operating a CANDU type of a small PHWR plant since 1971 when it was connected to the local Karachi grid. This paper describes PAEC's view in the light of KANUPP performance with respect to such factors associated with SMPR'smore » as selection of suitable reactor size and type, its operation in a grid of small capacity, flexibility of operation and its role as a reliable source of electrical power.« less

Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

Accelerator-driven transmutation of spent fuel elements

DOEpatents

Venneri, Francesco; Williamson, Mark A.; Li, Ning

2002-01-01

An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

Computed tomography of radioactive objects and materials

NASA Astrophysics Data System (ADS)

Sawicka, B. D.; Murphy, R. V.; Tosello, G.; Reynolds, P. W.; Romaniszyn, T.

1990-12-01

Computed tomography (CT) has been performed on a number of radioactive objects and materials. Several unique technical problems are associated with CT of radioactive specimens. These include general safety considerations, techniques to reduce background-radiation effects on CT images and selection criteria for the CT source to permit object penetration and to reveal accurate values of material density. In the present paper, three groups of experiments will be described, for objects with low, medium and high levels of radioactivity. CT studies on radioactive specimens will be presented. They include the following: (1) examination of individual ceramic reactor-fuel (uranium dioxide) pellets, (2) examination of fuel samples from the Three Mile Island reactor, (3) examination of a CANDU (CANada Deuterium Uraniun: registered trademark) nuclear-fuel bundle which underwent a simulated loss-of-coolant accident resulting in high-temperature damage and (4) examination of a PWR nuclear-reactor fuel assembly.

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

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

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

Nuclear Data Needs for Generation IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

Rullhusen, Peter

2006-04-01

Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S. Kozier, G. R. Dyck. The lead cooled fast reactor benchmark BREST-300: analysis with sensitivity method / V. Smirnov ... [et al.]. Sensitivity analysis of neutron cross-sections considered for design and safety studies of LFR and SFR generation IV systems / K. Tucek, J. Carlsson, H. Wider -- Experiments. INL capabilities for nuclear data measurements using the Argonne intense pulsed neutron source facility / J. D. Cole ... [et al.]. Cross-section measurements in the fast neutron energy range / A. Plompen. Recent measurements of neutron capture cross sections for minor actinides by a JNC and Kyoto University Group / H. Harada ... [et al.]. Determination of minor actinides fission cross sections by means of transfer reactions / M. Aiche ... [et al.] -- Evaluated data libraries. Nuclear data services from the NEA / H. Henriksson, Y. Rugama. Nuclear databases for energy applications: an IAEA perspective / R. Capote Noy, A. L. Nichols, A. Trkov. Nuclear data evaluation for generation IV / G. Noguère ... [et al.]. Improved evaluations of neutron-induced reactions on americium isotopes / P. Talou ... [et al.]. Using improved ENDF-based nuclear data for candu reactor calculations / J. Prodea. A comparative study on the graphite-moderated reactors using different evaluated nuclear data / Do Heon Kim ... [et al.].

Study of medical isotope production facility stack emissions and noble gas isotopic signature using automatic gamma-spectra analysis platform

NASA Astrophysics Data System (ADS)

Zhang, Weihua; Hoffmann, Emmy; Ungar, Kurt; Dolinar, George; Miley, Harry; Mekarski, Pawel; Schrom, Brian; Hoffman, Ian; Lawrie, Ryan; Loosz, Tom

2013-04-01

The nuclear industry emissions of the four CTBT (Comprehensive Nuclear-Test-Ban Treaty) relevant radioxenon isotopes are unavoidably detected by the IMS along with possible treaty violations. Another civil source of radioxenon emissions which contributes to the global background is radiopharmaceutical production companies. To better understand the source terms of these background emissions, a joint project between HC, ANSTO, PNNL and CRL was formed to install real-time detection systems to support 135Xe, 133Xe, 131mXe and 133mXe measurements at the ANSTO and CRL 99Mo production facility stacks as well as the CANDU (CANada Deuterium Uranium) primary coolant monitoring system at CRL. At each site, high resolution gamma spectra were collected every 15 minutes using a HPGe detector to continuously monitor a bypass feed from the stack or CANDU primary coolant system as it passed through a sampling cell. HC also conducted atmospheric monitoring for radioxenon at approximately 200 km distant from CRL. A program was written to transfer each spectrum into a text file format suitable for the automatic gamma-spectra analysis platform and then email the file to a server. Once the email was received by the server, it was automatically analysed with the gamma-spectrum software UniSampo/Shaman to perform radionuclide identification and activity calculation for a large number of gamma-spectra in a short period of time (less than 10 seconds per spectrum). The results of nuclide activity together with other spectrum parameters were saved into the Linssi database. This database contains a large amount of radionuclide information which is a valuable resource for the analysis of radionuclide distribution within the noble gas fission product emissions. The results could be useful to identify the specific mechanisms of the activity release. The isotopic signatures of the various radioxenon species can be determined as a function of release time. Comparison of 133mXe and 133Xe activity ratios showed distinct differences between the closed CANDU primary coolant system and radiopharmaceutical production releases. According to the concept proposed by Kalinowski and Pistner (2006), the relationship between different isotopic activity ratios based on three or four radioxenon isotopes was plotted in a log-log diagram for source characterisation (civil vs. nuclear test). The multiple isotopic activity ratios were distributed in three distinct areas: HC atmospheric monitoring ratios extended to far left; the CANDU primary coolant system ratios lay in the middle; and 99Mo stack monitoring ratios for ANSTO and CRL were located on the right. The closed CANDU primary coolant has the lowest logarithmic mean ratio that represents the nuclear power reactor operation. The HC atmospheric monitoring exhibited a broad range of ratios spreading over several orders of magnitude. In contrast, the ANSTO and CRL stack emissions showed the smallest range of ratios but the results indicate at least two processes involved in the 99Mo productions. Overall, most measurements were found to be shifted towards the reactor domain. The hypothesis is that this is due to an accumulation of the isotope 131mXe in the stack or atmospheric background as it has the longest half-life and extra 131mXe emissions from the decay of 131I. The contribution of older 131mXe to a fresh release shifts the ratio of 133mXe/131mXe to the left. It was also very interesting to note that there were some situations where isotopic ratios from 99Mo production emissions fell within the nuclear test domain. This is due to operational variability, such as shorter target irradiation times. Martin B. Kalinowski and Christoph Pistner, (2006), Isotopic signature of atmospheric xenon released from light water reactors, Journal of Environmental Radioactivity, 88, 215-235.

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

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

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

Forsberg, C.W.; Reich, W.J.

1991-09-01

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactormore » concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.« less

Emerging nuclear programs in Asia: The Phillipines, Thailand, Indonesia, and Pakistan

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

Williams, M.L.

This article is a review of the potential for nuclear energy development in the developing nations of Pakistan, Indonesia, Thailand, and the Philippines. In each country, there is a substantial need for new generating capacity, and each is exploring the idea of having nuclear energy supply a meaningful portion of this new capacity. Of the four countries, only Pakistan is currently a nuclear operator, and one vintage CANDU plant in operation and the Chashma unit under construction. Thailand and Indonesia have ambitious plans to have 12 reactors in service by the year 2015.

Enrichment of 57Fe isotope in neutron flux of nuclear reactors observed by Mössbauer spectroscopy.

PubMed

Sawicki, Jerzy A

2018-02-01

The abundance of 57 Fe isotope in nuclear reactor core materials can be considerably enriched by neutron-capture 56 Fe(n,γ) reactions. This is demonstrated using the sections of Zr-2.5 wt.%Nb pressure tubes removed from two CANDU* reactors. The tubes contained 0.11 and 0.04wt% Fe and were irradiated for about 10 effective full power years (EFPY) up to ~10 26 n/m 2 fast neutron (E > 1MeV) fluencies. The Mössbauer spectra of 57 Fe in irradiated samples indicated up to 10 times larger areas than unirradiated off-cuts from the same pressure tubes. The observed effect is in accord with the values calculated for known thermal neutron-capture cross-sections and resonance capture integrals, neutron flux profiles and spectra, and times of irradiation. The build-up of 57 Fe facilitated recording Mössbauer absorption spectra of alloys with minor amount of Fe down to ~ 400ppm, despite intense background radiation emitted by samples. These findings can open new possibilities in post-irradiation studies of alloys used in nuclear reactors and in other objects subjected to large neutron fluencies. Copyright © 2017 Elsevier Ltd. All rights reserved.

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

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

Rohatgi, Upendra S.

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary ofmore » appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/« less

Integrity assessment of grouted posttensioning cables and reinforced concrete of a nuclear containment building

NASA Astrophysics Data System (ADS)

Philipose, K.; Shenton, B.

2011-04-01

The Containment Buildings of CANDU Nuclear Generating Stations were designed to house nuclear reactors and process equipment and also to provide confinement of releases from a potential nuclear accident such as a Loss Of Coolant Accident (LOCA). To meet this design requirement, a post-tensioning system was designed to induce compressive stresses in the structure to counteract the internal design pressure. The CANDU reactor building at Gentilly-1 (G-1), Quebec, Canada (250 MWe) was built in the early 1970s and is currently in a decommissioned state. The structure at present is under surveillance and monitoring. In the year 2000, a field investigation was conducted as part of a condition assessment and corrosion was detected in some of the grouted post-tension cable strands. However, no further work was done at that time to determine the cause, nature, impact and extent of the corrosion. An investigation of the Gentilly-1 containment building is currently underway to assess the condition of grouted post-tensioning cables and reinforced concrete. At two selected locations, concrete and steel reinforcements were removed from the containment building wall to expose horizontal cables. Individual cable strands and reinforcement bars were instrumented and measurements were taken in-situ before removing them for forensic examination and destructive testing to determine the impact of ageing and corrosion. Concrete samples were also removed and tested in a laboratory. The purpose of the field investigation and laboratory testing, using this structure as a test bed, was also to collect material ageing data and to develop potential Nondestructive Examination (NDE) methods to monitor Containment Building Integrity. The paper describes the field work conducted and the test results obtained for concrete, reinforcement and post-tensioning cables.

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

Design of Test Loops for Forced Convection Heat Transfer Studies at Supercritical State

NASA Astrophysics Data System (ADS)

Balouch, Masih N.

Worldwide research is being conducted to improve the efficiency of nuclear power plants by using supercritical water (SCW) as the working fluid. One such SCW reactor considered for future development is the CANDU-Supercritical Water Reactor (CANDU-SCWR). For safe and accurate design of the CANDU-SCWR, a detailed knowledge of forced-convection heat transfer in SCW is required. For this purpose, two supercritical fluid loops, i.e. a SCW loop and an R-134a loop are developed at Carleton University. The SCW loop is designed to operate at pressures as high as 28 MPa, temperatures up to 600 °C and mass fluxes of up to 3000 kg/m2s. The R-134a loop is designed to operate at pressures as high as 6 MPa, temperatures up to 140 °C and mass fluxes in the range of 500-6000 kg/m2s. The test loops designs allow for up to 300 kW of heating power to be imparted to the fluid. Both test loops are of the closed-loop design, where flow circulation is achieved by a centrifugal pump in the SCW loop and three parallel-connected gear pumps in the R-134a loop, respectively. The test loops are pressurized using a high-pressure nitrogen cylinder and accumulator assembly, which allows independent control of the pressure, while simultaneously dampening pump induced pressure fluctuations. Heat exchangers located upstream of the pumps control the fluid temperature in the test loops. Strategically located measuring instrumentation provides information on the flow rate, pressure and temperature in the test loops. The test loops have been designed to accommodate a variety of test-section geometries, ranging from a straight circular tube to a seven-rod bundle, achieving heat fluxes up to 2.5 MW/m2 depending on the test-section geometry. The design of both test loops allows for easy reconfiguration of the test-section orientation relative to the gravitational direction. All the test sections are of the directly-heated design, where electric current passing through the pressure retaining walls of the test sections provides the Joule heating required to heat up the fluid to supercritical conditions. A high-temperature dielectric gasket isolates the current carrying parts of the test section from the rest of the assembly. Temperature and pressure drop data are collected at the inlet and outlet, and along the heated length of the test section. The test loops and test sections are designed according to American Society of Mechanical Engineers (ASME) Pressure Piping B31.1, and Boiler and Pressure Vessel Code, Section VIII-Division 1 rules. The final test loops and test sections assemblies are certified by Technical Standards and Safety Authority (TSSA). Every attempt is made to use off-the-shelf components where possible in order to streamline the design process and reduce costs. Following a rigorous selection process, stainless steel Types 316 and 316H are selected as the construction materials for the test loops, and Inconel 625 is selected as the construction material for the test sections. This thesis describes the design of the SCW and R-134a loops along with the three test-section geometries (i.e., tubular, annular and bundle designs).

Improvement of Pt/C/PTFE catalyst type used for hydrogen isotope separation

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

Vasut, F.; Preda, A.; Zamfirache, M.

2008-07-15

The CANDU reactor from the Nuclear Power plant Cernavoda (Romania)) is the most powerful tritium source from Europe. This reactor is moderated and cooled by heavy water that becomes continuously contaminated with tritium. Because of this reason, the National R and amp;D Inst. for Cryogenic and Isotopic Technologies developed a detritiation technology based on catalytic isotopic exchange and cryogenic distillation. The main effort of our Inst. was focused on finding more efficient catalysts with a longer operational life. Some of the tritium removal processes involved in Fusion Science and Technology use this type of catalyst 1. Several Pt/C/PTFE hydrophobic catalystsmore » that could be used in isotopic exchange process 2,3,4 were produced. The present paper presents a comparative study between the physical and morphological properties of different catalysts manufactured by impregnation at our institute. The comparison consists of a survey of specific surface, pores volume and pores distribution. (authors)« less

The environmental constraint needs for design improvements to the Saligny I/LLW-repository near Cernavoda NPP

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

Barariu, Gheorghe

2007-07-01

The paper presents the new perspectives on the development of the L/ILW Final Repository Project which will be built near Cernavoda NPP. The Repository is designed to satisfy the main performance objectives in accordance to IAEA recommendation. Starting in October 1996, Romania became a country with an operating nuclear power plant. Reactor 2 reached the criticality on May 6, 2007 and it will be put in commercial operation in September 2007. The Ministry of Economy and Finance has decided to proceed with the commissioning of Units 3 and 4 of Cernavoda NPP till 2014. The Strategy for radioactive waste managementmore » was elaborated by National Agency for Radioactive Waste (ANDRAD), the jurisdictional authority for definitive disposal and the coordination of nuclear spent fuel and radioactive waste management (Order 844/2004) with attributions established by Governmental Decision (GO) 31/2006. The Strategy specifies the commissioning of the Saligny L/IL Radwaste Repository near Cernavoda NPP in 2014. When designing the L/IL Radwaste Repository, the following prerequisites have been taken into account: 1) Cernavoda NPP will be equipped with 4 Candu 6 units. 2) National Legislation in radwaste management will be reviewed and/or completed to harmonize with UE standards 3) The selected site is now in process of confirmation after a comprehensive set of interdisciplinary investigations. (author)« less

A statistical approach to nuclear fuel design and performance

NASA Astrophysics Data System (ADS)

Cunning, Travis Andrew

As CANDU fuel failures can have significant economic and operational consequences on the Canadian nuclear power industry, it is essential that factors impacting fuel performance are adequately understood. Current industrial practice relies on deterministic safety analysis and the highly conservative "limit of operating envelope" approach, where all parameters are assumed to be at their limits simultaneously. This results in a conservative prediction of event consequences with little consideration given to the high quality and precision of current manufacturing processes. This study employs a novel approach to the prediction of CANDU fuel reliability. Probability distributions are fitted to actual fuel manufacturing datasets provided by Cameco Fuel Manufacturing, Inc. They are used to form input for two industry-standard fuel performance codes: ELESTRES for the steady-state case and ELOCA for the transient case---a hypothesized 80% reactor outlet header break loss of coolant accident. Using a Monte Carlo technique for input generation, 105 independent trials are conducted and probability distributions are fitted to key model output quantities. Comparing model output against recognized industrial acceptance criteria, no fuel failures are predicted for either case. Output distributions are well removed from failure limit values, implying that margin exists in current fuel manufacturing and design. To validate the results and attempt to reduce the simulation burden of the methodology, two dimensional reduction methods are assessed. Using just 36 trials, both methods are able to produce output distributions that agree strongly with those obtained via the brute-force Monte Carlo method, often to a relative discrepancy of less than 0.3% when predicting the first statistical moment, and a relative discrepancy of less than 5% when predicting the second statistical moment. In terms of global sensitivity, pellet density proves to have the greatest impact on fuel performance, with an average sensitivity index of 48.93% on key output quantities. Pellet grain size and dish depth are also significant contributors, at 31.53% and 13.46%, respectively. A traditional limit of operating envelope case is also evaluated. This case produces output values that exceed the maximum values observed during the 105 Monte Carlo trials for all output quantities of interest. In many cases the difference between the predictions of the two methods is very prominent, and the highly conservative nature of the deterministic approach is demonstrated. A reliability analysis of CANDU fuel manufacturing parametric data, specifically pertaining to the quantification of fuel performance margins, has not been conducted previously. Key Words: CANDU, nuclear fuel, Cameco, fuel manufacturing, fuel modelling, fuel performance, fuel reliability, ELESTRES, ELOCA, dimensional reduction methods, global sensitivity analysis, deterministic safety analysis, probabilistic safety analysis.

Quasi-heterogeneous efficient 3-D discrete ordinates CANDU calculations using Attila

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

Preeti, T.; Rulko, R.

2012-07-01

In this paper, 3-D quasi-heterogeneous large scale parallel Attila calculations of a generic CANDU test problem consisting of 42 complete fuel channels and a perpendicular to fuel reactivity device are presented. The solution method is that of discrete ordinates SN and the computational model is quasi-heterogeneous, i.e. fuel bundle is partially homogenized into five homogeneous rings consistently with the DRAGON code model used by the industry for the incremental cross-section generation. In calculations, the HELIOS-generated 45 macroscopic cross-sections library was used. This approach to CANDU calculations has the following advantages: 1) it allows detailed bundle (and eventually channel) power calculationsmore » for each fuel ring in a bundle, 2) it allows the exact reactivity device representation for its precise reactivity worth calculation, and 3) it eliminates the need for incremental cross-sections. Our results are compared to the reference Monte Carlo MCNP solution. In addition, the Attila SN method performance in CANDU calculations characterized by significant up scattering is discussed. (authors)« less

Current Status of The Romanian National Deep Geological Repository Program

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

Radu, M.; Nicolae, R.; Nicolae, D.

2008-07-01

Construction of a deep geological repository is a very demanding and costly task. By now, countries that have Candu reactors, have not processed the spent fuel passing to the interim storage as a preliminary step of final disposal within the nuclear fuel cycle back-end. Romania, in comparison to other nations, represents a rather small territory, with high population density, wherein the geological formation areas with radioactive waste storage potential are limited and restricted not only from the point of view of the selection criteria due to the rocks natural characteristics, but also from the point of view of their involvementmore » in social and economical activities. In the framework of the national R and D Programs, series of 'Map investigations' have been made regarding the selection and preliminary characterization of the host geological formation for the nation's spent fuel deep geological repository. The fact that Romania has many deposits of natural gas, oil, ore and geothermal water, and intensively utilizes soil and also is very forested, cause some of the apparent acceptable sites to be rejected in the subsequent analysis. Currently, according to the Law on the spent fuel and radioactive waste management, including disposal, The National Agency of Radioactive Waste is responsible and coordinates the national strategy in the field and, subsequently, further actions will be decided. The Romanian National Strategy, approved in 2004, projects the operation of a deep geological repository to begin in 2055. (authors)« less

Methodology, status and plans for development and assessment of TUF and CATHENA codes

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

Luxat, J.C.; Liu, W.S.; Leung, R.K.

1997-07-01

An overview is presented of the Canadian two-fluid computer codes TUF and CATHENA with specific focus on the constraints imposed during development of these codes and the areas of application for which they are intended. Additionally a process for systematic assessment of these codes is described which is part of a broader, industry based initiative for validation of computer codes used in all major disciplines of safety analysis. This is intended to provide both the licensee and the regulator in Canada with an objective basis for assessing the adequacy of codes for use in specific applications. Although focused specifically onmore » CANDU reactors, Canadian experience in developing advanced two-fluid codes to meet wide-ranging application needs while maintaining past investment in plant modelling provides a useful contribution to international efforts in this area.« less

ORCHID - a computer simulation of the reliability of an NDE inspection system

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

Moles, M.D.C.

1987-03-01

CANDU pressurized heavy water reactors contain several hundred horizontally-mounted zirconium alloy pressure tubes. Following a pressure tube failure, a pressure tube inspection system called CIGARette was rapidly designed, manufactured and put in operation. Defects called hydride blisters were found to be the cause of the failure, and were detected using a combination of eddy current and ultrasonic scans. A number of improvements were made to CIGARette during the inspection period. The ORCHID computer program models the operation of the delivery system, eddy current and ultrasonic systems by imitating the on-reactor decision-making procedure. ORCHID predicts that during the early stage ofmore » development, less than one blistered tube in three would be detected, while less than one in two would be detected in the middle development stage. However, ORCHID predicts that during the late development stage, probability of detection will be over 90%, primarily due to the inclusion of axial ultrasonic scans (a procedural modification). Rotational and axial slip could severely reduce probability of detection. Comparison of CIGARette's inspection data with ORCHID's predictions indicate that the latter are compatible with the actual inspection results, through the numbers are small and data uncertain. It should be emphasized that the CIGARette system has been essentially replaced with the much more reliable CIGAR system.« less

77 FR 60039 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-02

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [NRC-2011-0087] RIN 3150-AI96 Non-Power Reactor... the final regulatory basis for rulemaking to streamline non-power reactor license renewal. This final... Reactor (RTR) License Renewal Process. This contemplated rulemaking also recommends conforming changes to...

Heat-transfer analysis of double-pipe heat exchangers for indirect-cycle SCW NPP

NASA Astrophysics Data System (ADS)

Thind, Harwinder

SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 °C. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option for PT and PV SCWRs without introducing steam-reheat channels into a reactor. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, steam generators separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in a reactor containment building. This study analyzes the heat transfer from a SCW primary (reactor) loop to a SCW and Super-Heated Steam (SHS) secondary (turbine) loop using a double-pipe intermediate HX. The numerical model is developed with MATLAB and NIST REFPROP software. Water from the primary loop flows through the inner pipe, and water from the secondary loop flows through the annulus in the counter direction of the double-pipe HX. The analysis on the double-pipe HX shows temperature and profiles of thermophysical properties along the heated length of the HX. It was found that the pseudocritical region has a significant effect on the temperature profiles and heat-transfer area of the HX. An analysis shows the effect of variation in pressure, temperature, mass flow rate, and pipe size on the pseudocritical region and the heat-transfer area of the HX. The results from the numerical model can be used to optimize the heat-transfer area of the HX. The higher pressure difference on the hot side and higher temperature difference between the hot and cold sides reduces the pseudocritical-region length, thus decreases the heat-transfer surface area of the HX.

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.

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

NASA Astrophysics Data System (ADS)

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

1994-03-01

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

Application of perturbation theory to lattice calculations based on method of cyclic characteristics

NASA Astrophysics Data System (ADS)

Assawaroongruengchot, Monchai

Perturbation theory is a technique used for the estimation of changes in performance functionals, such as linear reaction rate ratio and eigenvalue affected by small variations in reactor core compositions. Here the algorithm of perturbation theory is developed for the multigroup integral neutron transport problems in 2D fuel assemblies with isotropic scattering. The integral transport equation is used in the perturbative formulation because it represents the interconnecting neutronic systems of the lattice assemblies via the tracking lines. When the integral neutron transport equation is used in the formulation, one needs to solve the resulting integral transport equations for the flux importance and generalized flux importance functions. The relationship between the generalized flux importance and generalized source importance functions is defined in order to transform the generalized flux importance transport equations into the integro-differential equations for the generalized adjoints. Next we develop the adjoint and generalized adjoint transport solution algorithms based on the method of cyclic characteristics (MOCC) in DRAGON code. In the MOCC method, the adjoint characteristics equations associated with a cyclic tracking line are formulated in such a way that a closed form for the adjoint angular function can be obtained. The MOCC method then requires only one cycle of scanning over the cyclic tracking lines in each spatial iteration. We also show that the source importance function by CP method is mathematically equivalent to the adjoint function by MOCC method. In order to speed up the MOCC solution algorithm, a group-reduction and group-splitting techniques based on the structure of the adjoint scattering matrix are implemented. A combined forward flux/adjoint function iteration scheme, based on the group-splitting technique and the common use of a large number of variables storing tracking-line data and exponential values, is proposed to reduce the computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and keff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and keff-EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these techniques to the CVR-BOC, CVR-EOC and keff-EOC adjustment of a CANDU lattice of which the burnup period is extended from 300 to 450 FPDs. The cases with the central pin containing either Dysprosium or Gadolinium in the natural Uranium are considered in our study. (Abstract shortened by UMI.)

Thermodynamic Analysis of the Use a Chemical Heat Pump to Link a Supercritical Water-Cooled Nuclear Reactor and a Thermochemical Water-Splitting Cycle for Hydrogen Production

NASA Astrophysics Data System (ADS)

Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved “steam” parameters (outlet temperatures up to 625°C and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600°C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the “nuclear” heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of the combined system comprising a SCW nuclear power generation plant and a chemical heat pump, which provides high-temperature heat to a thermochemical water splitting cycle for hydrogen production. It is concluded that the proposed chemical heat pump permits the utilization efficiency of nuclear energy to be improved by at least 2% without jeopardizing nuclear reactor safety. Based on this analysis, further research appears to be merited on the proposed advanced design of a nuclear power generation plant combined with a chemical heat pump, and implementation in appropriate applications seems worthwhile.

Eddy current proximity measurement of perpendicular tubes from within pressure tubes in CANDU nuclear reactors

NASA Astrophysics Data System (ADS)

Bennett, P. F. D.; Underhill, P. R.; Morelli, J.; Krause, T. W.

2018-04-01

Fuel channels in CANDU® (CANada Deuterium Uranium) nuclear reactors consist of two non-concentric tubes; an inner pressure tube (PT) and a larger diameter calandria tube (CT). Up to 400 horizontally mounted fuel channels are contained within a calandria vessel, which also holds the heavy water moderator. Certain fuel channels pass perpendicularly over horizontally oriented tubes (nozzles) that are part of the reactor's liquid injection shutdown system (LISS). Due to sag, these fuel channels are at risk of coming into contact with the LISS nozzles. In the event of contact between the LISS nozzle and CT, flow-induced vibrations from within the moderator could lead to fretting and deformation of the CT. LISS nozzle proximity to CTs is currently measured optically from within the calandria vessel, but from outside the fuel channels. Measurement by an independent means would provide confidence in optical results and supplement cases where optical observations are not possible. Separation of PT and CT, known as gap, is monitored from within the PT using a transmit-receive eddy current probe. Investigation of the eddy current based gap probe as a tool to also measure proximity of LISS nozzles was carried out experimentally in this work. Eddy current response as a function of LISS-PT proximity was recorded. When PT-CT gap, PT wall thickness, PT resistivity and probe lift-off variations were not present this dependence could be used to determine the LISS-PT proximity. This method has the potential to provide LISS-CT proximity using existing gap measurement data. Obtaining LISS nozzle proximity at multiple inspection intervals could be used to provide an estimate of the time to LISS-CT contact, and thereby provide a means of optimizing maintenance schedules.

The Results From the First High-Pressure Melt Ejection Test Completed in the Molten Fuel Moderator Interaction Facility at Chalk River Laboratories

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

Nitheanandan, T.; Kyle, G.; O'Connor, R.

2006-07-01

A high-pressure melt ejection test using prototypical corium was conducted at Atomic Energy of Canada Limited Chalk River Laboratories. This test was planned by the CANDU Owners Group to study the potential for an energetic interaction between molten fuel and water under postulated single-channel flow-blockage events. The experiments were designed to address regulator concerns surrounding this very low probability postulated accident events in CANDU Pressurized Heavy Water Reactors. The objective of the experimental program is to determine whether a highly energetic 'steam explosion' and associated high-pressure pulse, is possible when molten material is finely fragmented as it is ejected frommore » a fuel channel into the heavy-water moderator. The finely fragmented melt particles would transfer energy to the moderator as it is dispersed, creating a modest pressure pulse in the calandria vessel. The high-pressure melt ejection test consisted of heating up a {approx} 5 kg thermite mixture of U, U{sub 3}O{sub 8}, Zr, and CrO{sub 3} inside a 1.14-m length of insulated pressure tube. When the molten material reached the desired temperature of {approx} 2400 deg C, the pressure inside the tube was raised to 11.6 MPa, failing the pressure tube at a pre-machined flaw, and releasing the molten material into the surrounding tank of 68 deg C water. The experiment investigated the dynamic pressure history, debris size, and the effects of the material interacting with tubes representing neighbouring fuel channels. The measured mean particle size was 0.686 mm and the peak dynamic pressures were between 2.54 and 4.36 MPa, indicating that an energetic interaction between the melt and the water did not occur in the test. (authors)« less

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Wavelet analysis of poorly-focused ultrasonic signal of pressure tube inspection in nuclear industry

NASA Astrophysics Data System (ADS)

Zhao, Huan; Gachagan, Anthony; Dobie, Gordon; Lardner, Timothy

2018-04-01

Pressure tube fabrication and installment challenges combined with natural sagging over time can produce issues with probe alignment for pressure tube inspection of the primary circuit of CANDU reactors. The ability to extract accurate defect depth information from poorly focused ultrasonic signals would reduce additional inspection procedures, which leads to a significant time and cost saving. Currently, the defect depth measurement protocol is to simply calculate the time difference between the peaks of the echo signals from the tube surface and the defect from a single element probe focused at the back-wall depth. When alignment issues are present, incorrect focusing results in interference within the returning echo signal. This paper proposes a novel wavelet analysis method that employs the Haar wavelet to decompose the original poorly focused A-scan signal and reconstruct detailed information based on a selected high frequency component range within the bandwidth of the transducer. Compared to the original signal, the wavelet analysis method provides additional characteristic defect information and an improved estimate of defect depth with errors less than 5%.

A Course in Chemical Reactor Design.

ERIC Educational Resources Information Center

Takoudis, Christos G.

1983-01-01

Presents course outline, topics covered, and final project (doubling as a take home final exam) for a one-semester, interdisciplinary course on the design and behavior of chemical reactors. Interplay of chemical and physical rate processes is stressed in the course. (JM)

Evolution of interphase and intergranular strain in zirconium-niobium alloys during deformation at room temperature

NASA Astrophysics Data System (ADS)

Cai, Song

Zr-2.5Nb is currently used for pressure tubes in the CANDU (CANada Deuterium Uranium) reactor. A complete understanding of the deformation mechanism of Zr-2.5Nb is important if we are to accurately predict the in-reactor performance of pressure tubes and guarantee normal operation of the reactors. This thesis is a first step in gaining such an understanding; the deformation mechanism of ZrNb alloys at room temperature has been evaluated through studying the effect of texture and microstructure on deformation. In-situ neutron diffraction was used to monitor the evolution of the lattice strain of individual grain families along both the loading and Poisson's directions and to track the development of interphase and intergranular strains during deformation. The following experiments were carried out with data interpreted using elasto-plastic modeling techniques: (1) Compression tests of a 100%betaZr material at room temperature. (2) Tension and compression tests of hot rolled Zr-2.5Nb plate material. (3) Compression of annealed Zr-2.5Nb. (4) Cyclic loading of the hot rolled Zr-2.5Nb. (5) Compression tests of ZrNb alloys with different Nb and oxygen contents. The experimental results were interpreted using a combination of finite element (FE) and elasto-plastic self-consistent (EPSC) models. The phase properties and phase interactions well represented by the FE model, the EPSC model successfully captured the evolution of intergranular constraint during deformation and provided reasonable estimates of the critical resolved shear stress and hardening parameters of different slip systems under different conditions. The consistency of the material parameters obtained by the EPSC model allows the deformation mechanism at room temperature and the effect of textures and microstructures of ZrNb alloys to be understood. This work provides useful information towards manufacturing of Zr-2.5Nb components and helps in producing ideal microstructures and material properties for pressure tubes. Also it is helpful in guiding the development of new materials for the next generation of nuclear reactors. Furthermore, the large data set obtained from this study can be used in evaluation and improving current and future polycrystalline deformation models.

Treatment of smuggled cigarette tobacco by composting process in facultative reactors.

PubMed

Zittel, Rosimara; Pinto da Silva, Cleber; Domingues, Cinthia Eloise; de Oliveira Stremel, Tatiana Roselena; de Almeida, Thiago Eduardo; Vieira Damiani, Gislaine; Xavier de Campos, Sandro

2018-01-01

This paper presents a study on the degradation of smuggled cigarette tobacco combined with domestic organic waste and sawdust or wood chips, using facultative reactor. Four reactors with different amounts of residue were assembled. For the study of the quality of the compost obtained, physicochemical, phytotoxicity and microbiological analyses were carried out. The mixture with wood chips presented the best temperature conditions and pH variation optimizing the degradation. The final germination index (GI) values of all treatments were above the recommended GI value (50%) and the final C/N ratio between 8 and 13 indicated a mature compost. The concentration of metals under study was below the limit allowed for the commercialization. The composting carried out in all facultative reactors provided ideal conditions for the total sterilization of the final compost. Therefore, the treatment of smuggled cigarettes through facultative reactors was efficient to produce stable and mature compost. Copyright © 2017 Elsevier Ltd. All rights reserved.

75 FR 57302 - Advisory Committee on Reactor Safeguards; Public Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-20

... Monday, October 14, 2009, (74 FR 52829-52830). Thursday, October 7, 2010, Conference Room T2-B1, Two....: Final Safety Evaluation Report Associated with the Economic Simplified Boiling Water Reactor (ESBWR..., Inc. regarding the final Safety Evaluation Report associated with the ESBWR design certification...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

Localization of a Robotic Crawler for CANDU Fuel Channel Inspection

NASA Astrophysics Data System (ADS)

Manning, Mark

This thesis discusses the design and development of a pipe crawling robot for the purpose of CANDU fuel channel inspection. The pipe crawling robot shall be capable of deploying the existing CIGAR (Channel Inspection and Gauging Apparatus for Reactors) sensor head. The main focus of this thesis is the design of the localization system for this robot and the many tests that were completed to demonstrate its accuracy. The proposed localization system consists of three redundant resolver wheels mounted to the robot's frame and two resolvers that are mounted inside a custom made cable drum. This cable drum shall be referred to in this thesis as the emergency retrieval device. This device serves the dual-purpose of providing absolute position measurements (via the cable that is tethered to the robot) as well as retrieving the robot if it is inoperable. The estimated accuracy of the proposed design is demonstrated with the use of a proof-of-concept prototype and a custom made test bench that uses a vision system to provide a more accurate estimate of the robot's position. The only major difference between the proof-of-concept prototype and the proposed solution is that the more expensive radiation hardened components were not used in the proof-of-concept prototype design. For example, the proposed solution shall use radiation hardened resolver wheels, whereas the proof-of-concept prototype used encoder wheels. These encoder wheels provide the same specified accuracy as the radiation hardened resolvers for the most realistic results possible. The rationale behind the design of the proof-of-concept prototype, the proposed final design, the design of the localization system test bench, and the test plan for developing all of the components of the design related to the robot's localization system are discussed in the thesis. The test plan provides a step by step guide to the configuration and optimization of an Unscented Kalman Filter (UKF). The UKF was selected as the ideal sensor fusion algorithm for use in this application. Benchmarking was completed to compare the accuracy achieved by the UKF algorithm to other data fusion algorithms. When compared to other algorithms, the UKF demonstrated the best accuracy when considering all likely sources of error such as sensor failure and surface unevenness. The test results show that the localization system is able to achieve a worst case positional accuracy of +/- 3.6 mm for the robot crawler over the full 6350 mm distance that the robot travels inside the pressure tube. This is extrapolated from the test results completed over the shorter length test bench with simulated surface unevenness. The key benefits of the pipe crawling robot when compared to the current system include: reduced dosage to workers and the reduced outage time. The advantages are due to the fact that the robot can be automated and multiple inspection robots can be deployed simultaneously. The current inspection system is only able to complete one inspection at a time.

76 FR 52715 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-23

... review Draft Final Regulatory Guide (RG) 1.93, ``Availability of Electric Power Sources,'' Revision 1 and new Draft Final RG 1.218, ``Condition Monitoring Techniques for Electric Cables Used in Nuclear Power... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS...

76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-16

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

Investigation related to hydrogen isotopes separation by cryogenic distillation

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

Bornea, A.; Zamfirache, M.; Stefanescu, I.

2008-07-15

Research conducted in the last fifty years has shown that one of the most efficient techniques of removing tritium from the heavy water used as moderator and coolant in CANDU reactors (as that operated at Cernavoda (Romania)) is hydrogen cryogenic distillation. Designing and implementing the concept of cryogenic distillation columns require experiments to be conducted as well as computer simulations. Particularly, computer simulations are of great importance when designing and evaluating the performances of a column or a series of columns. Experimental data collected from laboratory work will be used as input for computer simulations run at larger scale (formore » The Pilot Plant for Tritium and Deuterium Separation) in order to increase the confidence in the simulated results. Studies carried out were focused on the following: - Quantitative analyses of important parameters such as the number of theoretical plates, inlet area, reflux flow, flow-rates extraction, working pressure, etc. - Columns connected in series in such a way to fulfil the separation requirements. Experiments were carried out on a laboratory-scale installation to investigate the performance of contact elements with continuous packing. The packing was manufactured in our institute. (authors)« less

Pourbaix Diagrams at Elevated Temperatures A Study of Zinc and Tin

NASA Astrophysics Data System (ADS)

Palazhchenko, Olga

Metals in industrial settings such as power plants are often subjected to high temperature and pressure aqueous environments, where failure to control corrosion compromises worker and environment safety. For instance, zircaloy (1.2-1.7 wt.% Sn) fuel rods are exposed to aqueous 250-310 °C coolant in CANDU reactors. The Pourbaix (EH-pH) diagram is a plot of electrochemical potential versus pH, which shows the domains of various metal species and by inference, corrosion susceptibility. Elevated temperature data for tin +II and tin +IV species were obtained using solid-aqueous phase equilibria with the respective oxides, in a batch vessel with in-situ pH measurement. Solubilities, determined via spectroscopic techniques, were used to calculate equilibrium constants and the Gibbs energies of Sn complexes for E-pH diagram construction. The SnOH3+ and Sn(OH )-5 species were incorporated, for the first time, into the 298.15 K and 358.15 K diagrams, with novel Go values determined at 358.15 K. Key words: Pourbaix diagrams, EH-pH, elevated temperatures, solubility, equilibrium, metal oxides, hydrolysis, redox potential, pH, thermochemical data, tin, zinc, zircaloy, corrosion, passivity.

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2014 CFR

2014-01-01

...), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation... 10 Energy 1 2014-01-01 2014-01-01 false Finality of partial decision on site suitability issues in a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2013 CFR

2013-01-01

...), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation... 10 Energy 1 2013-01-01 2013-01-01 false Finality of partial decision on site suitability issues in a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES...

NORTHROP REACTOR. REVISION NO. 1 TO THE FINAL SAFEGUARDS REPORT

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

Duncan, J.M.; Shimizu, B.; Romine, R.A.

1962-10-01

Additions and changes related to the original application concerning construction and operation of the Northrop Reactor are given. Answers to 13 questions relative to the Final Safeguards Report are included. Answers are also included concerning 12 questions regarding receipt, possession, and storage of fuel elements. Other information is included concerning personnel changes and changes from Al-clad fuel elements to improved stainless steelclad hydride elements. It is concluded that the improved Northrop Reactor does not present any undue hazard to the health and safety of the operating personnel or the public. (J.R.D.)

Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Lizon-A-Lugrin, Laure

The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a mechanical power of 1200 MW. It is observed that in most cases the landscape of Pareto's front is mostly controlled only by few key parameters. These results may be very useful for future plant design engineers. Furthermore, some calculations for pipe sizing and temperature variation between coolant and fuel have been carried out to provide an idea on their order of magnitude.

Remote Field Eddy Curent Signal Modeling for the Gap Measurement of Neighboring Tubes

NASA Astrophysics Data System (ADS)

Jung, H. K.; Lee, D. H.; Lee, Y. S.

2005-04-01

The fuel channels in the Canadian Deuterium Uranium (CANDU) reactor consist of the coaxial pressure tube (PT) and the calandria tube (CT). The Liquid injection nozzle (LIN) is cross aligned with the fuel channel to control the reactor by injecting poison. For a safe operation, the gap between the LIN and CT should be maintained in order to prevent a contact of the neighboring tubes. The remote field eddy current (RFEC) method was applied to measure the gap between a nonmagnetic Zircaloy-2 liquid injection nozzle (LIN) and a Zircaloy-2 calandria tube. Under the condition of inserting the RFEC probe into the coaxial tubes and of crossing a LIN above or under the CT, the modeling of a LIN signal is needed to check the possibility of a gap measurement. The Volume Integral Code S/W which covers the axi-symmetric 3D configuration has been very rarely applied to obtain a LIN signal. This problem was solved by assuming a LIN as a flaw which can be described as a complete 3D object. This simulated LIN signal was verified by performing the laboratory experiment. The gap between the LIN and CT can be correlated with the amplitude of the LIN signals in the voltage plane. Typical noises in the fuel channel were the relative constriction, the change in the pressure tube diameter (fill-factor), thickness variation, and so on. These noise signals were simulated by using the modeling and were analyzed by considering their dependency on the phase angle and amplitude of the voltage plane in order to separate the gap signal from them. It could be concluded that the voltage plane analysis of the simulated RFEC signals were effective for obtaining the gap measurement of the neighboring tube.

The slightly-enriched spectral shift control reactor. Final report, September 30, 1988--September 30, 1991

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

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

A low-cost municipal sewage treatment system with a combination of UASB and the "fourth-generation" downflow hanging sponge reactors.

PubMed

Tandukar, M; Uemura, S; Machdar, I; Ohashi, A; Harada, H

2005-01-01

This paper presents an evaluation of the process performance of a pilot-scale "fourth generation" downflow hanging sponge (DHS) post-treatment system combined with a UASB pretreatment unit treating municipal wastewater. After the successful operation of the second- and third-generation DHS reactors, the fourth-generation DHS reactor was developed to overcome a few shortcomings of its predecessors. This reactor was designed to further enhance the treatment efficiency and simplify the construction process in real scale, especially for the application in developing countries. Configuration of the reactor was modified to enhance the dissolution of air into the wastewater and to avert the possible clogging of the reactor especially during sudden washout from the UASB reactor. The whole system was operated at a total hydraulic retention time (HRT) of 8 h (UASB: 6 h and DHS: 2 h) for a period of over 600 days. The combined system was able to remove 96% of unfiltered BOD with only 9 mg/L remaining in the final effluent. Likewise, F. coli were removed by 3.45 log with the final count of 10(3) to 10(4) MPN/100 ml. Nutrient removal by the system was also satisfactory.

The Munich accelerator for fission fragments MAFF

NASA Astrophysics Data System (ADS)

Habs, D.; Groß, M.; Assmann, W.; Ames, F.; Bongers, H.; Emhofer, S.; Heinz, S.; Henry, S.; Kester, O.; Neumayr, J.; Ospald, F.; Reiter, P.; Sieber, T.; Szerypo, J.; Thirolf, P. G.; Varentsov, V.; Wilfart, T.; Faestermann, T.; Krücken, R.; Maier-Komor, P.

2003-05-01

The Munich Accelerator for Fission Fragments MAFF has been designed for the new Munich research reactor FRM-II. It will deliver several intense beams (˜3×10 11 s -1) of very neutron-rich fission fragments with a final energy of 30 keV (low-energy beam) or energies between 3.7 and 5.9 MeV· A (high-energy beam). Such beams are of interest for the creation of super-heavy elements by fusion reactions, nuclear spectroscopy of exotic nuclei, but they also have a potential for applications, e.g. in medicine. Presently the Munich research reactor FRM-II is ready for operation, but authorities delay the final permission to turn the reactor critical probably till the end of 2002. Only after this final permission the financing of the major parts of MAFF can start. On the other hand all major components have been designed and special components have been tested in separate setups.

SNAP (Space Nuclear Auxiliary Power) reactor overview. Final report, June 1982-December 1983

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

Voss, S.S.

1984-08-01

The SNAP reactor programs are outlined in this report. A summary of the program is included along with a technical outline of the SER, S2DR, SNAP 10A/SNAPSHOT, S8ER, and S8DR reactor systems. Specifications of the designs, the design logic and a conclusion outlining some of the program weaknesses are given.

Development of a tritium recovery system from CANDU tritium removal facility

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

Draghia, M.; Pasca, G.; Porcariu, F.

2015-03-15

The main purpose of the Tritium Recovery System (TRS) is to reduce to a maximum possible extent the release of tritium from the facility following a tritium release in confinement boundaries and also to have provisions to recover both elemental and vapors tritium from the purging gases during maintenance and components replacement from various systems processing tritium. This work/paper proposes a configuration of Tritium Recovery System wherein elemental tritium and water vapors are recovered in a separated, parallel manner. The proposed TRS configuration is a combination of permeators, a platinum microreactor (MR) and a trickle bed reactor (TBR) and consistsmore » of two branches: one branch for elemental tritium recovery from tritiated deuterium gas and the second one for tritium recovery from streams containing a significant amount of water vapours but a low amount, below 5%, of tritiated gas. The two branches shall work in a complementary manner in such a way that the bleed stream from the permeators shall be further processed in the MR and TBR in view of achieving the required decontamination level. A preliminary evaluation of the proposed TRS in comparison with state of the art tritium recovery system from tritium processing facilities is also discussed. (authors)« less

Modelling of the anti-neutrino production and spectra from a Magnox reactor

NASA Astrophysics Data System (ADS)

Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

2018-01-01

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor operation was calculated. The results of the preliminary calculations are shown and limitations in the methods and data discussed.

The slightly-enriched spectral shift control reactor

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

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

Advanced shield development for a fission surface power system for the lunar surface

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

A. E. Craft; I. J. Silver; C. M. Clark

A nuclear reactor power system such as the affordable fission surface power system enables a potential outpostonthemoon.Aradiation shieldmustbe included in the reactor system to reduce the otherwise excessive dose to the astronauts and other vital system components. The radiation shield is typically the most massive component of a space reactor system, and thus must be optimized to reduce mass asmuchas possible while still providing the required protection.Various shield options for an on-lander reactor system are examined for outpost distances of 400m and 1 kmfromthe reactor. Also investigated is the resulting mass savings from the use of a high performance cermetmore » fuel. A thermal analysis is performed to determine the thermal behaviours of radiation shields using borated water. For an outpost located 1000m from the core, a tetramethylammonium borohydride shield is the lightest (5148.4 kg), followed by a trilayer shield (boron carbide–tungsten–borated water; 5832.3 kg), and finally a borated water shield (6020.7 kg). In all of the final design cases, the temperature of the borated water remains below 400 K.« less

The Management of the Radioactive Waste Generated by Cernavoda NPP, Romania, an Example of International Cooperation - 13449

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

Barariu, Gheorghe

2013-07-01

The design criteria and constraints for the development of the management strategy for radioactive waste generated from operating and decommissioning of CANDU Nuclear Units from Cernavoda NPP in Romania, present many specific aspects. The main characteristics of CANDU type waste are its high concentrations of tritium and radiocarbon. Also, the existing management strategy for radioactive waste at Cernavoda NPP provides no treatment or conditioning for radioactive waste disposal. These characteristics embodied a challenging effort, in order to select a proper strategy for radioactive waste management at present, when Romania is an EU member and a signatory country of the Jointmore » Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. The helping of advanced countries in radioactive waste management, directly or into the frame of the international organizations, like IAEA, become solve the aforementioned challenges at adequate level. (authors)« less

The effect of mixing ratio variation of sludge and organic solid waste on biodrying process

NASA Astrophysics Data System (ADS)

Nasution, A. C.; Kristanto, G. A.

2018-01-01

In this study, organic waste was co-biodried with sludge cake to determine which mixing ratio gave the best result. The organic waste was consisted of dried leaves and green leaves, while the sludge cake was obtained from a waste water treatment plant in Bekasi. The experiment was performed on 3 lab-scale reactors with same specifications. After 21 days of experiment, it was found that the reactor with the lowest mixing fraction of sludge (5:1) has the best temperature profile and highest moisture content depletion compared with others. Initial moisture content and initial volatile solid content of this reactor’s feedstock was 52.25% and 82.4% respectively. The airflow rate was 10 lpm. After biodrying was done, the final moisture content of the feedstock from Reactor C was 22.0% and the final volatile solid content was 75.9%.The final calorific value after biodrying process was 3179,28kcal/kg.

Reactor-Scale Cultivation of the Hyperthermophilic Methanarchaeon Methanococcus jannaschii to High Cell Densities

PubMed Central

Mukhopadhyay, Biswarup; Johnson, Eric F.; Wolfe, Ralph S.

1999-01-01

For the hyperthermophilic and barophilic methanarchaeon Methanococcus jannaschii, we have developed a medium and protocols for reactor-scale cultivation that improved the final cell yield per liter from ∼0.5 to ∼7.5 g of packed wet cells (∼1.8 g dry cell mass) under autotrophic growth conditions and to ∼8.5 g of packed wet cells (∼2 g dry cell mass) with yeast extract (2 g liter−1) and tryptone (2 g liter−1) as medium supplements. For growth in a sealed bottle it was necessary to add Se to the medium, and a level of 2 μM for added Se gave the highest final cell yield. In a reactor M. jannaschii grew without added Se in the medium; it is plausible that the cells received Se as a contaminant from the reactor vessel and the H2S supply. But, for the optimal performance of a reactor culture, an addition of Se to a final concentration of 50 to 100 μM was needed. Also, cell growth in a reactor culture was inhibited at much higher Se concentrations. These observations and the data from previous work with methanogen cell extracts (B. C. McBride and R. S. Wolfe, Biochemistry 10:4312–4317, 1971) suggested that from a continuously sparged reactor culture Se was lost in the exhaust gas as volatile selenides, and this loss raised the apparent required level of and tolerance for Se. In spite of having a proteinaceous cell wall, M. jannaschii withstood an impeller tip speed of 235.5 cms−1, which was optimal for achieving high cell density and also was the higher limit for the tolerated shear rate. The organism secreted one or more acidic compounds, which lowered pH in cultures without pH control; this secretion continued even after cessation of growth. PMID:10543823

Graphite distortion ``C`` Reactor. Final report

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

Wood, N.H.

1962-02-08

This report covers the efforts of the Laboratory in an investigation of the graphite distortion in the ``C`` reactor at Hanford. The particular aspects of the problem to be covered by the Laboratory were possible ``fixes`` to the control rod sticking problem caused by VSR channel distortion.

Decommissioning of the Northrop TRIGA reactor

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

Cozens, George B.; Woo, Harry; Benveniste, Jack

1986-07-01

An overview of the administrative and operational aspects of decommissioning and dismantling the Northrop Mark F TRIGA Reactor, including: planning and preparation, personnel requirements, government interfacing, costs, contractor negotiations, fuel shipments, demolition, disposal of low level waste, final survey and disposition of the concrete biological shielding. (author)

Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

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

Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

2016-12-01

A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulationmore » tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.« less

Synthesis of carbohydrates in a continuous flow reactor by immobilized phosphatase and aldolase.

PubMed

Babich, Lara; Hartog, Aloysius F; van Hemert, Lieke J C; Rutjes, Floris P J T; Wever, Ron

2012-12-01

Herein, we report a new flow process with immobilized enzymes to synthesize complex chiral carbohydrate analogues from achiral inexpensive building blocks in a three-step cascade reaction. The first reactor contained immobilized acid phosphatase, which phosphorylated dihydroxyacetone to dihydroxyacetone phosphate using pyrophosphate as the phosphate donor. The second flow reactor contained fructose-1,6-diphosphate aldolase (RAMA, rabbit muscle aldolase) or rhamnulose-1-phosphate aldolase (RhuA from Thermotoga maritima) and acid phosphatase. The immobilized aldolases coupled the formed dihydroxyacetone phosphate to aldehydes, resulting in phosphorylated carbohydrates. A final reactor containing acid phosphatase that dephosphorylated the phosphorylated product yielded the final product. Different aldehydes were used to synthesize carbohydrates on a gram scale. To demonstrate the feasibility of the flow systems, we synthesized 0.6 g of the D-fagomine precursor. By using immobilized aldolase RhuA we were also able to obtain other stereoisomers of the D-fagomine precursor. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

PubMed

Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

2011-01-01

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

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

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely themore » total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.« less

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2010 CFR

2010-01-01

... required for issuance of a construction permit, operating license, combined license, in any hearing under... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases for not incorporating severe accident mitigation design alternatives into the design of the reactor to...

Safety approach to the selection of design criteria for the CRBRP reactor refueling system

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

Meisl, C J; Berg, G E; Sharkey, N F

1979-01-01

The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents.more » The process steps are illustrated by examples.« less

Reactor-Produced Medical Radionuclides

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

Mirzadeh, Saed; Mausner, Leonard; Garland, Marc A

2011-01-01

The therapeutic use of radionuclides in nuclear medicine, oncology and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by beta emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chaptermore » is concluded by providing a number of examples encompassing the various possible reaction routes for production of a number of medical radionuclides in a reactor.« less

SP-100 Program: space reactor system and subsystem investigations

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

Harty, R.B.

1983-09-30

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

SP-100 program: Space reactor system and subsystem investigations

NASA Astrophysics Data System (ADS)

Harty, R. B.

1983-09-01

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. The nuclear safety review/approval process that is required for a space reactor system is summarized. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that is expected and to provide information that could be usable in future programs.

10 CFR 50.82 - Termination of license.

Code of Federal Regulations, 2011 CFR

2011-01-01

... to the NRC, consistent with the requirements of § 50.4(b)(8); (ii) Once fuel has been permanently... fuel from the reactor vessel, or when a final legally effective order to permanently cease operations... emplacement or retention of fuel into the reactor vessel. (3) Decommissioning will be completed within 60...

Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report

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

Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

1992-03-01

This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

Bioaugmentation of activated sludge towards 3-chloroaniline removal with a mixed bacterial population carrying a degradative plasmid.

PubMed

Bathe, Stephan; Schwarzenbeck, Norbert; Hausner, Martina

2009-06-01

A bioaugmentation approach combining several strategies was applied to achieve degradation of 3-chloroaniline (3CA) in semicontinuous activated sludge reactors. In a first step, a 3CA-degrading Comamonas testosteroni strain carrying the degradative plasmid pNB2 was added to a biofilm reactor, and complete 3CA degradation together with spread of the plasmid within the indigenous biofilm population was achieved. A second set of reactors was then bioaugmented with either a suspension of biofilm cells removed from the carrier material or with biofilm-containing carrier material. 3CA degradation was established rapidly in all bioaugmented reactors, followed by a slow adaptation of the non-bioaugmented control reactors. In response to variations in 3CA concentration, all reactors exhibited temporary performance breakdowns. Whereas duplicates of the control reactors deviated in their behaviour, the bioaugmented reactors appeared more reproducible in their performance and population dynamics. Finally, the carrier-bioaugmented reactors showed an improved performance in the presence of high 3CA influent concentrations over the suspension-bioaugmented reactors. In contrast, degradation in one control reactor failed completely, but was rapidly established in the remaining control reactor.

Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

DTIC Science & Technology

1987-06-01

computer modeling remains at best semiempirical (C-i), this large variation in scaling factor makes extrapolation of data impossible. The DIDO Water...in a full scale PWR are not practical. The reactor plant is not controlled to tolerances necessary for research, and utilities are reluctant to vary...MIT Reactor Safeguards Committee, in revision 1 to the PCCL Safety Evaluation Report (SER), for final approval to begin in-pile testing and

DE-NE0008277_PROTEUS final technical report 2018

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

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

Calculation to experiment comparison of SPND signals in various nuclear reactor environments

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

Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien

2015-07-01

In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first partmore » of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)« less

Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 1

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

None

1975-12-01

Information is presented under the following section headings: LMFBR program options and their compatibility with the major issues affecting commercial development, Proposed Final Environmental Statement for the LMFBR program, December 1974, WASH-1535, supplemental material, and material relating to Proposed Final Environmental Statement review. (DG)

Taming The Next Set of Strategic Weapons Threats

DTIC Science & Technology

2006-06-01

Reactors Victor Gilinsky . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 6. Coping with Biological Threats after...Regime (MTCR) is not yet optimized to cope with these challenges. Finally, nuclear technologies have become much more difficult to control. New...resistance of the most popular type of power reactor concludes that the current international nuclear safeguards system needs to be modified to cope

Chemical vapor deposition of epitaxial silicon

DOEpatents

Berkman, Samuel

1984-01-01

A single chamber continuous chemical vapor deposition (CVD) reactor is described for depositing continuously on flat substrates, for example, epitaxial layers of semiconductor materials. The single chamber reactor is formed into three separate zones by baffles or tubes carrying chemical source material and a carrier gas in one gas stream and hydrogen gas in the other stream without interaction while the wafers are heated to deposition temperature. Diffusion of the two gas streams on heated wafers effects the epitaxial deposition in the intermediate zone and the wafers are cooled in the final zone by coolant gases. A CVD reactor for batch processing is also described embodying the deposition principles of the continuous reactor.

Changing concepts of geologic structure and the problem of siting nuclear reactors: Examples from Washington State

NASA Astrophysics Data System (ADS)

Tabor, R. W.

1986-09-01

The conflict between regulation and healthy evolution of geological science has contributed to the difficulties of siting nuclear reactors. On the Columbia Plateau in Washington, but for conservative design of the Hanford reactor facility, the recognition of the little-understood Olympic-Wallowa lineament as a major, possibly still active structural alinement might have jeopardized the acceptability of the site for nuclear reactors. On the Olympic Peninsula, evolving concepts of compressive structures and their possible recent activity and the current recognition of a subducting Juan de Fuca plate and its potential for generating great earthquakes—both concepts little-considered during initial site selection—may delay final acceptance of the Satsop site. Conflicts of this sort are inevitable but can be accommodated if they are anticipated in the reactor-licensing process. More important, society should be increasing its store of geologic knowledge now, during the current recess in nuclear reactor siting.

Development of a Model and Computer Code to Describe Solar Grade Silicon Production Processes

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

Mathematical models and computer codes based on these models, which allow prediction of the product distribution in chemical reactors for converting gaseous silicon compounds to condensed-phase silicon were developed. The following tasks were accomplished: (1) formulation of a model for silicon vapor separation/collection from the developing turbulent flow stream within reactors of the Westinghouse (2) modification of an available general parabolic code to achieve solutions to the governing partial differential equations (boundary layer type) which describe migration of the vapor to the reactor walls, (3) a parametric study using the boundary layer code to optimize the performance characteristics of the Westinghouse reactor, (4) calculations relating to the collection efficiency of the new AeroChem reactor, and (5) final testing of the modified LAPP code for use as a method of predicting Si(1) droplet sizes in these reactors.

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

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

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Development of a reactor with carbon catalysts for modular-scale, low-cost electrochemical generation of H 2O 2

DOE PAGES

Chen, Zhihua; Chen, Shucheng; Siahrostami, Samira; ...

2017-03-01

The development of small-scale, decentralized reactors for H 2O 2 production that can couple to renewable energy sources would be of great benefit, particularly for water purification in the developing world. Herein, we describe our efforts to develop electrochemical reactors for H 2O 2 generation with high Faradaic efficiencies of >90%, requiring cell voltages of only ~1.6 V. The reactor employs a carbon-based catalyst that demonstrates excellent performance for H 2O 2 production under alkaline conditions, as demonstrated by fundamental studies involving rotating-ring disk electrode methods. Finally, the low-cost, membrane-free reactor design represents a step towards a continuous, modular-scale, de-centralizedmore » production of H 2O 2.« less

Long lifetime fast spectrum reactor for lunar surface power system

NASA Astrophysics Data System (ADS)

Kambe, Mitsuru

1993-01-01

In the framework of innovative reactor research activities, a conceptual design study of fast spectrum reactor and primary system for 800 kWe lunar surface power system to be combined with potassium Rankine cycle power conversion has been conducted to meet the power requirements of the lunar base activities in the next century. The reactor subsystem is characterized by RAPID (Refueling by All Pins Integrated Design) concept to enhance inherent safety and to enable quick and simplifed refueling in every 10 years. RAPID concept affords power plant design lifetime of up to 30 years. Integrity of the reactor structure and replacement of failed primary circuits are also discussed. Substantial reduction in per-kWh cost on considering launch, emplacement, and final disposition can be expected by a long system lifetime.

Microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

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

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

78 FR 3470 - DTE Electric Company (Formerly the Detroit Edison Company), Notice of Availability of Final...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-16

...--specific Web page at http://www.nrc.gov/reactors/new-reactors/col/fermi.html . The Ellis Library and... possesses and are publicly-available, using any of the following methods: Federal Rulemaking Web site: Go to... Documents Access and Management System (ADAMS): You may access publicly-available documents online in the...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2012 CFR

2012-01-01

... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, shall determine whether the... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2011 CFR

2011-01-01

... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, shall determine whether the... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF...

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

A simple, space constrained NIRIM type reactor for chemical vapour deposition of diamond

NASA Astrophysics Data System (ADS)

Thomas, Evan L. H.; Ginés, Laia; Mandal, Soumen; Klemencic, Georgina M.; Williams, Oliver A.

2018-03-01

In this paper the design of a simple, space constrained chemical vapour deposition reactor for diamond growth is detailed. Based on the design by NIRIM, the reactor is composed of a quartz discharge tube placed within a 2.45 GHz waveguide to create the conditions required for metastable growth of diamond. Utilising largely off-the-shelf components and a modular design, the reactor allows for easy modification, repair, and cleaning between growth runs. The elements of the reactor design are laid out with the CAD files, parts list, and control files made easily available to enable replication. Finally, the quality of nanocrystalline diamond films produced are studied with SEM and Raman spectroscopy, with the observation of clear faceting and a large diamond fraction suggesting the design offers deposition of diamond with minimal complexity.

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

Damper mechanism for nuclear reactor control elements

DOEpatents

Taft, William Elwood

1976-01-01

A damper mechanism which provides a nuclear reactor control element decelerating function at the end of the scram stroke. The total damping function is produced by the combination of two assemblies, which operate in sequence. First, a tapered dashram assembly decelerates the control element to a lower velocity, after which a spring hydraulic damper assembly takes over to complete the final damping.

Glycerol Production and Transformation: A Critical Review with Particular Emphasis on Glycerol Reforming Reaction for Producing Hydrogen in Conventional and Membrane Reactors.

PubMed

Bagnato, Giuseppe; Iulianelli, Adolfo; Sanna, Aimaro; Basile, Angelo

2017-03-23

Glycerol represents an emerging renewable bio-derived feedstock, which could be used as a source for producing hydrogen through steam reforming reaction. In this review, the state-of-the-art about glycerol production processes is reviewed, with particular focus on glycerol reforming reactions and on the main catalysts under development. Furthermore, the use of membrane catalytic reactors instead of conventional reactors for steam reforming is discussed. Finally, the review describes the utilization of the Pd-based membrane reactor technology, pointing out the ability of these alternative fuel processors to simultaneously extract high purity hydrogen and enhance the whole performances of the reaction system in terms of glycerol conversion and hydrogen yield.

Glycerol Production and Transformation: A Critical Review with Particular Emphasis on Glycerol Reforming Reaction for Producing Hydrogen in Conventional and Membrane Reactors

PubMed Central

Bagnato, Giuseppe; Iulianelli, Adolfo; Sanna, Aimaro; Basile, Angelo

2017-01-01

Glycerol represents an emerging renewable bio-derived feedstock, which could be used as a source for producing hydrogen through steam reforming reaction. In this review, the state-of-the-art about glycerol production processes is reviewed, with particular focus on glycerol reforming reactions and on the main catalysts under development. Furthermore, the use of membrane catalytic reactors instead of conventional reactors for steam reforming is discussed. Finally, the review describes the utilization of the Pd-based membrane reactor technology, pointing out the ability of these alternative fuel processors to simultaneously extract high purity hydrogen and enhance the whole performances of the reaction system in terms of glycerol conversion and hydrogen yield. PMID:28333121

Enzymatic membrane reactors for biodegradation of recalcitrant compounds. Application to dye decolourisation.

PubMed

López, C; Mielgo, I; Moreira, M T; Feijoo, G; Lema, J M

2002-11-13

Membrane bioreactors are being increasingly used in enzymatic catalysed transformations. However, the application of enzymatic-based treatment systems in the environmental field is rather unusual. The aim of this paper is to overview the application of enzymatic membrane reactors to wastewater treatment, more specifically to dye decolourisation. Firstly, the basic aspects such as different configurations of enzymatic reactors, advantages and disadvantages associated to their utilisation are revised as well as the application of this technology to wastewater treatment. Secondly, dye decolourisation by white-rot fungi and their oxidative enzymes are discussed, presenting an overall view from for in vivo and in vitro systems. Finally, dye decolourisation by manganese peroxidase in an enzymatic membrane reactor in continuous operation is presented.

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.

75 FR 32972 - Office of New Reactors: Notice of Availability of the Final Staff Guidance; Section 14.3.12 on...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-10

... the Final Staff Guidance; Section 14.3.12 on Physical Security Hardware Inspections, Tests, Analyses...: The NRC is issuing its Final Revision 1 to NUREG-0800, ``Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants,'' Section 14.3.12 on ``Physical Security Hardware...

75 FR 68009 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-04

... the Final Staff Guidance Standard Review Plan Section 13.6.3, Revision 1 on Physical Security--Early... NRC is issuing its Final Revision 1 to NUREG-0800, ``Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants,'' Section 13.6.3, Revision 1 on Physical Security...

75 FR 68009 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-04

... the Final Staff Guidance Standard Review Plan Section 13.6.2, Revision 1 on Physical Security--Design... issuing its Final Revision 1 to NUREG-0800, ``Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants,'' Section 13.6.2, Revision 1 on Physical Security--Design Certification...

75 FR 68009 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-04

... the Final Staff Guidance Standard Review Plan, Section 13.6.1, Revision 1 on Physical Security... Availability. SUMMARY: The NRC is issuing its Final Revision 1 to NUREG-0800, ``Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants,'' Section 13.6.1, Revision 1 on...

Entropy Production in Chemical Reactors

NASA Astrophysics Data System (ADS)

Kingston, Diego; Razzitte, Adrián C.

2017-06-01

We have analyzed entropy production in chemically reacting systems and extended previous results to the two limiting cases of ideal reactors, namely continuous stirred tank reactor (CSTR) and plug flow reactor (PFR). We have found upper and lower bounds for the entropy production in isothermal systems and given expressions for non-isothermal operation and analyzed the influence of pressure and temperature in entropy generation minimization in reactors with a fixed volume and production. We also give a graphical picture of entropy production in chemical reactions subject to constant volume, which allows us to easily assess different options. We show that by dividing a reactor into two smaller ones, operating at different temperatures, the entropy production is lowered, going as near as 48 % less in the case of a CSTR and PFR in series, and reaching 58 % with two CSTR. Finally, we study the optimal pressure and temperature for a single isothermal PFR, taking into account the irreversibility introduced by a compressor and a heat exchanger, decreasing the entropy generation by as much as 30 %.

Interim Safe Storage of Plutonium Production Reactors at the US DOE Hanford Site - 13438

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

Schilperoort, Daryl L.; Faulk, Darrin

2013-07-01

Nine plutonium production reactors located on DOE's Hanford Site are being placed into an Interim Safe Storage (ISS) period that extends to 2068. The Environmental Impact Statement (EIS) for ISS [1] was completed in 1993 and proposed a 75-year storage period that began when the EIS was finalized. Remote electronic monitoring of the temperature and water level alarms inside the safe storage enclosure (SSE) with visual inspection inside the SSE every 5 years are the only planned operational activities during this ISS period. At the end of the ISS period, the reactor cores will be removed intact and buried inmore » a landfill on the Hanford Site. The ISS period allows for radioactive decay of isotopes, primarily Co-60 and Cs-137, to reduce the dose exposure during disposal of the reactor cores. Six of the nine reactors have been placed into ISS by having an SSE constructed around the reactor core. (authors)« less

Bioelectrochemical enhancement of methane production from highly concentrated food waste in a combined anaerobic digester and microbial electrolysis cell.

PubMed

Park, Jungyu; Lee, Beom; Tian, Donjie; Jun, Hangbae

2018-01-01

A microbial electrolysis cell (MEC) is a promising technology for enhancing biogas production from an anaerobic digestion (AD) reactor. In this study, the effects of the MEC on the rate of methane production from food waste were examined by comparing an AD reactor with an AD reactor combined with a MEC (AD+MEC). The use of the MEC accelerated methane production and stabilization via rapid organic oxidation and rapid methanogenesis. Over the total experimental period, the methane production rate and stabilization time of the AD+MEC reactor were approximately 1.7 and 4.0 times faster than those of the AD reactor. Interestingly however, at the final steady state, the methane yields of both the reactors were similar to the theoretical maximum methane yield. Based on these results, the MEC did not increase the methane yield over the theoretical value, but accelerated methane production and stabilization by bioelectrochemical reactions. Copyright © 2017 Elsevier Ltd. All rights reserved.

NACA Zero Power Reactor Facility Hazards Summary

NASA Technical Reports Server (NTRS)

1957-01-01

The Lewis Flight Propulsion Laboratory of the National Advisory Committee for Aeronautics proposes to build a zero power research reactor facility which will be located in the laboratory grounds near Clevelaurd, Ohio. The purpose of this report is to inform the Advisory Commit tee on Reactor Safeguards of the U. S. Atomic Energy Commission in re gard to the design of the reactor facility, the cha,acteristics of th e site, and the hazards of operation at this location, The purpose o f this reactor is to perform critical experiments, to measure reactiv ity effects, to serve as a neutron source, and to serve as a training tool. The reactor facility is described. This is followed by a discu ssion of the nuclear characteristics and the control system. Site cha racteristics are then discussed followed by a discussion of the exper iments which may be conducted in the facility. The potential hazards of the facility are then considered, particularly, the maximum credib le accident. Finally, the administrative procedure is discussed.

Etude thermo-hydraulique de l'ecoulement du moderateur dans le reacteur CANDU-6

NASA Astrophysics Data System (ADS)

Mehdi Zadeh, Foad

Etant donne la taille (6,0 m x 7,6 m) ainsi que le domaine multiplement connexe qui caracterisent la cuve des reacteurs CANDU-6 (380 canaux dans la cuve), la physique qui gouverne le comportement du fluide moderateur est encore mal connue de nos jours. L'echantillonnage de donnees dans un reacteur en fonction necessite d'apporter des changements a la configuration de la cuve du reacteur afin d'y inserer des sondes. De plus, la presence d'une zone intense de radiations empeche l'utilisation des capteurs courants d'echantillonnage. En consequence, l'ecoulement du moderateur doit necessairement etre etudie a l'aide d'un modele experimental ou d'un modele numerique. Pour ce qui est du modele experimental, la fabrication et la mise en fonction de telles installations coutent tres cher. De plus, les parametres de la mise a l'echelle du systeme pour fabriquer un modele experimental a l'echelle reduite sont en contradiction. En consequence, la modelisation numerique reste une alternative importante. Actuellement, l'industrie nucleaire utilise une approche numerique, dite de milieu poreux, qui approxime le domaine par un milieu continu ou le reseau des tubes est remplace par des resistances hydrauliques distribuees. Ce modele est capable de decrire les phenomenes macroscopiques de l'ecoulement, mais ne tient pas compte des effets locaux ayant un impact sur l'ecoulement global, tel que les distributions de temperatures et de vitesses a proximite des tubes ainsi que des instabilites hydrodynamiques. Dans le contexte de la surete nucleaire, on s'interesse aux effets locaux autour des tubes de calandre. En effet, des simulations faites par cette approche predisent que l'ecoulement peut prendre plusieurs configurations hydrodynamiques dont, pour certaines, l'ecoulement montre un comportement asymetrique au sein de la cuve. Ceci peut provoquer une ebullition du moderateur sur la paroi des canaux. Dans de telles conditions, le coefficient de reactivite peut varier de maniere importante, se traduisant par l'accroissement de la puissance du reacteur. Ceci peut avoir des consequences majeures pour la surete nucleaire. Une modelisation CFD (Computational Fluid Dynamics) detaillee tenant compte des effets locaux s'avere donc necessaire. Le but de ce travail de recherche est de modeliser le comportement complexe de l'ecoulement du moderateur au sein de la cuve d'un reacteur nucleaire CANDU-6, notamment a proximite des tubes de calandre. Ces simulations servent a identifier les configurations possibles de l'ecoulement dans la calandre. Cette etude consiste ainsi a formuler des bases theoriques a l'origine des instabilites macroscopiques du moderateur, c.-a-d. des mouvements asymetriques qui peuvent provoquer l'ebullition du moderateur. Le defi du projet est de determiner l'impact de ces configurations de l'ecoulement sur la reactivite du reacteur CANDU-6.

Conceptual design of BNCT facility based on the TRR medical room

NASA Astrophysics Data System (ADS)

Golshanian, M.; Rajabi, A. A.; Kasesaz, Y.

2017-10-01

This paper presents a conceptual design of the Boron Neutron Capture Therapy (BNCT) facility based on the medical room of Tehran Research Reactor (TRR). The medical room is located behind the east wall of the reactor pool. The designed beam line is an in-pool Beam Shaping Assembly (BSA) which is considered between the reactor core and the medical room wall. The final designed BSA can provide 2.96× 109 n/cm2ṡs epithermal neutron flux at the irradiation position with acceptable beam contamination to use as a clinical BNCT.

Calculation and Experiment of Adding Top Beryllium Shims for Iran MNSR

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

Ebadati, Javad; Rezvanifard, Mehdi; Shahabi, Iraj

2006-07-01

Miniature Neutron Source Reactor which is called MNSR were put into operation on June 1994 in Esfahan Nuclear Technology Center (ENTC). At that time the excess reactivity at the cold condition was 3.85 mk. After 7 years of operation and fuel consumption the reactivity was reduced to 2.90 mk. To compensate this reduction and upgrade the reactor, Beryllium Shim were used at the top of the core. This paper discusses the steps for this accurate and sensitive task. Finally a layer of 1.5 mm Beryllium were added to restore the reactor life. (authors)

N.S. SAVANNAH, DRAFT OF FINAL SAFEGUARDS REPORT TEST, START-UP AND TRIALS, NEW YORK SHIPBUILDING CORPORATION

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

None

1960-04-01

The N. S. Savannah program for testing, start-up, and initial operation of all reactor and propulsion components and systems is discussed. Definitions of test phases are given and various stages of the test program are outlined. A list of tests for the various reactor, propulsion, and other system components is included. (C.J.G.)

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Finality of manufacturing licenses; information requests. 52.171 Section 52.171 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases...

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Finality of manufacturing licenses; information requests. 52.171 Section 52.171 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases...

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Finality of manufacturing licenses; information requests. 52.171 Section 52.171 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases...

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Finality of manufacturing licenses; information requests. 52.171 Section 52.171 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases...

The prototype fast reactor at Dounreay, Scotland. Process and engineering development for sodium removal

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

Mann, A.; Herrick, R.; Gunn, J.

2007-07-01

Dounreay was home to commercial fast reactor development in the UK. Following the construction and operation of the Dounreay Fast Reactor, a sodium-cooled Prototype Fast Reactor (PFR), was constructed. PFR started operating in 1974, closed in 1994 and is presently being decommissioned. To date the bulk of the sodium has been removed and treated. Due to the design of the existing extraction system however, a sodium pool will remain in the heel of the reactor. To remove this sodium, a pump/camera system was developed, tested and deployed. The Water Vapour Nitrogen (WVN) process has been selected to allow removal ofmore » the final sodium residues from the reactor. Due to the design of the reactor and potential for structural damage should Normal WVN (which produces hydrated sodium hydroxide) be used, Low Concentration WVN (LC WVN) has been developed. Pilot scale testing has shown that it is possible treat the reactor within 18 months at a WVN concentration of up to 4% v/v and temperature of 120 deg. C. At present the equipment that will be used to apply LC WVN to the reactor is being developed at the detail design stage. and is expected to be deployed within the next few years. (authors)« less

Final report on the oxidation of energetic materials in supercritical water. Final Air Force report

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

Buelow, S.J.; Allen, D.; Anderson, G.K.

1995-04-03

The objective of this project was to determine the suitability of oxidation in supercritical fluids (SCO), particularly water (SCWO), for disposal of propellants, explosives, and pyrotechnics (PEPs). The SCO studies of PEPs addressed the following issues: The efficiency of destruction of the substrate. The products of destruction contained in the effluents. Whether the process can be conducted safely on a large scale. Whether energy recovery from the process is economically practicable. The information essential for process development and equipment design was also investigated, including issues such as practical throughput of explosives through a SCWO reactor, reactor materials and corrosion, andmore » models for process design and optimization.« less

75 FR 70304 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-17

..., Rockville, Maryland. Thursday, December 2, 2010, Conference Room T2-B1, Two White Flint North, Rockville... opening remarks regarding the conduct of the meeting. 8:35 a.m.-10 a.m.: Final Safety Evaluation Report... Kewaunee, Inc., regarding the final safety evaluation report associated with the License Renewal...

Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 3

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

None

1975-12-01

Included are copies of thirty-four comment letters on the Proposed Final Environmental Statement together with the ERDA replies to these letters. The letters were received from Federal, State, and local agencies, environmental and public interest groups, members of the academic and industrial communities, and individual citizens. (DG)

Characterization of deformation mechanisms in zirconium alloys: effect of temperature and irradiation

NASA Astrophysics Data System (ADS)

Long, Fei

Zirconium alloys have been widely used in the CANDU (CANada Deuterium Uranium) reactor as core structural materials. Alloy such as Zircaloy-2 has been used for calandria tubes; fuel cladding; the pressure tube is manufactured from alloy Zr-2.5Nb. During in-reactor service, these alloys are exposed to a high flux of fast neutron at elevated temperatures. It is important to understand the effect of temperature and irradiation on the deformation mechanism of zirconium alloys. Aiming to provide experimental guidance for future modeling predictions on the properties of zirconium alloys this thesis describes the result of an investigation of the change of slip and twinning modes in Zircaloy-2 and Zr-2.5Nb as a function of temperature and irradiation. The aim is to provide scientific fundamentals and experimental evidences for future industry modeling in processing technique design, and in-reactor property change prediction of zirconium components. In situ neutron diffraction mechanical tests carried out on alloy Zircaloy-2 at three temperatures: 100¢ªC, 300¢ªC, and 500¢ªC, and described in Chapter 3. The evolution of the lattice strain of individual grain families in the loading and Poisson's directions during deformation, which probes the operation of slip and twinning modes at different stress levels, are described. By using the same type of in situ neutron diffraction technique, tests on Zr-2.5Nb pressure tube material samples, in either the fast-neutron irradiated or un-irradiated condition, are reported in Chapter 4. In Chapter 5, the measurement of dislocation density by means of line profile analysis of neutron diffraction patterns, as well as TEM observations of the dislocation microstructural evolution, is described. In Chapter 6 a hot-rolled Zr-2.5Nb with a larger grain size compared with the pressure tubing was used to study the development of dislocation microstructures with increasing plastic strain. In Chapter 7, in situ loading of heavy ion irradiated hot-rolled Zr-2.5Nb alloy is described, providing evidence for the interaction between moving dislocations and irradiation induced loops. Chapter 8 gives the effect on the dislocation structure of different levels of compressive strains along two directions in the hot-rolled Zr-2.5Nb alloy. By using high resolution neutron diffraction and TEM observations, the evolution of type and dislocation densities, as well as changes of dislocation microstructure with plastic strain were characterized.

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

Nygaard, E. T.; Williams, M. M. R.; Angelo, P. L.

Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W'smore » proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)« less

Reactor Operations Monitoring System

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

Hart, M.M.

1989-01-01

The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

GUM Analysis for TIMS and SIMS Isotopic Ratios in Graphite

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

Heasler, Patrick G.; Gerlach, David C.; Cliff, John B.

2007-04-01

This report describes GUM calculations for TIMS and SIMS isotopic ratio measurements of reactor graphite samples. These isotopic ratios are used to estimate reactor burn-up, and currently consist of various ratios of U, Pu, and Boron impurities in the graphite samples. The GUM calculation is a propagation of error methodology that assigns uncertainties (in the form of standard error and confidence bound) to the final estimates.

Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing 235U, 233U, and 232Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

NASA Astrophysics Data System (ADS)

Ioffe, B. L.; Kochurov, B. P.

2012-02-01

A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of 235U. It operates in the open-cycle mode involving 233U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

Integrated process of distillation with side reactors for synthesis of organic acid esters

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

Panchal, Chandrakant B; Prindle, John C; Kolah, Aspri

An integrated process and system for synthesis of organic-acid esters is provided. The method of synthesizing combines reaction and distillation where an organic acid and alcohol composition are passed through a distillation chamber having a plurality of zones. Side reactors are used for drawing off portions of the composition and then recycling them to the distillation column for further purification. Water is removed from a pre-reactor prior to insertion into the distillation column. An integrated heat integration system is contained within the distillation column for further purification and optimizing efficiency in the obtaining of the final product.

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

Recovery of elemental sulphur from anaerobic effluents through the biological oxidation of sulphides.

PubMed

de Sousa, José Tavares; Lima, Jéssyca de Freitas; da Silva, Valquíria Cordeiro; Leite, Valderi Duarte; Lopes, Wilton Silva

2017-03-01

The aim of the present study was to evaluate the biological oxidation of sulphide in two different UASB reactors by assessing the occurrence of oxidized forms of sulphur in the effluents and the amount of S 0 that could be recovered in the process. The bioreactors employed were an anaerobic hybrid (AH) reactor employing porous polyurethane foam as support media and a micro-aerated UASB reactor equipped with an aeration device above the digestion zone. The AH reactor produced a final effluent containing low concentrations of S 2- (3.87% of total sulphur load). It was achieved due to a complete oxidation of 56.1% of total sulphur. The partial biological oxidation that occurred in the AH reactor allowed the recovery of 30% of the sulphur load as S 0 . The effluent from the micro-aerated UASB reactor contained 5% of the sulphur load in the form of S 2- , while 20.9% was present as dissolved SO 4 2- and 46% was precipitated as S 0 . It is concluded that the AH reactor or micro-aeration carried out above the digestion zone of the UASB reactor favoured the biological oxidation of S 2- and the release of odourless effluents. Both technologies represent feasible and low-cost alternatives for the anaerobic treatment of domestic sewage.

Radiochemistry Lab Decommissioning and Dismantlement. AECL, Chalk River Labs, Ontario, Canada

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

Kenny, Stephen

2008-01-15

Atomic Energy of Canada (AECL) was originally founded in the mid 1940's to perform research in radiation and nuclear areas under the Canadian Defense Department. In the mid 50's The Canadian government embarked on several research and development programs for the development of the Candu Reactor. AECL was initially built as a temporary site and is now faced with many redundant buildings. Prior to 2004 small amounts of Decommissioning work was in progress. Many reasons for deferring decommissioning activities were used with the predominant ones being: 1. Reduction in radiation doses to workers during the final dismantlement, 2. Development ofmore » a long-term solution for the management of radioactive wastes in Canada, 3. Financial constraints presented by the number of facilities shutdown that would require decommissioning funds and the absence of an approved funding strategy. This has led to the development of a comprehensive decommissioning plan that is all inclusive of AECL's current and legacy liabilities. Canada does not have a long-term disposal site; therefore waste minimization becomes the driving factor behind decontamination for decommissioning before and during dismantlement. This decommissioning job was a great learning experience for decommissioning and the associated contractors who worked on this project. Throughout the life of the project there was a constant focus on waste minimization. This focus was constantly in conflict with regulatory compliance primarily with respect to fire regulations and protecting the facility along with adjacent facilities during the decommissioning activities. Discrepancies in historical documents forced the project to treat every space as a contaminated space until proven differently. Decommissioning and dismantlement within an operating site adds to the complexity of the tasks especially when it is being conducted in the heart of the plant. This project was very successful with no lost time accidents in over one hundred thousand hours worked, on schedule and under budget despite some significant changes throughout the decommissioning phases. The actual cost to decommission this building will come in under 9 million dollars vs. an estimated 14.5 million dollars. This paper will cover some of the unique aspects of dismantling a radioactive building that has seen pretty much every element of the periodic table pass through it with the client requirement focused on minimization of radioactive waste volumes.« less

Biodegradation of Methylene Blue Dye by Sequential Treatment Using Anaerobic Hybrid Reactor and Submerged Aerobic Fixed Film Bioreactor

NASA Astrophysics Data System (ADS)

Farooqi, Izharul H.; Basheer, Farrukh; Tiwari, Pradeepika

2017-12-01

Laboratory scale experiments were carried out to access the feasibility of sequential anaerobic/aerobic biological treatment for the biodegradation of Methylene Blue (MB) dye. Anaerobic studies were performed using anaerobic hybrid reactor (consisting of UASB and Anaerobic filter) whereas submerged aerobic fixed film reactor was used as aerobic reactor. Degradation of MB dye was attempted using neutralized acetic acid (1000 mg/L) as co-substrate. MB dye concentration was stepwise increased from 10 to 70 mg/L after reaching steady state in each dye concentration. Such a gradual increase in the dye concentration helps in the proper acclimatization of the sludge to dyes thereby avoiding the possible inhibitory effects to biological activities at high dye concentrations. The overall treatment efficiency of MB through sequential anaerobic-aerobic reactor operation was 90% at maximum attempted dye concentration of 70 mg/L. The effluent from anaerobic reactor was analysed for intermediate biodegradation products through HPLC. It was observed that catechol, quinone, amino pyrine, 1,4 diamino benzene were present. However they were absent in final effluent.

Waste tyre pyrolysis: modelling of a moving bed reactor.

PubMed

Aylón, E; Fernández-Colino, A; Murillo, R; Grasa, G; Navarro, M V; García, T; Mastral, A M

2010-12-01

This paper describes the development of a new model for waste tyre pyrolysis in a moving bed reactor. This model comprises three different sub-models: a kinetic sub-model that predicts solid conversion in terms of reaction time and temperature, a heat transfer sub-model that calculates the temperature profile inside the particle and the energy flux from the surroundings to the tyre particles and, finally, a hydrodynamic model that predicts the solid flow pattern inside the reactor. These three sub-models have been integrated in order to develop a comprehensive reactor model. Experimental results were obtained in a continuous moving bed reactor and used to validate model predictions, with good approximation achieved between the experimental and simulated results. In addition, a parametric study of the model was carried out, which showed that tyre particle heating is clearly faster than average particle residence time inside the reactor. Therefore, this fast particle heating together with fast reaction kinetics enables total solid conversion to be achieved in this system in accordance with the predictive model. Copyright © 2010 Elsevier Ltd. All rights reserved.

POWER-BURST FACILITY (PBF) CONCEPTUAL DESIGN

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

Wasserman, A.A.; Johnson, S.O.; Heffner, R.E.

1963-06-21

A description is presented of the conceptual design of a high- performance, pulsed reactor called the Power Burst Facility (PBF). This reactor is designed to generate power bursts with initial asymptotic periods as short as 1 msec, producing energy releases large enough to destroy entire fuel subassemblies placed in a capsule or flow loop mounted in the reactor, all without damage to the reactor itself. It will be used primarily to evaluate the consequences and hazards of very rapid destructive accidents in reactors representing the entire range of current nuclear technology as applied to power generation, propulsion, and testing. Itmore » will also be used to carry out detailed studies of nondestructive reactivity feedback mechanisms in the shortperiod domain. The facility was designed to be sufficiently flexible to accommodate future cores of even more advanced design. The design for the first reactor core is based upon proven technology; hence, completion of the final design of this core will involve no significant development delays. Construction of the PBF is proposed to begin in September 1984, and is expected to take approximately 20 months to complete. (auth)« less

The Effect of Stochastic Perturbation of Fuel Distribution on the Criticality of a One Speed Reactor and the Development of Multi-Material Multinomial Line Statistics

NASA Technical Reports Server (NTRS)

Jahshan, S. N.; Singleterry, R. C.

2001-01-01

The effect of random fuel redistribution on the eigenvalue of a one-speed reactor is investigated. An ensemble of such reactors that are identical to a homogeneous reference critical reactor except for the fissile isotope density distribution is constructed such that it meets a set of well-posed redistribution requirements. The average eigenvalue, , is evaluated when the total fissile loading per ensemble element, or realization, is conserved. The perturbation is proven to increase the reactor criticality on average when it is uniformly distributed. The various causes of the change in reactivity, and their relative effects are identified and ranked. From this, a path towards identifying the causes. and relative effects of reactivity fluctuations for the energy dependent problem is pointed to. The perturbation method of using multinomial distributions for representing the perturbed reactor is developed. This method has some advantages that can be of use in other stochastic problems. Finally, some of the features of this perturbation problem are related to other techniques that have been used for addressing similar problems.

GUM Analysis for SIMS Isotopic Ratios in BEP0 Graphite Qualification Samples, Round 2

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

Gerlach, David C.; Heasler, Patrick G.; Reid, Bruce D.

2009-01-01

This report describes GUM calculations for TIMS and SIMS isotopic ratio measurements of reactor graphite samples. These isotopic ratios are used to estimate reactor burn-up, and currently consist of various ratios of U, Pu, and Boron impurities in the graphite samples. The GUM calculation is a propagation of error methodology that assigns uncertainties (in the form of standard error and confidence bound) to the final estimates.

Diagnosing ion-beam targets, data acquisition, reactor conditions

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

Mendel, Jr., C. W.

1982-01-01

The final lecture will discuss diagnostics of the target. These are very difficult because of the short times, small spatial extent, and extreme values of temperature and pressure. Diagnostics for temperature, density profile, and neutron production will be discussed. A few minutes will be devoted to data acquisition needs. The lecture will end with a discussion of current areas where improvements are needed and future diagnostics that will be required for reactor conditions.

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

Venkataraman, M.; Natarajan, R.; Raj, Baldev

The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR)more » spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)« less

Hydrodynamic study of an internal airlift reactor for microalgae culture.

PubMed

Rengel, Ana; Zoughaib, Assaad; Dron, Dominique; Clodic, Denis

2012-01-01

Internal airlift reactors are closed systems considered today for microalgae cultivation. Several works have studied their hydrodynamics but based on important solid concentrations, not with biomass concentrations usually found in microalgae cultures. In this study, an internal airlift reactor has been built and tested in order to clarify the hydrodynamics of this system, based on microalgae typical concentrations. A model is proposed taking into account the variation of air bubble velocity according to volumetric air flow rate injected into the system. A relationship between riser and downcomer gas holdups is established, which varied slightly with solids concentrations. The repartition of solids along the reactor resulted to be homogenous for the range of concentrations and volumetric air flow rate studied here. Liquid velocities increase with volumetric air flow rate, and they vary slightly when solids are added to the system. Finally, liquid circulation time found in each section of the reactor is in concordance with those employed in microalgae culture.

Co-cultivation of Lactobacillus zeae and Veillonella cricetifor the production of propionic acid

PubMed Central

2013-01-01

In this work a defined co-culture of the lactic acid bacterium Lactobacillus zeae and the propionate producer Veillonella criceti has been studied in continuous stirred tank reactor (CSTR) and in a dialysis membrane reactor. It is the first time that this reactor type is used for a defined co-culture fermentation. This reactor allows high mixing rates and working with high cell densities, making it ideal for co-culture investigations. In CSTR experiments the co-culture showed over a broad concentration range an almost linear correlation in consumption and production rates to the supply with complex nutrients. In CSTR and dialysis cultures a strong growth stimulation of L. zeae by V. criceti was shown. In dialysis cultures very high propionate production rates (0.61 g L-1h-1) with final titers up to 28 g L-1 have been realized. This reactor allows an individual, intracellular investigation of the co-culture partners by omic-technologies to provide a better understanding of microbial communities. PMID:23705662

ENGINEERING AND CONSTRUCTING THE HALLAM NUCLEAR POWER FACILITY REACTOR STRUCTURE

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

Mahlmeister, J E; Haberer, W V; Casey, D F

1960-12-15

The Hallam Nuclear Power Facility reactor structure, including the cavity liner, is described, and the design philosophy and special design requirements which were developed during the preliminary and final engineering phases of the project are explained. The structure was designed for 600 deg F inlet and 1000 deg F outlet operating sodium temperatures and fabricated of austenitic and ferritic stainless steels. Support for the reactor core components and adequate containment for biological safeguards were readily provided even though quite conservative design philosophy was used. The calculated operating characteristics, including heat generation, temperature distributions and stress levels for full-power operation, aremore » summarized. Ship fabrication and field installation experiences are also briefly related. Results of this project have established that the sodium graphite reactor permits practical and economical fabrication and field erection procedures; considerably higher operating design temperatures are believed possible without radical design changes. Also, larger reactor structures can be similarly constructed for higher capacity (300 to 1000 Mwe) nuclear power plants. (auth)« less

Successive and large-scale synthesis of InP/ZnS quantum dots in a hybrid reactor and their application to white LEDs

NASA Astrophysics Data System (ADS)

Kim, Kyungnam; Jeong, Sohee; Woo, Ju Yeon; Han, Chang-Soo

2012-02-01

We report successive and large-scale synthesis of InP/ZnS core/shell nanocrystal quantum dots (QDs) using a customized hybrid flow reactor, which is based on serial combination of a batch-type mixer and a flow-type furnace. InP cores and InP/ZnS core/shell QDs were successively synthesized in the hybrid reactor in a simple one-step process. In this reactor, the flow rate of the solutions was typically 1 ml min-1, 100 times larger than that of conventional microfluidic reactors. In order to synthesize high-quality InP/ZnS QDs, we controlled both the flow rate and the crystal growth temperature. Finally, we obtained high-quality InP/ZnS QDs in colors from bluish green to red, and we demonstrated that these core/shell QDs could be incorporated into white-light-emitting diode (LED) devices to improve color rendering performance.

Successive and large-scale synthesis of InP/ZnS quantum dots in a hybrid reactor and their application to white LEDs.

PubMed

Kim, Kyungnam; Jeong, Sohee; Woo, Ju Yeon; Han, Chang-Soo

2012-02-17

We report successive and large-scale synthesis of InP/ZnS core/shell nanocrystal quantum dots (QDs) using a customized hybrid flow reactor, which is based on serial combination of a batch-type mixer and a flow-type furnace. InP cores and InP/ZnS core/shell QDs were successively synthesized in the hybrid reactor in a simple one-step process. In this reactor, the flow rate of the solutions was typically 1 ml min(-1), 100 times larger than that of conventional microfluidic reactors. In order to synthesize high-quality InP/ZnS QDs, we controlled both the flow rate and the crystal growth temperature. Finally, we obtained high-quality InP/ZnS QDs in colors from bluish green to red, and we demonstrated that these core/shell QDs could be incorporated into white-light-emitting diode (LED) devices to improve color rendering performance.

Anaerobic digestion of pressed off leachate from the organic fraction of municipal solid waste

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

Nayono, Satoto E.; Institute of Biology for Engineers and Biotechnology of Wastewater, University of Karlsruhe, Am Fasanengarten, 76131 Karlsruhe; Winter, Josef, E-mail: josef.winter@iba.uka.d

2010-10-15

A highly polluted liquid ('press water') was obtained from the pressing facility for the organic fraction of municipal solid waste in a composting plant. Methane productivity of the squeezed-off leachate was investigated in batch assays. To assess the technical feasibility of 'press water' as a substrate for anaerobic digestion, a laboratory-scale glass column reactor was operated semi-continuously at 37 {sup o}C. A high methane productivity of 270 m{sup -3} CH{sub 4} ton{sup -1} COD{sub added} or 490 m{sup -3} CH{sub 4} ton{sup -1} VS{sub added} was achieved in the batch experiment. The semi-continuously run laboratory-scale reactor was initially operated atmore » an organic loading rate of 10.7 kg COD m{sup -3} d{sup -1}. The loading was increased to finally 27.7 kg COD m{sup -3} d{sup -1}, corresponding to a reduction of the hydraulic retention time from initially 20 to finally 7.7 days. During the digestion, a stable elimination of organic material (measured as COD elimination) of approximately 60% was achieved. Linearly with the increment of the OLR, the volumetric methane production of the reactor increased from 2.6 m{sup 3} m{sub reactor}{sup -3} d{sup -1} to 7.1 m{sup 3} m{sub reactor}{sup -3} d{sup -1}. The results indicated that 'press water' from the organic fraction of municipal solid waste was a suitable substrate for anaerobic digestion which gave a high biogas yield even at very high loading rates.« less

RADSOURCE. Volume 1, Part 1, A scaling factor prediction computer program technical manual and code validation: Final report

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

Vance, J.N.; Holderness, J.H.; James, D.W.

1992-12-01

Waste stream scaling factors based on sampling programs are vulnerable to one or more of the following factors: sample representativeness, analytic accuracy, and measurement sensitivity. As an alternative to sample analyses or as a verification of the sampling results, this project proposes the use of the RADSOURCE code, which accounts for the release of fuel-source radionuclides. Once the release rates of these nuclides from fuel are known, the code develops scaling factors for waste streams based on easily measured Cobalt-60 (Co-60) and Cesium-137 (Cs-137). The project team developed mathematical models to account for the appearance rate of 10CFR61 radionuclides inmore » reactor coolant. They based these models on the chemistry and nuclear physics of the radionuclides involved. Next, they incorporated the models into a computer code that calculates plant waste stream scaling factors based on reactor coolant gamma- isotopic data. Finally, the team performed special sampling at 17 reactors to validate the models in the RADSOURCE code.« less

Optimization of post-column reactor radius in capillary high performance liquid chromatography Effect of chromatographic column diameter and particle diameter

PubMed Central

Xu, Hongjuan; Weber, Stephen G.

2006-01-01

A post-column reactor consisting of a simple open tube (Capillary Taylor Reactor) affects the performance of a capillary LC in two ways: stealing pressure from the column and adding band spreading. The former is a problem for very small radius reactors, while the latter shows itself for large reactor diameters. We derived an equation that defines the observed number of theoretical plates (Nobs) taking into account the two effects stated above. Making some assumptions and asserting certain conditions led to a final equation with a limited number of variables, namely chromatographic column radius, reactor radius and chromatographic particle diameter. The assumptions and conditions are that the van Deemter equation applies, the mass transfer limitation is for intraparticle diffusion in spherical particles, the velocity is at the optimum, the analyte’s retention factor, k′, is zero, the post-column reactor is only long enough to allow complete mixing of reagents and analytes and the maximum operating pressure of the pumping system is used. Optimal ranges of the reactor radius (ar) are obtained by comparing the number of observed theoretical plates (and theoretical plates per time) with and without a reactor. Results show that the acceptable reactor radii depend on column diameter, particle diameter, and maximum available pressure. Optimal ranges of ar become narrower as column diameter increases, particle diameter decreases or the maximum pressure is decreased. When the available pressure is 4000 psi, a Capillary Taylor Reactor with 12 μm radius is suitable for all columns smaller than 150 μm (radius) packed with 2–5 μm particles. For 1 μm packing particles, only columns smaller than 42.5 μm (radius) can be used and the reactor radius needs to be 5 μm. PMID:16494886

Biodegradation of endocrine disruptors in urban wastewater using Pleurotus ostreatus bioreactor.

PubMed

Křesinová, Zdena; Linhartová, Lucie; Filipová, Alena; Ezechiáš, Martin; Mašín, Pavel; Cajthaml, Tomáš

2018-07-25

The white rot fungus Pleurotus ostreatus HK 35, which is also an edible industrial mushroom commonly cultivated in farms, was tested in the degradation of typical representatives of endocrine disrupters (EDCs; bisphenol A, estrone, 17β-estradiol, estriol, 17α-ethinylestradiol, triclosan and 4-n-nonylphenol); its degradation efficiency under model laboratory conditions was greater than 90% within 12 days and better than that of another published strain P. ostreatus 3004. A spent mushroom substrate from a local farm was tested for its applicability in various batch and trickle-bed reactors in degrading EDCs in model fortified and real communal wastewater. The reactors were tested under various regimes including a pilot-scale trickle-bed reactor, which was finally tested at a wastewater treatment plant. The result revealed that the spent substrate is an efficient biodegradation agent, where the fungus was usually able to remove about 95% of EDCs together with suppression of the estrogenic activity of the sample. The results showed the fungus was able to operate in the presence of bacterial microflora in wastewater without any substantial negative effects on the degradation abilities. Finally, a pilot-scale trickle-bed reactor was installed in a wastewater treatment plant and successfully operated for 10days, where the bioreactor was able to remove more than 76% of EDCs present in the wastewater. Copyright © 2017 Elsevier B.V. All rights reserved.

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...

Method and apparatus for producing synthesis gas

DOEpatents

Hemmings, John William; Bonnell, Leo; Robinson, Earl T.

2010-03-03

A method and apparatus for reacting a hydrocarbon containing feed stream by steam methane reforming reactions to form a synthesis gas. The hydrocarbon containing feed is reacted within a reactor having stages in which the final stage from which a synthesis gas is discharged incorporates expensive high temperature materials such as oxide dispersed strengthened metals while upstream stages operate at a lower temperature allowing the use of more conventional high temperature alloys. Each of the reactor stages incorporate reactor elements having one or more separation zones to separate oxygen from an oxygen containing feed to support combustion of a fuel within adjacent combustion zones, thereby to generate heat to support the endothermic steam methane reforming reactions.

Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

2010-07-01

Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor.

PubMed

Essadki, A H; Gourich, B; Vial, Ch; Delmas, H; Bennajah, M

2009-09-15

Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12mA/cm(2), but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H(2) microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

Utilization of 134Cs/137Cs in the environment to identify the reactor units that caused atmospheric releases during the Fukushima Daiichi accident

NASA Astrophysics Data System (ADS)

Chino, Masamichi; Terada, Hiroaki; Nagai, Haruyasu; Katata, Genki; Mikami, Satoshi; Torii, Tatsuo; Saito, Kimiaki; Nishizawa, Yukiyasu

2016-08-01

The Fukushima Daiichi nuclear power reactor units that generated large amounts of airborne discharges during the period of March 12-21, 2011 were identified individually by analyzing the combination of measured 134Cs/137Cs depositions on ground surfaces and atmospheric transport and deposition simulations. Because the values of 134Cs/137Cs are different in reactor units owing to fuel burnup differences, the 134Cs/137Cs ratio measured in the environment was used to determine which reactor unit ultimately contaminated a specific area. Atmospheric dispersion model simulations were used for predicting specific areas contaminated by each dominant release. Finally, by comparing the results from both sources, the specific reactor units that yielded the most dominant atmospheric release quantities could be determined. The major source reactor units were Unit 1 in the afternoon of March 12, 2011, Unit 2 during the period from the late night of March 14 to the morning of March 15, 2011. These results corresponded to those assumed in our previous source term estimation studies. Furthermore, new findings suggested that the major source reactors from the evening of March 15, 2011 were Units 2 and 3 and that the dominant source reactor on March 20, 2011 temporally changed from Unit 3 to Unit 2.

Improving biogas production from continuous co-digestion of oily wastewater and waste-activated sludge by hydrodynamic cavitation pre-treatment.

PubMed

Habashi, Nima; Alighardashi, Abolghasem; Mennerich, Artur; Mehrdadi, Nasser; Torabian, Ali

2018-04-01

Hydrodynamic cavitation (HC) was evaluated as a pretreatment for synthetic oily wastewater (OWW) to be co-digested with waste-activated sludge (WAS). The main objective of the present research was the enhancement of biogas production by the application of HC pretreatment. HC was applied to the OWW, and the OWW and WAS were added to a 50 L continuous digestion reactor. As a control system, an identical digestion reactor was set up for co-digestion of the WAS and the OWW without pretreatment. The reactors were initially filled with inoculum and the hydraulic retention time (HRT) was set to 22 d. The HRT was gradually reduced to 19, 16, and finally 13 d, but the substrate quality was kept constant. The loading rate, accordingly, increased from 0.86 to 1.46 g TVS/(L d). The biogas volume was recorded online and its quality was analyzed regularly. The HC improved biogas production up to 43% at 22 d of HRT. Reducing the HRT decreased biogas production from the main reactor while that of the control reactor was more or less constant. HC also increased the biogas methane content; the methane concentration of the main reactor was about 3% higher than the methane concentration of the control reactor. The main reactor experienced no clogging or accumulation of fatty materials.

Utilization of (134)Cs/(137)Cs in the environment to identify the reactor units that caused atmospheric releases during the Fukushima Daiichi accident.

PubMed

Chino, Masamichi; Terada, Hiroaki; Nagai, Haruyasu; Katata, Genki; Mikami, Satoshi; Torii, Tatsuo; Saito, Kimiaki; Nishizawa, Yukiyasu

2016-08-22

The Fukushima Daiichi nuclear power reactor units that generated large amounts of airborne discharges during the period of March 12-21, 2011 were identified individually by analyzing the combination of measured (134)Cs/(137)Cs depositions on ground surfaces and atmospheric transport and deposition simulations. Because the values of (134)Cs/(137)Cs are different in reactor units owing to fuel burnup differences, the (134)Cs/(137)Cs ratio measured in the environment was used to determine which reactor unit ultimately contaminated a specific area. Atmospheric dispersion model simulations were used for predicting specific areas contaminated by each dominant release. Finally, by comparing the results from both sources, the specific reactor units that yielded the most dominant atmospheric release quantities could be determined. The major source reactor units were Unit 1 in the afternoon of March 12, 2011, Unit 2 during the period from the late night of March 14 to the morning of March 15, 2011. These results corresponded to those assumed in our previous source term estimation studies. Furthermore, new findings suggested that the major source reactors from the evening of March 15, 2011 were Units 2 and 3 and that the dominant source reactor on March 20, 2011 temporally changed from Unit 3 to Unit 2.

Utilization of 134Cs/137Cs in the environment to identify the reactor units that caused atmospheric releases during the Fukushima Daiichi accident

PubMed Central

Chino, Masamichi; Terada, Hiroaki; Nagai, Haruyasu; Katata, Genki; Mikami, Satoshi; Torii, Tatsuo; Saito, Kimiaki; Nishizawa, Yukiyasu

2016-01-01

The Fukushima Daiichi nuclear power reactor units that generated large amounts of airborne discharges during the period of March 12–21, 2011 were identified individually by analyzing the combination of measured 134Cs/137Cs depositions on ground surfaces and atmospheric transport and deposition simulations. Because the values of 134Cs/137Cs are different in reactor units owing to fuel burnup differences, the 134Cs/137Cs ratio measured in the environment was used to determine which reactor unit ultimately contaminated a specific area. Atmospheric dispersion model simulations were used for predicting specific areas contaminated by each dominant release. Finally, by comparing the results from both sources, the specific reactor units that yielded the most dominant atmospheric release quantities could be determined. The major source reactor units were Unit 1 in the afternoon of March 12, 2011, Unit 2 during the period from the late night of March 14 to the morning of March 15, 2011. These results corresponded to those assumed in our previous source term estimation studies. Furthermore, new findings suggested that the major source reactors from the evening of March 15, 2011 were Units 2 and 3 and that the dominant source reactor on March 20, 2011 temporally changed from Unit 3 to Unit 2. PMID:27546490

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

Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy researchmore » and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.« less

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

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

John Darrell Bess

2008-06-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, suchmore » as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.« less

Features of postfailure fuel behavior in transient overpower and transient undercooled/overpower tests in the transient reactor test facility

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

Doerner, R.C.; Bauer, T.H.; Morman, J.A.

Prototypic oxide fuel was subjected to simulated, fast reactor severe accident conditions in a series of in-pile tests in the Transient Reactor Test Facility reactor. Seven experiments were performed on fresh and previously irradiated oxide fuel pins under transient overpower and transient undercooled. overpower accident conditions. For each of the tests, fuel motions were observed by the hodoscope. Hodoscope data are correlated with coolant flow, pressure, and temperature data recorded by the loop instrumentation. Data were analyzed from the onset of initial failure to a final mass distribution at the end of the test. In this paper results of thesemore » analyses are compared to pre- and posttest accident calculations and to posttest metallographic accident calculations and to posttest metallographic examinations and computed tomographic reconstructions from neutron radiographs.« less

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

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

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D&D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D&D plans for the turbine building were prepared from 1979 through 1990. D&D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and the absence of loosemore » contamination, the D&D activities were completed with no radiation exposure to the workers. The D&D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

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

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D D plans for the turbine building were prepared from 1979 through 1990. D D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and themore » absence of loose contamination, the D D activities were completed with no radiation exposure to the workers. The D D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Production of edible carbohydrates from formaldehyde in a spacecraft. pH variations in the calcium hydroxide catalyzed formose reaction. Final Report, 1 Jul. 1973 - 30 Jun. 1974. M.S. Thesis

NASA Technical Reports Server (NTRS)

Weiss, A. H.; Kohler, J. T.; John, T.

1974-01-01

The study of the calcium hydroxide catalyzed condensation of formaldehyde was extended to a batch reactor system. Decreases in pH were observed, often in the acid regime, when using this basic catalyst. This observation was shown to be similar to results obtained by others using less basic catalysts in the batch mode. The relative rates of these reactions are different in a batch reactor than in a continuous stirred tank reactor. This difference in relative rates is due to the fact that at any degree of advancement in the batch system, the products have a history of previous products, pH, and dissolved catalyst. The relative rate differences can be expected to yield a different nature of product sugars for the two types of reactors.

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.

A beta-ray spectrometer based on a two-or three silicon detector coincidence telescope

NASA Astrophysics Data System (ADS)

Horowitz, Y. S.; Weizman, Y.; Hirning, C. R.

1996-02-01

This report describes the operation of a beta-ray energy spectrometer based on a silicon detector telescope using two or three elements. The front detector is a planar, totally-depleted, silicon surface barrier detector that is 97 μm thick, the back detector is a room-temperature, lithium compensated, silicon detector that is 5000 μm thick, and the intermediate detector is similar to the front detector but 72 μm thick and intended to be used only in intense photon fields. The three detectors are mounted in a light-tight aluminum housing. The capability of the spectrometer to reject photons is based upon the fact that the incident photon will have a small probability of simultaneously losing detectable energy in two detectors, and an even smaller probability of losing detectable energy in all three detectors. Electrons will, however, almost always record measurable events in either the front two or all three detectors. A coincidence requirement between the detectors thus rejects photon induced events. With a 97 μm thick detector the lower energy coincidence threshold is approximately 110 keV. With an ultra-thin 40 μm thick front detector, and operated at 15°C, the spectrometer is capable of detecting even 60-70 keV electrons with a coincidence efficiency of 60%. The spectrometer has been used to measure beta radiation fields in CANDU reactor working environments, and the spectral information is intended to support dose algorithms for the LiF TLD chips used in the Ontario Hydro dosimetry program.

TEMPEST code simulations of hydrogen distribution in reactor containment structures. Final report

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

Trent, D.S.; Eyler, L.L.

The mass transport version of the TEMPEST computer code was used to simulate hydrogen distribution in geometric configurations relevant to reactor containment structures. Predicted results of Battelle-Frankfurt hydrogen distribution tests 1 to 6, and 12 are presented. Agreement between predictions and experimental data is good. Best agreement is obtained using the k-epsilon turbulence model in TEMPEST in flow cases where turbulent diffusion and stable stratification are dominant mechanisms affecting transport. The code's general analysis capabilities are summarized.

Computational Modeling as a Design Tool in Microelectronics Manufacturing

NASA Technical Reports Server (NTRS)

Meyyappan, Meyya; Arnold, James O. (Technical Monitor)

1997-01-01

Plans to introduce pilot lines or fabs for 300 mm processing are in progress. The IC technology is simultaneously moving towards 0.25/0.18 micron. The convergence of these two trends places unprecedented stringent demands on processes and equipments. More than ever, computational modeling is called upon to play a complementary role in equipment and process design. The pace in hardware/process development needs a matching pace in software development: an aggressive move towards developing "virtual reactors" is desirable and essential to reduce design cycle and costs. This goal has three elements: reactor scale model, feature level model, and database of physical/chemical properties. With these elements coupled, the complete model should function as a design aid in a CAD environment. This talk would aim at the description of various elements. At the reactor level, continuum, DSMC(or particle) and hybrid models will be discussed and compared using examples of plasma and thermal process simulations. In microtopography evolution, approaches such as level set methods compete with conventional geometric models. Regardless of the approach, the reliance on empricism is to be eliminated through coupling to reactor model and computational surface science. This coupling poses challenging issues of orders of magnitude variation in length and time scales. Finally, database development has fallen behind; current situation is rapidly aggravated by the ever newer chemistries emerging to meet process metrics. The virtual reactor would be a useless concept without an accompanying reliable database that consists of: thermal reaction pathways and rate constants, electron-molecule cross sections, thermochemical properties, transport properties, and finally, surface data on the interaction of radicals, atoms and ions with various surfaces. Large scale computational chemistry efforts are critical as experiments alone cannot meet database needs due to the difficulties associated with such controlled experiments and costs.

Hydrogel coated monoliths for enzymatic hydrolysis of penicillin G

PubMed Central

Smeltink, M. W.; Straathof, A. J. J.; Paasman, M. A.; van de Sandt, E. J. A. X.; Kapteijn, F.; Moulijn, J. A.

2008-01-01

The objective of this work was to develop a hydrogel-coated monolith for the entrapment of penicillin G acylase (E. coli, PGA). After screening of different hydrogels, chitosan was chosen as the carrier material for the preparation of monolithic biocatalysts. This protocol leads to active immobilized biocatalysts for the enzymatic hydrolysis of penicillin G (PenG). The monolithic biocatalyst was tested in a monolith loop reactor (MLR) and compared with conventional reactor systems using free PGA, and a commercially available immobilized PGA. The optimal immobilization protocol was found to be 5 g l−1 PGA, 1% chitosan, 1.1% glutaraldehyde and pH 7. Final PGA loading on glass plates was 29 mg ml−1 gel. For 400 cpsi monoliths, the final PGA loading on functionalized monoliths was 36 mg ml−1 gel. The observed volumetric reaction rate in the MLR was 0.79 mol s−1 m−3monolith. Apart from an initial drop in activity due to wash out of PGA at higher ionic strength, no decrease in activity was observed after five subsequent activity test runs. The storage stability of the biocatalysts is at least a month without loss of activity. Although the monolithic biocatalyst as used in the MLR is still outperformed by the current industrial catalyst (immobilized preparation of PGA, 4.5 mol s−1 m−3catalyst), the rate per gel volume is slightly higher for monolithic catalysts. Good activity and improved mechanical strength make the monolithic bioreactor an interesting alternative that deserves further investigation for this application. Although moderate internal diffusion limitations have been observed inside the gel beads and in the gel layer on the monolith channel, this is not the main reason for the large differences in reactor performance that were observed. The pH drop over the reactor as a result of the chosen method for pH control results in a decreased performance of both the MLR and the packed bed reactor compared to the batch system. A different reactor configuration including an optimal pH profile is required to increase the reactor performance. The monolithic stirrer reactor would be an interesting alternative to improve the performance of the monolith-PGA combination. PMID:18427849

Two-phase anaerobic digestion of source sorted OFMSW (organic fraction of municipal solid waste): performance and kinetic study.

PubMed

Pavan, P; Battistoni, P; Cecchi, F; Mata-Alvarez, J

2000-01-01

The results of a two-phase system operated in different conditions, treating the source-sorted organic fraction of municipal solid waste (SS-OFMSW), coming mainly from fruit and vegetable markets, are presented. Hydraulic retention time (HRT) in the hydrolytic reactor and in the methanogenic reactor and also the temperature in the hydrolytic reactor (mesophilic and thermophilic conditions) are varied in order to evaluate the effect of these factors. The methanogenic reactor is always operated within the thermophilic range. Optimum operating conditions are found to be around 12 days (total system) using the mesophilic range of temperature in the first reactor. Specific gas production (SGP) in these conditions is around 0.6 m3/kg TVS. A kinetic study is also carried out, using the first and the step diffusional models. The latter gives much better results, with fitted constants comparable to other studies. Finally, a comparison with a one-phase system is carried out, showing that a two-phase system is much more appropriate for the digestion of this kind of highly biodegradable substrate in thermophilic conditions.

Production of oxalic acid from sugar beet molasses by formed nitrogen oxides.

PubMed

Gürü, M; Bilgesü, A Y; Pamuk, V

2001-03-01

Production of oxalic acid from sugar beet molasses was developed in a series of three reactors. Nitrogen oxides formed were used to manufacture oxalic acid in the second and third reactor. Parameters affecting the reaction were determined to be, air flow rate, temperature, the amount of V2O5 catalyst and the concentrations of molasses and H2SO4. The maximum yields in the second and third reactors were 78.9% and 74.6% of theoretical yield, respectively. Also, kinetic experiments were performed and the first-order rate constants were determined for the glucose consumption rate. Nitrogen oxides in off-gases from the final reactor were absorbed in water and concentrated sulphuric acid and reused in the following reactors giving slightly lower yields under similar conditions. In this novel way, it was possible to recover NO(x) and to prevent air pollution. Meanwhile, it was possible to reduce the unit cost of reactant for oxalic acid production. A maximum 77.5% and 74.1% of theoretical yield was obtained by using the absorption solutions with NO(x).

Improved measurement of the reactor antineutrino flux and spectrum at Daya Bay

DOE PAGES

An, F. P.; Balantekin, A. B.; Band, H. R.; ...

2017-01-01

Here, a new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay reactor neutrino experiment is reported. The antineutrinos were generated by six 2.9 GW th nuclear reactors and detected by eight antineutrino detectors deployed in two near (560 m and 600 m flux-weighted baselines) and one far (1640 m flux-weighted baseline) underground experimental halls. With 621 days of data, more than 1.2 million inverse beta decay (IBD) candidates were detected. The IBD yield in the eight detectors was measured, and the ratio of measured to predicted flux was found to be 0.946 ± 0.020 (0.992more » ± 0.021) for the Huber+Mueller (ILL+Vogel) model. A 2.9σ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4$-$6 MeV was found in the measured spectrum, with a local significance of 4.4σ. Finally, a reactor antineutrino spectrum weighted by the IBD cross section is extracted for model-independent predictions.« less

Review of heat transfer problems associated with magnetically-confined fusion reactor concepts

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

Hoffman, M.A.; Werner, R.W.; Carlson, G.A.

1976-04-01

Conceptual design studies of possible fusion reactor configurations have revealed a host of interesting and sometimes extremely difficult heat transfer problems. The general requirements imposed on the coolant system for heat removal of the thermonuclear power from the reactor are discussed. In particular, the constraints imposed by the fusion plasma, neutronics, structure and magnetic field environment are described with emphasis on those aspects which are unusual or unique to fusion reactors. Then the particular heat transfer characteristics of various possible coolants including lithium, flibe, boiling alkali metals, and helium are discussed in the context of these general fusion reactor requirements.more » Some specific areas where further experimental and/or theoretical work is necessary are listed for each coolant along with references to the pertinent research already accomplished. Specialized heat transfer problems of the plasma injection and removal systems are also described. Finally, the challenging heat transfer problems associated with the superconducting magnets are reviewed, and once again some of the key unsolved heat transfer problems are enumerated.« less

78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-19

... 100 Grams,'' dated October 10, 2002, and subsequent similar security orders issued to licensees... than 100 grams from theft, diversion, or radiological sabotage. The requirements in the final rule...

Osiris and SOMBRERO inertial confinement fusion power plant designs. Volume 2, Designs, assessments, and comparisons, Final report

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

Meier, W.R.; Bieri, R.L.; Monsler, M.J.

1992-03-01

The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of ourmore » effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs.« less

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Technical Reports Server (NTRS)

Martin, James; Salvail, Pat

2003-01-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final "wet in". A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/- 1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to less than10(exp -10) std cc/sec helium and vacuum conditioned at 250 C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Astrophysics Data System (ADS)

Martin, James; Salvail, Pat

2004-02-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final ``wet in''. A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/-1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to <10-10 std cc/sec helium and vacuum conditioned at 250 °C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 °C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

NPR design basis

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

Locke, G.L.

1958-09-08

The design basis is composed of requirements and conditions for the design of the reactor plant (composed of the reactor and heat dissipation system). Its intent is to insure that the final product meets the economic, safety, and technical objectives of the project. The design basis is dependent on the ground rules, objectives, technical criteria, and practical design considerations. This document is being issued with the understanding that these items are not yet firmly established in all respects, and therefore, the numbers put down here are subject to change. Consideration of the spectrum of probable changes that might be mademore » leads to the conclusion that the numbers here are close to the final ones and are satisfactory as a basis for the initial stages of design. Some numbers are omitted because of insufficient data at this time.« less

Maintenance method and its critical issues for a fast-ignition laser fusion reactor based on a dry wall chamber

NASA Astrophysics Data System (ADS)

Someya, Y.; Matsumoto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Goto, T.; Ogawa, Y.

2008-05-01

The neutronics analysis has been carried out for feasibility study of the FALCON-D concept by Monte Carlo N-paticle transport code (MCNP), in order to inspect the cooling performance of in-vessel and ex-vessel components, and a connection pipe between Vacuum Vessel and reactor room. The nuclear heating rate in the Vacuum Vessel was at the same level as that of NBI duct of the ITER. The temperature of the connection pipe was found to be 345·, ·which was smaller than the melting point of structure materials (F82H). Moreover, the radiation damage of the final optics was also investigated. We propose a sliding changer concept for replacement. This method could be adapted for the replacement of one FPY cycle in the final optics system.

Final summary report for 1989 inservice inspection (ISI) of SRS (Savannah River Site) 100-P Reactor tank

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

Morrison, J.M.; Loibl, M.W.

1989-12-15

The integrity of the SRS reactor tanks is a key factor affecting their suitability for continued service since, unlike the external piping system and components, the tanks are virtually irreplaceable. Cracking in various areas of the process water piping systems has occurred beginning in 1960 as a result of several degradation mechanisms, chiefly intergranular stress corrosion cracking (IGSCC) and chloride-induced transgranular cracking. IGSCC, currently the primary degradation mechanism, also occurred in the knuckle'' region (tank wall-to-bottom tube sheet transition piece) unique to C Reactor and was eventually responsible for that reactor being deactivated in 1985. A program of visual examinationsmore » of the SRS reactor tanks was initiated in 1968, which used a specially designed immersible periscope. Under that program the condition of the accessible tank welds and associated heat affected zones (HAZ) was evaluated on a five-year frequency. Prior to 1986, the scope of these inspections comprised approximately 20 percent of the accessible weld area. In late 1986 and early 1987 the scope of the inspections was expanded and a 100 percent visual inspection of accessible welds was performed of the P-, L-, and K-Reactor tanks. Supplemental dye penetrant examinations were performed in L Reactor on selected areas which showed visual indications. No evidence of cracking was detected in any of these inspections of the P-, L-, and K-Reactor tanks. 17 refs., 7 figs.« less

Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

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

Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A

2010-01-01

We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is inmore » support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to consolidate the locations of special nuclear material (SNM) to reduce the cost of securing many SNM facilities. The removal of SPR-II from SNL was a significant accomplishment in SNL's de-inventory efforts and played a key role in reducing the number of locations requiring the expensive security measures required for category 1 and 2 SNM facilities. A similar pulse reactor was fabricated at the Y-12 National Security Complex beginning in the late 1960's. This Aberdeen Pulse Reactor (APR) was operated at the Army Pulse Radiation Facility (APRF) located at the Aberdeen Test Center (ATC) in Maryland. When the APRF was shut down in 2003, a portion of the DOE-owned Special Nuclear Material (SNM) was shipped to an interim facility for storage. Subsequently, the DOE determined that the material from both the SPR-II and the APR would be processed in the H-Canyon at the Savannah River Site (SRS). Because of the SRS receipt requirements some of the material was sent to the Los Alamos National Laboratory (LANL) for size-reduction prior to shipment to the SRS for final disposition.« less

The PROSPECT physics program

DOE PAGES

Ashenfelter, J.; Balantekin, A. B.; Band, H. R.; ...

2016-10-17

The precision reactor oscillation and spectrum experiment, PROSPECT, is designed to make a precise measurement of the antineutrino spectrum from a highly-enriched uranium reactor and probe eV-scale sterile neutrinos by searching for neutrino oscillations over a distance of several meters. The subject of this paper, PROSPECT, is conceived as a 2-phase experiment utilizing segmented 6Li-doped liquid scintillator detectors for both efficient detection of reactor antineutrinos through the inverse beta decay reaction and excellent background discrimination. PROSPECT Phase I consists of a movable 3 ton antineutrino detector at distances of 7–12 m from the reactor core. It will probe the best-fitmore » point of the ν e disappearance experiments at 4σ in 1 year and the favored region of the sterile neutrino parameter space at > 3σ in 3 years. With a second antineutrino detector at 15–19 m from the reactor, Phase II of PROSPECT can probe the entire allowed parameter space below 10 eV 2 at 5σ in 3 additional years. Finally, the measurement of the reactor antineutrino spectrum and the search for short-baseline oscillations with PROSPECT will test the origin of the spectral deviations observed in recent θ 13 experiments, search for sterile neutrinos, and conclusively address the hypothesis of sterile neutrinos as an explanation of the reactor anomaly.« less

Effects of the Fabrication Process and Thermal Cycling on the Oxidation of Zirconium-Niobium Pressure Tubes

NASA Astrophysics Data System (ADS)

Nam, Cheol

2009-12-01

Pressure tubes made of Zr-2.5%Nb alloy are used to contain fuels and coolant in CANDU nuclear power reactors The pressure tube oxidizes during reactor operation and hydrogen ingress through the oxide grown on the tube limits its lifetime. Little attention was paid to the intermediate tube manufacturing processes in enhancing the oxidation resistance. In addition, the oxide grown on the tube experiences various thermal cycles depending on the reactor shutdown and startup cycles. To address these two aspects and to better understand the oxidation process of the Zr-2.5Nb tube, research was conducted in two parts: (i) effects of tube fabrication on oxidation behavior, and (ii) thermal cycling behaviors of oxides grown on a pressure tube. In the first part, the optimum manufacturing process was pursued to improve the corrosion resistance of Zr-2.5Nb tubes. Experimental micro-tubes were fabricated with various manufacturing routes in the stages of billet preparation, hot extrusion and cold drawing. These were oxidized in air at 400°C and 500°C, and in an autoclave at 360°C lithiated water. Microstructure and texture of the tubes and oxides were characterized with X-ray diffraction, scanning electron microscope and optical microscope. Special emphasis was given to examinations of the metal/oxide interface structures. A correlation between the manufacturing process and oxidation resistance was investigated in terms of tube microstructure and the metal/oxide interface structure. As a result, it was consistently observed that uniform interface structures were formed on the tubes which had a fine distribution of secondary phases. These microstructures were found to be beneficial in enhancing the oxidation resistance as opposed to the tubes that had coarse and continuous beta-Zr phases. Based on these observations, a schematic model of the oxidation process was proposed with respect to the oxidation resistance under oxidizing temperatures of 360°C, 400°C and 500°C. In the second part, the oxides grown on a standard Zr-2.5Nb pressure tube were analyzed by X-ray diffraction peak broadening and line shift. Crystallite size, t-ZrO2 fraction and residual stress of the zirconium oxides were investigated upon several thermal cycles at DeltaT range of 500°C--750°C. The oxide residual stresses measured by the sin2psi method were always compressive around 2 GPa. Different stress-states were noticed with the oxides grown on different sections of pressure tube. The compressive stress was released when the oxide was thermally cycled at the highest DeltaT of 750°C. Discussion was given to the effects of anisotropic nature of thermal expansion coefficients and crystallographic texture on the stress-state of Zr oxides.

75 FR 69709 - Office of New Reactors; Notice of Availability of the Final Staff Guidance; Standard Review Plan...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-15

... the Final Staff Guidance; Standard Review Plan, Section 13.6.6, Revision 0 on Cyber Security Plan... Reports for Nuclear Power Plants,'' Section 13.6.6, Revision 0 on ``Cyber Security Plan'' (Agencywide.... Disposition: On May 13, 2010, the NRC staff issued the proposed Revision 0 on SRP Section 13.6.6 on ``Cyber...

Project Luna Succendo: The Lunar Evolutionary Growth-Optimized (LEGO) Reactor

NASA Astrophysics Data System (ADS)

Bess, John Darrell

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched within lunar shipments from the Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides 5 kWe using a free-piston Stirling space converter. The overall envelope for a single unit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. The subunits can be placed with centerline distances of approximately 0.6 m in a hexagonal-lattice pattern to provide sufficient neutronic coupling while allowing room for heat rejection and interstitial control. A lattice of six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network Future improvements include advances in reactor control methods, fuel form and matrix, determination of shielding requirements, as well as power conversion and heat rejection techniques to generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces such as Mars, other moons, and asteroids.

Co-composting of palm oil mill sludge-sawdust.

PubMed

Yaser, Abu Zahrim; Abd Rahman, Rakmi; Kalil, Mohd Sahaid

2007-12-15

Composting of Palm Oil Mill Sludge (POMS) with sawdust was conducted in natural aerated reactor. Composting using natural aerated reactor is cheap and simple. The goal of this study is to observe the potential of composting process and utilizing compost as media for growing Cymbopogun citratus, one of Malaysia herbal plant. The highest maximum temperature achieved is about 40 degrees C and to increase temperature bed, more biodegradable substrate needs to be added. The pH value decrease along the process with final pH compost is acidic (pH 5.7). The highest maximum organic losses are about 50% with final C/N ratio of the compost is about 19. Final compost also showed some fertilizing value but need to be adjusted to obtain an ideal substrate. Addition of about 70% sandy soil causes highest yield and excellent root development for C. citratus in potted media. Beside that, compost from POMS-sawdust also found to have fertilizer value and easy to handle. Composting of POMS with sawdust shows potential as an alternative treatment to dispose and recycle waste components.

Design improvement and performance evaluation of solar photocatalytic reactor for industrial effluent treatment.

PubMed

Nair, Ranjith G; Bharadwaj, P J; Samdarshi, S K

2016-12-01

This work reports the details of the design components and materials used in a linear compound parabolic trough reactor constructed with an aim to use the photocatalyst for solar photocatalytic applications. A compound parabolic trough reactor has been designed and engineered to exploit both UV and visible part of the solar irradiation. The developed compound parabolic trough reactor could receive almost 88% of UV radiation along with a major part of visible radiation. The performance of the reactor has been evaluated in terms of degradation of a probe pollutant using the parameters such as rate constant, residence time and photonic efficiency. An attempt has been made to assess the performance in different ranges of solar spectrum. Finally the developed reactor has been employed for the photocatalytic treatment of a paper mill effluent using Degussa P25 as the photocatalyst. The paper mill effluent collected from Nagaon paper mill, Assam, India has been treated under both batch mode and continuous mode using Degussa P25 photocatalyst under artificial and natural solar radiation, respectively. The photocatalytic degradation kinetics of the paper mill effluent has been determined using the reduction in total organic carbon (TOC) values of the effluent. Copyright © 2015 Elsevier Inc. All rights reserved.

Different cultivation methods to acclimatise ammonia-tolerant methanogenic consortia.

PubMed

Tian, Hailin; Fotidis, Ioannis A; Mancini, Enrico; Angelidaki, Irini

2017-05-01

Bioaugmentation with ammonia tolerant-methanogenic consortia was proposed as a solution to overcome ammonia inhibition during anaerobic digestion process recently. However, appropriate technology to generate ammonia tolerant methanogenic consortia is still lacking. In this study, three basic reactors (i.e. batch, fed-batch and continuous stirred-tank reactors (CSTR)) operated at mesophilic (37°C) and thermophilic (55°C) conditions were assessed, based on methane production efficiency, incubation time, TAN/FAN (total ammonium nitrogen/free ammonia nitrogen) levels and maximum methanogenic activity. Overall, fed-batch cultivation was clearly the most efficient method compared to batch and CSTR. Specifically, by saving incubation time up to 150%, fed-batch reactors were acclimatised to nearly 2-fold higher FAN levels with a 37%-153% methanogenic activity improvement, compared to batch method. Meanwhile, CSTR reactors were inhibited at lower ammonia levels. Finally, specific methanogenic activity test showed that hydrogenotrophic methanogens were more active than aceticlastic methanogens in all FAN levels above 540mgNH 3 -NL -1 . Copyright © 2017 Elsevier Ltd. All rights reserved.

PCI fuel failure analysis: a report on a cooperative program undertaken by Pacific Northwest Laboratory and Chalk River Nuclear Laboratories.

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

Mohr, C.L.; Pankaskie, P.J.; Heasler, P.G.

Reactor fuel failure data sets in the form of initial power (P/sub i/), final power (P/sub f/), transient increase in power (..delta..P), and burnup (Bu) were obtained for pressurized heavy water reactors (PHWRs), boiling water reactors (BWRs), and pressurized water reactors (PWRs). These data sets were evaluated and used as the basis for developing two predictive fuel failure models, a graphical concept called the PCI-OGRAM, and a nonlinear regression based model called PROFIT. The PCI-OGRAM is an extension of the FUELOGRAM developed by AECL. It is based on a critical threshold concept for stress dependent stress corrosion cracking. The PROFITmore » model, developed at Pacific Northwest Laboratory, is the result of applying standard statistical regression methods to the available PCI fuel failure data and an analysis of the environmental and strain rate dependent stress-strain properties of the Zircaloy cladding.« less

The Nucifer Experiment

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

Cucoanes, A.S., E-mail: cucoanes@subatech.in2p3.fr

In nuclear reactors, a large number of antineutrinos are generated in the decay chains of the fission products; thus a survey of the antineutrino flux could provide valuable information related to the uranium and plutonium content of the core. This application generated interest by the IAEA in using antineutrino detectors as a potential safeguard tool. Here we present the Nucifer experiment, developed in France, by CEA and CNRS/IN2P3. The design of this new antineutrino detector has focused on safety, size reduction, reliability and high detection efficiency with a good background rejection. The Nucifer detector is currently taking data at themore » OSIRIS research reactor, inside CEA-Saclay. Presently, the ongoing analyses are considering the main sources of background for the antineutrino detection; the first antineutrino result is expected in 2013. A possible contribution to the understanding of the so called “reactor antineutrino anomaly” is also discussed. Finally, we present a brief description of the proposed experiments at very short baselines (VSBL) from reactors in France.« less

Determining Reactor Fuel Type from Continuous Antineutrino Monitoring

NASA Astrophysics Data System (ADS)

Jaffke, Patrick; Huber, Patrick

2017-09-01

We investigate the ability of an antineutrino detector to determine the fuel type of a reactor. A hypothetical 5-ton antineutrino detector is placed 25 m from the core and measures the spectral shape and rate of antineutrinos emitted by fission fragments in the core for a number of 90-d periods. Our results indicate that four major fuel types can be differentiated from the variation of fission fractions over the irradiation time with a true positive probability of detection at approximately 95%. In addition, we demonstrate that antineutrinos can identify the burnup at which weapons-grade mixed-oxide (MOX) fuel would be reduced to reactor-grade MOX, on average, providing assurance that plutonium-disposition goals are met. We also investigate removal scenarios where plutonium is purposefully diverted from a mixture of MOX and low-enriched uranium fuel. Finally, we discuss how our analysis is impacted by a spectral distortion around 6 MeV observed in the antineutrino spectrum measured from commercial power reactors.

Partial nitrification using aerobic granules in continuous-flow reactor: rapid startup.

PubMed

Wan, Chunli; Sun, Supu; Lee, Duu-Jong; Liu, Xiang; Wang, Li; Yang, Xue; Pan, Xiangliang

2013-08-01

This study applied a novel strategy to rapid startup of partial nitrification in continuous-flow reactor using aerobic granules. Mature aerobic granules were first cultivated in a sequencing batch reactor at high chemical oxygen demand in 16 days. The strains including the Pseudoxanthomonas mexicana strain were enriched in cultivated granules to enhance their structural stability. Then the cultivated granules were incubated in a continuous-flow reactor with influent chemical oxygen deamnad being stepped decreased from 1,500 ± 100 (0-19 days) to 750 ± 50 (20-30 days), and then to 350 ± 50 mg l(-1) (31-50 days); while in the final stage 350 mg l(-1) bicarbonate was also supplied. Using this strategy the ammonia-oxidizing bacterium, Nitrosomonas europaea, was enriched in the incubated granules to achieve partial nitrification efficiency of 85-90% since 36 days and onwards. The partial nitrification granules were successfully harvested after 52 days, a period much shorter than those reported in literature. Copyright © 2013 Elsevier Ltd. All rights reserved.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

DOE PAGES

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; ...

2014-06-30

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in ordermore » to reduce the overall numerical uncertainty while leveraging available computational resources. Finally, the coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.« less

Preliminary Tritium Management Design Activities at ORNL

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

Harrison, Thomas J.; Felde, David K.; Logsdon, Randall J.

2016-09-01

Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritiummore » mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year’s efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.« less

Final report, PT IP-535-C: Test of smaller VSR`s in DR reactor

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

Vaughn, A.D.

1963-04-17

Because of rod-sticking problems at DR Reactor, a knuckle rod of B Reactor design was installed in vertical safety channel number 28. The substitute VSR, which has a smaller diameter than the original DR rod, has been tested for its operational feasibility including both drop time and reactivity effect. The reactivity effect of the rod was estimated by comparison of the reactivity transient caused by insertion of the specific B-type rod after scramming into the pile, with similar transients caused by normal vertical safety rod in an adjacent channel. This document lists the indicated relative control strength of the rodmore » as an empirical basis for future safety calculations. Results indicate that the B-type knuckel rod in DR reactor is about 80% as strong as a normal DR vertical safety rod if used in equivalent flux distribution and location; this makes it reasonable to assume that the local control strength of the B-type knuckel rod is 98 {mu}b.« less

The startup performance and microbial distribution of an anaerobic baffled reactor (ABR) treating medium-strength synthetic industrial wastewater.

PubMed

Jiang, Hao; Nie, Hong; Ding, Jiangtao; Stinner, Walter; Sun, Kaixuan; Zhou, Hongjun

2018-01-02

In this study, an anaerobic baffled reactor (ABR) with seven chambers was applied to treat medium-strength synthetic industrial wastewater (MSIW). The performance of startup and shock test on treating MSIW was investigated. During the acclimation process, the chemical oxygen demand (COD) of MSIW gradually increased from 0 to 2,000 mg L -1 , and the COD removal finally reached 90%. At shock test, the feeding COD concentration increased by one-fifth and the reactor adapted very well with a COD removal of 82%. In a stable state, Comamonas, Smithella, Syntrophomonas and Pseudomonas were the main populations of bacteria, while the predominant methanogen was Methanobacterium. The results of chemical and microbiological analysis indicated the significant advantages of ABR, including buffering shocks, separating stages with matching microorganisms and promoting syntrophism. Meanwhile, the strategies for acclimation and operation were of great importance. Further work can test reactor performance in the treatment of actual industrial wastewater.

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

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

Yamada, K.; Aksan, S. N.

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present,more » 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)« less

Investigating the kinetics of the enzymatic depolymerization of polygalacturonic acid in continuous UF-membrane reactors.

PubMed

Gallifuoco, Alberto; Cantarella, Maria; Marucci, Mariagrazia

2007-01-01

A stirred tank membrane reactor is used to study the kinetics of polygalacturonic acid (PGA) enzymatic hydrolysis. The reactor operates in semicontinuous configuration: the native biopolymer is loaded at the initial time and the system is continuously fed with the buffer. The effect of retention time (from 101 to 142 min) and membrane molecular weight cutoff (from 1 to 30 kDa) on the rate of permeable oligomers production is investigated. Reaction products are clustered in two different classes, those sized below the membrane cutoff and those above. The reducing power measured in the permeate is used as an estimate of total product concentration. The characteristic breakdown times range from 40 to 100 min. The overall kinetics obeys a first-order law with a characteristic time estimated to 24 min. New mathematical data handling are developed and illustrated using the experimental data obtained. Finally, the body of the experimental results suggests useful indications (reactor productivity, breakdown induction period) for implementing the bioprocess at the industrial scale.

Study of the Open Loop and Closed Loop Oscillator Techniques

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

Imel, George R.; Baker, Benjamin; Riley, Tony

This report presents the progress and completion of a five-year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques.The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this report we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign tomore » measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems.« less

Study of the open loop and closed loop oscillator techniques

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

Baker, Benjamin; Riley, Tony; Langbehn, Adam

This paper presents some aspects of a five year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques. The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this paper we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign tomore » measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems. (authors)« less

Reactor antineutrino shoulder explained by energy scale nonlinearities?

NASA Astrophysics Data System (ADS)

Mention, G.; Vivier, M.; Gaffiot, J.; Lasserre, T.; Letourneau, A.; Materna, T.

2017-10-01

The Daya Bay, Double Chooz and RENO experiments recently observed a significant distortion in their detected reactor antineutrino spectra, being at odds with the current predictions. Although such a result suggests to revisit the current reactor antineutrino spectra modeling, an alternative scenario, which could potentially explain this anomaly, is explored in this letter. Using an appropriate statistical method, a study of the Daya Bay experiment energy scale is performed. While still being in agreement with the γ calibration data and 12B measured spectrum, it is shown that a O (1%) deviation of the energy scale reproduces the distortion observed in the Daya Bay spectrum, remaining within the quoted calibration uncertainties. Potential origins of such a deviation, which challenge the energy calibration of these detectors, are finally discussed.

Use of Gelidium amansii as a promising resource for bioethanol: a practical approach for continuous dilute-acid hydrolysis and fermentation.

PubMed

Park, Jeong-Hoon; Hong, Ji-Yeon; Jang, Hyun Chul; Oh, Seung Geun; Kim, Sang-Hyoun; Yoon, Jeong-Jun; Kim, Yong Jin

2012-03-01

A facile continuous method for dilute-acid hydrolysis of the representative red seaweed species, Gelidium amansii was developed and its hydrolysate was subsequently evaluated for fermentability. In the hydrolysis step, the hydrolysates obtained from a batch reactor and a continuous reactor were systematically compared based on fermentable sugar yield and inhibitor formation. There are many advantages to the continuous hydrolysis process. For example, the low melting point of the agar component in G. amansii facilitates improved raw material fluidity in the continuous reactor. In addition, the hydrolysate obtained from the continuous process delivered a high sugar and low inhibitor concentration, thereby leading to both high yield and high final ethanol titer in the fermentation process. Copyright © 2011 Elsevier Ltd. All rights reserved.

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

Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR).

PubMed

Tawfik, A; El-Kamah, H

2012-01-01

This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 degrees C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic retention time (HRT) of 10.2 h and organic loading rate (OLR) of 11.8 kg COD m(-3) d(-1). The reactor achieved a removal efficiency of 42% for chemical oxygen demand (COD), 50.8% for biochemical oxygen demand (BOD5), 50.3% for volatile fatty acids (VFA) and 56.4% for total suspended solids (TSS). In the second experiment, two AH reactors connected in series achieved a higher removal efficiency for COD (67.4%), BOD5 (77%), and TSS (71.5%) at a total HRT of 20 h and an OLR of 5.9 kg COD m(-3) d(-1). For removal of the remaining portions of COD, BOD5 and TSS from the effluent of the two-stage AH system, a sequencing batch reactor (SBR) was investigated as a post-treatment unit. The reactor achieved a substantial reduction in total COD, resulting in an average effluent concentration of 50 mg L(-1) at an HRT of 11 h and OLR of 5.3 kg COD m(-3) d(-1). Almost complete removal of total BOD5 and oil and grease was achieved, i.e. 10 mg L(-1) and 1.2 mg L(-1), respectively, remained in the final effluent of the SBR.

Decommissioning of eight surplus production reactors at the Hanford Site, Richland, Washington. Addendum (Final Environmental Impact Statement)

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

Not Available

1992-12-01

The first section of this volume summarizes the content of the draft environmental impact statement (DEIS) and this Addendum, which together constitute the final environmental impact statement (FEIS) prepared on the decommissioning of eight surplus plutonium production reactors at Hanford. The FEIS consists of two volumes. The first volume is the DEIS as written. The second volume (this Addendum) consists of a summary; Chapter 9, which contains comments on the DEIS and provides DOE`s responses to the comments; Appendix F, which provides additional health effects information; Appendix K, which contains costs of decommissioning in 1990 dollars; Appendix L, which containsmore » additional graphite leaching data; Appendix M, which contains a discussion of accident scenarios; Appendix N, which contains errata; and Appendix 0, which contains reproductions of the letters, transcripts, and exhibits that constitute the record for the public comment period.« less

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

Isotopic composition and neutronics of the Okelobondo natural reactor

NASA Astrophysics Data System (ADS)

Palenik, Christopher Samuel

The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (<1 ppm) in uraninite were estimated. Related to the burnup measured in Okelobondo samples (0.7 to 13.8 GWd/MTU), the final fission product inventories of Nd (90 to 1200 ppm) and Te (10 to 110 ppm) were calculated. By the same means, the ranges of all other fission products and actinides produced during reactor operation were calculated as a function of burnup. These results provide a source term against which the present elemental and decay abundances at the fission reactor can be compared. Furthermore, they provide new insights into the extent to which a "fossil" nuclear reactor can be characterized on the basis of its isotopic signatures. In addition, results from the study of two other natural systems related to the radionuclide and fission product transport are included. A detailed mineralogical characterization of the uranyl mineralogy at the Bangombe uranium deposit in Gabon, Africa was completed to improve geochemical models of the solubility-limiting phase. A study of the competing effects of radiation damage and annealing in a U-bearing crystal of zircon shows that low temperature annealing in actinide-bearing phases is significant in the annealing of radiation damage.

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

Krieg, R.

For future pressurized-water reactors, which should be designed against core-meltdown accidents, missiles generated inside the containment present a severe problem for its integrity. The masses and geometries of the missiles, as well as their velocities, may vary to a great extent. Therefore a reliable proof of the containment integrity is very difficult. In this article the potential sources of missiles are discussed, and the conclusion was reached that the generation of heavy missiles must be prevented. Steam explosions must not damage the reactor vessel head. Thus fragments of the head cannot become missiles that endanger the containment shell. Furthermore, duringmore » a melt-through failure of the reactor vessel under high pressure, the resulting forces must not catapult the whole vessel against the containment shell. Only missiles caused by hydrogen explosions may be tolerable, but shielding structures that protect the containment shell may be required. Further investigations are necessary. Finally, measures are described showing that the generation of heavy missiles can indeed be prevented. Investigations are currently being carried out that will confirm the strength of the reactor vessel head. In addition, a device for retaining the fragments of a failing reactor vessel is discussed.« less

Mass transfer study on the electrochemical removal of copper ions from synthetic effluents using reticulated vitreous carbon.

PubMed

Britto-Costa, Pedro H; Ruotolo, Luís Augusto M

2013-01-01

Porous electrodes have been successfully used for metal electrodeposition from diluted aqueous solution due to their high porosity and specific surface area, which lead to high mass transfer rates. This work studies the mass transfer of copper electrodeposition on reticulated vitreous carbon in a flow reactor without membrane. The flow configuration, otherwise the filter-press electrochemical reactors, was designed in order to minimize the pressure drop. The mass transfer coefficient was determined by voltammetric and galvanostatic electrodeposition. In the voltammetric experiments a Luggin capillary was used to measure the current-potential curves and to determine the limiting current (and, consequently, the mass transfer coefficient). In the galvanostatic experiments the concentration-time curves were obtained and considering a limiting current kinetics model, the mass transfer coefficient (k(m)) was determined for different flow velocities. The results showed that both methods give similar values of k(m), thus the voltammetric method can be recommended because it is faster and simpler. Finally, the reactor performance was compared with others from literature, and it was observed that the proposed reactor design has high Sherwood numbers similar to other reactor configurations using membranes and reticulated vitreous carbon electrodes.

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

VVER-440 and VVER-1000 reactor dosimetry benchmark - BUGLE-96 versus ALPAN VII.0

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

Duo, J. I.

2011-07-01

Document available in abstract form only, full text of document follows: Analytical results of the vodo-vodyanoi energetichesky reactor-(VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Inst. Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reducedmore » computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10%) between BUGLE-96 and ALPAN VII.O libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15% with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements. (authors)« less

Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

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

Okrent, D.

1997-06-23

This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, withmore » the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.« less

75 FR 30077 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on ESBWR

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-28

... Final Safety Evaluation Report (FSER) associated with the ESBWR design certification. The Subcommittee... from the website cited above or by contacting the identified DFO. Moreover, in view of the possibility...

77 FR 56239 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-12

... Draft Final Revision 1 to Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test Program... inconvenience. If attending this meeting, please enter through the One White Flint North building, 11555...

Tritium laboratory with multiple purposes at NIPNE Magurele Romania

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

Matei, L.; Postolache, C.

2008-07-15

The Tritium Laboratory from NIPNE (Romania)) is part of Radioisotope Research and Production Center. The Tritium Laboratory has been in operation since 1960, and carries out R and D activities involving tritium sources in gaseous, liquids and solid state, provides specialized service to CANDU NPP Cernavoda (Romania)), and provides tritium assay services to internal and external customers. The paper presents the activities and perspectives of Tritium Laboratory and its performances in accordance with Quality System Management. (authors)

Modeling CANDU-6 liquid zone controllers for effects of thorium-based fuels

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

St-Aubin, E.; Marleau, G.

2012-07-01

We use the DRAGON code to model the CANDU-6 liquid zone controllers and evaluate the effects of thorium-based fuels on their incremental cross sections and reactivity worth. We optimize both the numerical quadrature and spatial discretization for 2D cell models in order to provide accurate fuel properties for 3D liquid zone controller supercell models. We propose a low computer cost parameterized pseudo-exact 3D cluster geometries modeling approach that avoids tracking issues on small external surfaces. This methodology provides consistent incremental cross sections and reactivity worths when the thickness of the buffer region is reduced. When compared with an approximate annularmore » geometry representation of the fuel and coolant region, we observe that the cluster description of fuel bundles in the supercell models does not increase considerably the precision of the results while increasing substantially the CPU time. In addition, this comparison shows that it is imperative to finely describe the liquid zone controller geometry since it has a strong impact of the incremental cross sections. This paper also shows that liquid zone controller reactivity worth is greatly decreased in presence of thorium-based fuels compared to the reference natural uranium fuel, since the fission and the fast to thermal scattering incremental cross sections are higher for the new fuels. (authors)« less

Estimating cancer risk in relation to tritium exposure from routine operation of a nuclear-generating station in Pickering, Ontario.

PubMed

Wanigaratne, S; Holowaty, E; Jiang, H; Norwood, T A; Pietrusiak, M A; Brown, P

2013-09-01

Evidence suggests that current levels of tritium emissions from CANDU reactors in Canada are not related to adverse health effects. However, these studies lack tritium-specific dose data and have small numbers of cases. The purpose of our study was to determine whether tritium emitted from a nuclear-generating station during routine operation is associated with risk of cancer in Pickering, Ontario. A retrospective cohort was formed through linkage of Pickering and north Oshawa residents (1985) to incident cancer cases (1985-2005). We examined all sites combined, leukemia, lung, thyroid and childhood cancers (6-19 years) for males and females as well as female breast cancer. Tritium estimates were based on an atmospheric dispersion model, incorporating characteristics of annual tritium emissions and meteorology. Tritium concentration estimates were assigned to each cohort member based on exact location of residence. Person-years analysis was used to determine whether observed cancer cases were higher than expected. Cox proportional hazards regression was used to determine whether tritium was associated with radiation-sensitive cancers in Pickering. Person-years analysis showed female childhood cancer cases to be significantly higher than expected (standardized incidence ratio [SIR] = 1.99, 95% confidence interval [CI]: 1.08-3.38). The issue of multiple comparisons is the most likely explanation for this finding. Cox models revealed that female lung cancer was significantly higher in Pickering versus north Oshawa (HR = 2.34, 95% CI: 1.23-4.46) and that tritium was not associated with increased risk. The improved methodology used in this study adds to our understanding of cancer risks associated with low-dose tritium exposure. Tritium estimates were not associated with increased risk of radiationsensitive cancers in Pickering.

Development of advanced high heat flux and plasma-facing materials

NASA Astrophysics Data System (ADS)

Linsmeier, Ch.; Rieth, M.; Aktaa, J.; Chikada, T.; Hoffmann, A.; Hoffmann, J.; Houben, A.; Kurishita, H.; Jin, X.; Li, M.; Litnovsky, A.; Matsuo, S.; von Müller, A.; Nikolic, V.; Palacios, T.; Pippan, R.; Qu, D.; Reiser, J.; Riesch, J.; Shikama, T.; Stieglitz, R.; Weber, T.; Wurster, S.; You, J.-H.; Zhou, Z.

2017-09-01

Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling materials, thereby minimizing the release of tritium under normal operation conditions. Finally, solutions for the unique bonding requirements of dissimilar material used in a fusion reactor are demonstrated by describing the current status and prospects of functionally graded materials.

Performance of a full scale prototype detector at the BR2 reactor for the SoLid experiment

NASA Astrophysics Data System (ADS)

Abreu, Y.; Amhis, Y.; Arnold, L.; Ban, G.; Beaumont, W.; Bongrand, M.; Boursette, D.; Castle, B. C.; Clark, K.; Coupé, B.; Cussans, D.; De Roeck, A.; D'Hondt, J.; Durand, D.; Fallot, M.; Ghys, L.; Giot, L.; Guillon, B.; Ihantola, S.; Janssen, X.; Kalcheva, S.; Kalousis, L. N.; Koonen, E.; Labare, M.; Lehaut, G.; Manzanillas, L.; Mermans, J.; Michiels, I.; Moortgat, C.; Newbold, D.; Park, J.; Pestel, V.; Petridis, K.; Piñera, I.; Pommery, G.; Popescu, L.; Pronost, G.; Rademacker, J.; Ryckbosch, D.; Ryder, N.; Saunders, D.; Schune, M.-H.; Simard, L.; Vacheret, A.; Van Dyck, S.; Van Mulders, P.; van Remortel, N.; Vercaemer, S.; Verstraeten, M.; Weber, A.; Yermia, F.

2018-05-01

The SoLid collaboration has developed a new detector technology to detect electron anti-neutrinos at close proximity to the Belgian BR2 reactor at surface level. A 288 kg prototype detector was deployed in 2015 and collected data during the operational period of the reactor and during reactor shut-down. Dedicated calibration campaigns were also performed with gamma and neutron sources. This paper describes the construction of the prototype detector with a high control on its proton content and the stability of its operation over a period of several months after deployment at the BR2 reactor site. All detector cells provide sufficient light yields to achieve a target energy resolution of better than 20%/√E(MeV). The capability of the detector to track muons is exploited to equalize the light response of a large number of channels to a precision of 3% and to demonstrate the stability of the energy scale over time. Particle identification based on pulse-shape discrimination is demonstrated with calibration sources. Despite a lower neutron detection efficiency due to triggering constraints, the main backgrounds at the reactor site were determined and taken into account in the shielding strategy for the main experiment. The results obtained with this prototype proved essential in the design optimization of the final detector.

International Research Reactor Decommissioning Project

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

Leopando, Leonardo; Warnecke, Ernst

2008-01-15

Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement tomore » the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.« less

Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

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

Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly casesmore » are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.« less

IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

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

Brumovsky, M.; Steele, L.E.

1997-02-01

In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracturemore » mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.« less

77 FR 46528 - Sunshine Act Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-03

... Suspend Final Decisions on Reactor License Applications Pending Completion of Remanded Waste Confidence...-nrc/policy-making/schedule.html . * * * * * The NRC provides reasonable accommodation to individuals... Benefits Branch, at 301-415-6200, TDD: 301-415-2100, or by email at [email protected] . Determinations...

Microstructure evolution and tensile properties of Zr-2.5 wt.% Nb pressure tubes processed from billets with different microstructures

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

Kapoor, K.; Saratchandran, N.; Muralidharan, K.

1999-02-01

Pressurized heavy water reactors (PHWR) use zirconium-base alloys for their low neutron-absorption cross section, good mechanical strength, low irradiation creep, and high corrosion resistance in reactor atmospheres. Starting with identical ingots, billets having different microstructures were obtained by three different processing methods for fabrication of Zr-2.5 wt%Nb pressure tubes., The billets were further processed by hot extrusion and cold Pilger tube reducing to the finished product. Microstructural characterization was done at each stage of processing. The effects of the initial billet microstructure on the intermediate and final microstructure and mechanical property results were determined. It was found that the structuremore » at each stage and the final mechanical properties depend strongly on the initial billet microstructure. The structure at the final stage consists of elongated alpha zirconium grains with a network of metastable beta zirconium phase. Some of this metastable phase transforms into stable beta niobium during thermomechanical processing. Billets with quenched structure resulted in less beta niobium at the final stage. The air cooled billets resulted in a large amount of beta niobium. The tensile properties, especially the percentage elongation, were found to vary for the different methods. Higher percentage elongation was observed for billets having quenched structure. Extrusion and forging did not produce any characteristic differences in the properties. The results were used to select a process flow sheet which yields the desired mechanical properties with suitable microstructure in the final product.« less

Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

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

Was, Gary

This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

THE FUEL ELEMENT GRAPHITE. Project DRAGON.

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

Graham, L.W.; Price, M.S.T.

1963-01-15

The main requirements of a fuel element graphite for reactors based on the Dragon concept are low transmission coefficient for fission products, dimensional stability under service conditions, high strength, high thermal conductivity, high purity, and high resistance to oxidation. Since conclusions reached in early 1960, a considerable amount of information has accumulated concerning the likely behaviour of graphites in high temperature reactor systems, particularly data on dimensional stability under irradiation. The influence of this new knowledge on the development of fuel element graphite with the Dragon Project is discussed in detail in the final section of this paper.

Simulation of chemical-vapor-deposited silicon carbide for a cold wall vertical reactor

NASA Astrophysics Data System (ADS)

Lee, Y. L.; Sanchez, J. M.

1997-07-01

The growth rate of silicon carbide obtained by low-pressure chemical vapor deposition from tetramethylsilane is numerically simulated for a cold wall vertical reactor. The transport equations for momentum, heat, and mass transfer are simultaneously solved by employing the finite volume method. A model for reaction rate is also proposed in order to predict the measured growth rates [A. Figueras, S. Garelik, J. Santiso, R. Rodroguez-Clemente, B. Armas, C. Combescure, R. Berjoan, J.M. Saurel and R. Caplain, Mater. Sci. Eng. B 11 (1992) 83]. Finally, the effects of thermal diffusion on the growth rate are investigated.

Structural behavior of the Bitter plate tf magnet for the Zephyr ignition test reactor

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

Bobrov, E.S.; Becker, H.

1981-01-01

This paper discusses methods and results of the computer structural analysis of the Bitter plate toroidal field magnet design for the ZEPHYR Ignition Test Reactor. The magnet provides a field of 7.06 T at the center of the bore which is 1.76 m from the major toroidal axis. The ignited plasma is located at a major radius of 1.36 m where the magnetic field is 9.11 T. The plasma is moved to this final position following compression in the major radius. The horizontal bore of the magnet is 1.8 m.

78 FR 7465 - Agency Information Collection Activities: Submission for the Office of Management and Budget (OMB...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-01

..., and applicants for facility (i.e., nuclear power and non-power research and test reactor) operating... the final supporting statement, at the NRC's Public Document Room, Room O-1F21, One White Flint North...

A RADIAL REACTOR FOR TRICHLOROETHYLENE STEAM REFORMING. (R822721C633)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

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

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

Self-actuated shutdown system for a commercial size LMFBR. Final report

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

Dupen, C.F.G.

1978-08-01

A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility andmore » reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power.« less

An investigation of the treatment of particulate matter from gasoline engine exhaust using non-thermal plasma.

PubMed

Ye, Dan; Gao, Dengshan; Yu, Gang; Shen, Xianglin; Gu, Fan

2005-12-09

A plasma reactor with catalysts was used to treat exhaust gas from a gasoline engine in order to decrease particulate matter (PM) emissions. The effect of non-thermal plasma (NTP) of the dielectric discharges on the removal of PM from the exhaust gas was investigated experimentally. The removal efficiency of PM was based on the concentration difference in PM for particle diameters ranging from 0.3 to 5.0 microm as measured by a particle counter. Several factors affecting PM conversion, including the density of plasma energy, reaction temperature, flow rate of exhaust gas, were investigated in the experiment. The results indicate that PM removal efficiency ranged approximately from 25 to 57% and increased with increasing energy input in the reactor, reaction temperature and residence time of the exhaust gas in the reactor. Enhanced removal of the PM was achieved by filling the discharge gap of the reactor with Cu-ZSM-5 catalyst pellets. In addition, the removal of unburned hydrocarbons was studied. Finally, available approaches for PM conversion were analyzed involving the interactions between discharge and catalytic reactions.

Novel windows for "solar commodities": a device for CO2 reduction using plasmonic catalyst activation.

PubMed

Navarrete, Alexander; Muñoz, Sergio; Sanz-Moral, Luis M; Brandner, Juergen J; Pfeifer, Peter; Martín, Ángel; Dittmeyer, Roland; Cocero, María J

2015-01-01

A novel plasmonic reactor concept is proposed and tested to work as a visible energy harvesting device while allowing reactions to transform CO2 to be carried out. Particularly the reverse water gas shift (RWGS) reaction has been tested as a means to introduce renewable energy into the economy. The development of the new reactor concept involved the synthesis of a new composite capable of plasmonic activation with light, the development of an impregnation method to create a single catalyst reactor entity, and finally the assembly of a reaction system to test the reaction. The composite developed was based on a Cu/ZnO catalyst dispersed into transparent aerogels. This allows efficient light transmission and a high surface area for the catalyst. An effective yet simple impregnation method was developed that allowed introduction of the composites into glass microchannels. The activation of the reaction was made using LEDs that covered all the sides of the reactor allowing a high power delivery. The results of the reaction show a stable process capable of low temperature transformations.

Small modular reactor: First-of-a-Kind (FOAK) and Nth-of-a-Kind (NOAK) Economic Analysis

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

Boldon, Lauren M.; Sabharwall, Piyush

2014-08-01

Small modular reactors (SMRs) refer to any reactor design in which the electricity generated is less than 300 MWe. Often medium sized reactors with power less than 700 MWe are also grouped into this category. Internationally, the development of a variety of designs for SMRs is booming with many designs approaching maturity and even in or nearing the licensing stage. It is for this reason that a generalized yet comprehensive economic model for first of a kind (FOAK) through nth of a kind (NOAK) SMRs based upon rated power, plant configuration, and the fiscal environment was developed. In the model,more » a particular project’s feasibility is assessed with regards to market conditions and by commonly utilized capital budgeting techniques, such as the net present value (NPV), internal rate of return (IRR), Payback, and more importantly, the levelized cost of energy (LCOE) for comparison to other energy production technologies. Finally, a sensitivity analysis was performed to determine the effects of changing debt, equity, interest rate, and conditions on the LCOE.« less

Sequential two-column electro-Fenton-photolytic reactor for the treatment of winery wastewater.

PubMed

Díez, A M; Sanromán, M A; Pazos, M

2017-01-01

The high amount of winery wastewaters produced each year makes their treatment a priority issue due to their problematic characteristics such as acid pH, high concentration of organic load and colourful compounds. Furthermore, some of these effluents can have dissolved pesticides, due to the previous grape treatments, which are recalcitrant to conventional treatments. Recently, photo-electro-Fenton process has been reported as an effective procedure to mineralize different organic contaminants and a promising technology for the treatment of these complex matrixes. However, the reactors available for applying this process are scarce and they show several limitations. In this study, a sequential two-column reactor for the photo-electro-Fenton treatment was designed and evaluated for the treatment of different pesticides, pirimicarb and pyrimethanil, used in wine production. Both studied pesticides were efficiently removed, and the transformation products were determined. Finally, the treatment of a complex aqueous matrix composed by winery wastewater and the previously studied pesticides was carried out in the designed sequential reactor. The high removals of TOC and COD reached and the low energy consumption demonstrated the efficiency of this new configuration.

Automated in Situ Measurement of Gas Solubility in Liquids with a Simple Tube-in-Tube Reactor.

PubMed

Zhang, Jisong; Teixeira, Andrew R; Zhang, Haomiao; Jensen, Klavs F

2017-08-15

Data on the solubilities of gases in liquids are foundational for assessing a variety of multiphase separations and gas-liquid reactions. Taking advantage of the tube-in-tube reactor design built with semipermeable Teflon AF-2400 tubes, liquids can be rapidly saturated without direct contacting of gas and liquid. The gas solubility can be determined by performing steady-state flux balances of both the gas and liquid flowing into the reactor system. Using this type of reactor, a fully automated strategy has been developed for the rapid in situ measurement of gas solubilities in liquids. The developed strategy enables precise gas solubility measurements within 2-5 min compared with 4-5 h using conventional methods. This technique can be extended to the discrete multipoint steady-state and continuous ramped-multipoint data acquisition methods. The accuracy of this method has been validated against several gas-liquid systems, showing less than 2% deviation from known values. Finally, this strategy has been extended to measure the temperature dependence of gas solubilities in situ and to estimate the local enthalpy of dissolution across a defined temperature range.

Modular fabrication and characterization of complex silicon carbide composite structures Advanced Reactor Technologies (ART) Research Final Report (Feb 2015 – May 2017)

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

Khalifa, Hesham

Advanced ceramic materials exhibit properties that enable safety and fuel cycle efficiency improvements in advanced nuclear reactors. In order to fully exploit these desirable properties, new processing techniques are required to produce the complex geometries inherent to nuclear fuel assemblies and support structures. Through this project, the state of complex SiC-SiC composite fabrication for nuclear components has advanced significantly. New methods to produce complex SiC-SiC composite structures have been demonstrated in the form factors needed for in-core structural components in advanced high temperature nuclear reactors. Advanced characterization techniques have been employed to demonstrate that these complex SiC-SiC composite structures providemore » the strength, toughness and hermeticity required for service in harsh reactor conditions. The complex structures produced in this project represent a significant step forward in leveraging the excellent high temperature strength, resistance to neutron induced damage, and low neutron cross section of silicon carbide in nuclear applications.« less

Conversion of NO with a catalytic packed-bed dielectric barrier discharge reactor

NASA Astrophysics Data System (ADS)

Xu, CAO; Weixuan, ZHAO; Renxi, ZHANG; Huiqi, HOU; Shanping, CHEN; Ruina, ZHANG

2017-11-01

This paper discusses the conversion of nitric oxide (NO) with a low-temperature plasma induced by a catalytic packed-bed dielectric barrier discharge (DBD) reactor. Alumina oxide (Al2O3), glass (SiO2) and zirconium oxide (ZrO2), three different spherical packed materials of the same size, were each present in the DBD reactor. The NO conversion under varying input voltage and specific energy density, and the effects of catalysts (titanium dioxide (TiO2) and manganese oxide (MnO x ) coated on Al2O3) on NO conversion were investigated. The experimental results showed that NO conversion was greatly enhanced in the presence of packed materials in the reactor, and the catalytic packed bed of MnO x /Al2O3 showed better performance than that of TiO2/Al2O3. The surface and crystal structures of the materials and catalysts were characterized through scanning electron microscopy analysis. The final products were clearly observed by a Fourier transform infrared spectrometer and provided a better understanding of NO conversion.

Disposal of radioactive iodine in space

NASA Technical Reports Server (NTRS)

Burns, R. E.; Defield, J. G.

1978-01-01

The possibility of space disposal of iodine waste from nuclear power reactors is investigated. The space transportation system utilized relies upon the space shuttle, a liquid hydrogen/liquid oxygen orbit transfer vehicle, and a solid propellant final stage. The iodine is assumed to be in the form of either an iodide or an iodate, and calculations assume that the final destination is either solar orbit or solar system escape. It is concluded that space disposal of iodine is feasible.

Physical-chemical treatment of wastes: a way to close turnover of elements in LSS

NASA Astrophysics Data System (ADS)

Kudenko, Yu A.; Gribovskaya, I. V.; Zolotukhin, I. G.

2000-05-01

"Man-plants-physical-chemical unit" system designed for space stations or terrestrial ecohabitats to close steady-state mineral, water and gas exchange is proposed. The physical-chemical unit is to mineralize all inedible plant wastes and physiological human wastes (feces, urine, gray water) by electromagnetically activated hydrogen peroxide in an oxidation reactor. The final product is a mineralized solution containing all elements balanced for plants' requirements. The solution has been successfully used in experiments to grow wheat, beans and radish. The solution was reusable: the evaporated moisture was replenished by the phytotron condensate. Sodium salination of plants was precluded by evaporating reactor-mineralized urine to sodium saturation concentration to crystallize out NaCl which can be used as food for the crew. The remaining mineralized product was brought back for nutrition of plants. The gas composition of the reactor comprises O 2, N 2, CO 2, NH 3, H 2. At the reactor's output hydrogen and oxygen were catalyzed into water, NH 3 was converted in a water trap into NH 4 and used for nutrition of plants. A special accessory at the reactor's output may produce hydrogen peroxide from intrasystem water and gas which makes possible to close gas loops between LSS components.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, Michael M.; Schulz, Terry L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, M.M.; Schulz, T.L.

1993-12-07

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

Dynamic System Simulation of the KRUSTY Experiment

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

Klein, Steven Karl; Kimpland, Robert Herbert

2016-05-09

The proposed KRUSTY experiment is a demonstration of a reactor operating at power. The planned experimental configuration includes a highly enriched uranium (HEU) reflected core, cooled by multiple heat pipes leading to Stirling engines for primary heat rejection. Operating power is expected to be approximately four (4) to five (5) kilowatts with a core temperature above 1,000 K. No data is available on any historical reactor employing HEU metal that operated over the temperature range required for the KRUSTY experiment. Further, no reactor has operated with heat pipes as the primary cooling mechanism. Historic power reactors have employed either naturalmore » or forced convection so data on their operation is not directly applicable to the KRUSTY experiment. The primary purpose of the system model once developed and refined by data from these component experiments, will be used to plan the KRUSTY experiment. This planning will include expected behavior of the reactor from start-up, through various transient conditions where cooling begins to become present and effective, and finally establishment of steady-state. In addition, the model can provide indicators of anticipated off-normal events and appropriate operator response to those conditions. This information can be used to develop specific experiment operating procedures and aids to guide the operators in conduct of the experiment.« less

Application of Radiation Chemistry to Some Selected Technological Issues Related to the Development of Nuclear Energy.

PubMed

Bobrowski, Krzysztof; Skotnicki, Konrad; Szreder, Tomasz

2016-10-01

The most important contributions of radiation chemistry to some selected technological issues related to water-cooled reactors, reprocessing of spent nuclear fuel and high-level radioactive wastes, and fuel evolution during final radioactive waste disposal are highlighted. Chemical reactions occurring at the operating temperatures and pressures of reactors and involving primary transients and stable products from water radiolysis are presented and discussed in terms of the kinetic parameters and radiation chemical yields. The knowledge of these parameters is essential since they serve as input data to the models of water radiolysis in the primary loop of light water reactors and super critical water reactors. Selected features of water radiolysis in heterogeneous systems, such as aqueous nanoparticle suspensions and slurries, ceramic oxides surfaces, nanoporous, and cement-based materials, are discussed. They are of particular concern in the primary cooling loops in nuclear reactors and long-term storage of nuclear waste in geological repositories. This also includes radiation-induced processes related to corrosion of cladding materials and copper-coated iron canisters, dissolution of spent nuclear fuel, and changes of bentonite clays properties. Radiation-induced processes affecting stability of solvents and solvent extraction ligands as well oxidation states of actinide metal ions during recycling of the spent nuclear fuel are also briefly summarized.

Composting in small laboratory pilots: performance and reproducibility.

PubMed

Lashermes, G; Barriuso, E; Le Villio-Poitrenaud, M; Houot, S

2012-02-01

Small-scale reactors (<10 l) have been employed in composting research, but few attempts have assessed the performance of composting considering the transformations of organic matter. Moreover, composting at small scales is often performed by imposing a fixed temperature, thus creating artificial conditions, and the reproducibility of composting has rarely been reported. The objectives of this study are to design an innovative small-scale composting device safeguarding self-heating to drive the composting process and to assess the performance and reproducibility of composting in small-scale pilots. The experimental setup included six 4-l reactors used for composting a mixture of sewage sludge and green wastes. The performance of the process was assessed by monitoring the temperature, O(2) consumption and CO(2) emissions, and characterising the biochemical evolution of organic matter. A good reproducibility was found for the six replicates with coefficients of variation for all parameters generally lower than 19%. An intense self-heating ensured the existence of a spontaneous thermophilic phase in all reactors. The average loss of total organic matter (TOM) was 46% of the initial content. Compared to the initial mixture, the hot water soluble fraction decreased by 62%, the hemicellulose-like fraction by 68%, the cellulose-like fraction by 50% and the lignin-like fractions by 12% in the final compost. The TOM losses, compost stabilisation and evolution of the biochemical fractions were similar to observed in large reactors or on-site experiments, excluding the lignin degradation, which was less important than in full-scale systems. The reproducibility of the process and the quality of the final compost make it possible to propose the use of this experimental device for research requiring a mass reduction of the initial composted waste mixtures. Copyright © 2011 Elsevier Ltd. All rights reserved.

Role of H2O2 in the fluctuating patterns of COD (chemical oxygen demand) during the treatment of palm oil mill effluent (POME) using pilot scale triple frequency ultrasound cavitation reactor.

PubMed

Manickam, Sivakumar; Abidin, Norhaida binti Zainal; Parthasarathy, Shridharan; Alzorqi, Ibrahim; Ng, Ern Huay; Tiong, Timm Joyce; Gomes, Rachel L; Ali, Asgar

2014-07-01

Palm oil mill effluent (POME) is a highly contaminating wastewater due to its high chemical oxygen demand (COD) and biochemical oxygen demand (BOD). Conventional treatment methods require longer residence time (10-15 days) and higher operating cost. Owing to this, finding a suitable and efficient method for the treatment of POME is crucial. In this investigation, ultrasound cavitation technology has been used as an alternative technique to treat POME. Cavitation is the phenomenon of formation, growth and collapse of bubbles in a liquid. The end process of collapse leads to intense conditions of temperature and pressure and shock waves which assist various physical and chemical transformations. Two different ultrasound systems i.e. ultrasonic bath (37 kHz) and a hexagonal triple frequency ultrasonic reactor (28, 40 and 70 kHz) of 15 L have been used. The results showed a fluctuating COD pattern (in between 45,000 and 60,000 mg/L) while using ultrasound bath alone, whereas a non-fluctuating COD pattern with a final COD of 27,000 mg/L was achieved when hydrogen peroxide was introduced. Similarly for the triple frequency ultrasound reactor, coupling all the three frequencies resulted into a final COD of 41,300 mg/L compared to any other individual or combination of two frequencies. With the possibility of larger and continuous ultrasonic cavitational reactors, it is believed that this could be a promising and a fruitful green process engineering technique for the treatment of POME. Copyright © 2014 Elsevier B.V. All rights reserved.

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

Monteleone, S.

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updatedmore » and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.« less

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE PAGES

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

2018-02-02

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Environmental radiation protection studies related to nuclear industries, using AMS

NASA Astrophysics Data System (ADS)

Hellborg, Ragnar; Erlandsson, Bengt; Faarinen, Mikko; Hâkansson, Helena; Hâkansson, Kjell; Kiisk, Madis; Magnusson, Carl-Erik; Persson, Per; Skog, Göran; Stenström, Kristina; Mattsson, Sören; Thornberg, Charlotte

2001-07-01

14C is produced in nuclear reactors during normal operation and part of it is continuously released into the environment. Because of the biological importance of carbon and the long physical half-life of 14C it is of interest to study these releases. The 14C activity concentrations in the air and vegetation around some Swedish as well as foreign nuclear facilities have been measured by accelerator mass spectrometry (AMS). 59Ni is produced by neutron activation in the stainless steel close to the core of a nuclear reactor. The 59Ni levels have been measured in order to be able to classify the different parts of the reactor with respect to their content of long-lived radionuclides before final storage. The technique used to measure 59Ni at a small accelerator such as the Lund facility has been developed over the past few years and material from the Swedish nuclear industry has been analyzed.

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

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

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Field Testing of a Pneumatic Regolith Feed System During a 2010 ISRU Field Campaign on Mauna Kea, Hawaii

NASA Technical Reports Server (NTRS)

Craft, Jack; Zacny, Kris; Chu, Philip; Wilson, Jack; Santoro, Chris; Carlson, Lee; Maksymuk, Michael; Townsend, Ivan I.; Mueller, Robert P.; Mantovani, James G.

2010-01-01

Lunar In Situ Resource Utilization (ISRU) consists of a number of tasks starting with mining of lunar regolith, followed by the transfer of regolith to an oxygen extraction reactor and finally processing the regolith and storing of extracted oxygen. The transfer of regolith from the regolith hopper at the ground level to an oxygen extraction reactor many feet above the surface could be accomplished in different ways, including using a mechanical auger, bucket ladder system or a pneumatic system. The latter system is commonly used on earth when moving granular materials since it offers high reliability and simplicity of operation. In this paper, we describe a pneumatic regolith feed system, delivering feedstock to a Carbothermal reactor and lessons learned from deploying the system during the 2010 ISRU field campaign on the Mauna Kea, Hawaii.

Vibration and acoustic noise emitted by dry-type air-core reactors under PWM voltage excitation

NASA Astrophysics Data System (ADS)

Li, Jingsong; Wang, Shanming; Hong, Jianfeng; Yang, Zhanlu; Jiang, Shengqian; Xia, Shichong

2018-05-01

According to coupling way between the magnetic field and the structural order, structure mode is discussed by engaging finite element (FE) method and both natural frequency and modal shape for a dry-type air-core reactor (DAR) are obtained in this paper. On the basis of harmonic response analysis, electromagnetic force under PWM (Pulse Width Modulation) voltage excitation is mapped with the structure mesh, the vibration spectrum is gained and the consequences represents that the whole structure vibration predominates in the radial direction, with less axial vibration. Referring to the test standard of reactor noise, the rules of emitted noise of the DAR are measured and analyzed at chosen switching frequency matches the sample resonant frequency and the methods of active vibration and noise reduction are put forward. Finally, the low acoustic noise emission of a prototype DAR is verified by measurement.

Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

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

Volkov, Victor; Danilovich, Alexey; Zverkov, Yuri

2013-07-01

In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channelmore » of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66A, 66B, 72, 64, 63) - as well as from water and gas loop corridors - was dismantled, with the total radwaste weight of 53 tons and the total removed activity of 5,0 x 10{sup 10} Bq; - loop-type channel equipment from underground reactor hall premises was dismantled; - 93 loop-type channels were characterized, chopped and removed, with radwaste of 2.6 x 10{sup 13} Bq ({sup 60}Co) and 1.5 x 10{sup 13} Bq ({sup 137}Cs) total activity removed from the reactor pool, fragmented and packaged. Some of this waste was placed into the high-level waste (HLW) repository of the Center. Dismantling works were executed with application of remotely operated mechanisms, which promoted decrease of radiation impact on the personnel. The average individual dose for the personnel was 1.9 mSv/year in 2011, and the collective dose is estimated as 0.0605 man x Sv/year. (authors)« less

Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

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

Woods, Brian; Gutowska, Izabela; Chiger, Howard

Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion tomore » ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with an electrical heat source. The peak core region temperature capability is 1400°C. As part of this project, an inventory of test facilities that could be used for these experimental programs was completed. Several of these facilities showed some promise, however, upon further investigation it became clear that only the OSU HTTF had the power and/or peak temperature limits that would allow for the experimental programs envisioned herein. Thus the conceptual design and feasibility study development focused on examining the feasibility of configuring the current HTTF to collect validation data for these experimental programs. In addition to the scaling analyses and conceptual design development, a test plan was developed for the envisioned modified test facility. This test plan included a discussion on an appropriate shakedown test program as well as the specific matrix tests. Finally, a feasibility study was completed to determine the cost and schedule considerations that would be important to any test program developed to investigate these designs and events.« less

FINAL PROJECT REPORT - EVALUATION AND TESTING OF HTGR REACTOR BUILDING RESPONSE TO DEPRESSURIZATION ACCIDENTS

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

ALLIANCE LIMITED, NGNP INDUSTRY

This report provides a description of the project, summarizes each phase of the project, and ends with project conclusions. In addition, the report contains a descriptive index of the technical reports generated during the course of the project.

76 FR 80409 - Draft Supplement 2 to Final Environmental Statement Related to the Operation of Watts Bar Nuclear...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-23

... image files of the NRC's public documents. If you do not have access to ADAMS or if there are problems.... Stephen J. Campbell, Chief, Watts Bar Special Projects Branch, Division of Operating Reactor Licensing...

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

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

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

2014-06-01

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

Composting domestic sewage sludge with natural zeolites in a rotary drum reactor.

PubMed

Villaseñor, J; Rodríguez, L; Fernández, F J

2011-01-01

This work aimed the influence of zeolites addition on a sludge-straw composting process using a pilot-scale rotary drum reactor. The type and concentration of three commercial natural zeolites were considered: a mordenite and two clinoptilolites (Klinolith and Zeocat). Mordenite caused the greatest carbon removal (58%), while the clinoptilolites halved losses of ammonium. All zeolites removed 100% of Ni, Cr, Pb, and significant amounts (more than 60%) of Cu, Zn and Hg. Zeocat displayed the greatest retention of ammonium and metals, and retention efficiencies increased as Zeocat concentration increased. The addition of 10% Zeocat produced compost compliant with Spanish regulations. Zeolites were separated from the final compost, and leaching studies suggested that zeolites leachates contained very low metals concentrations (<1 mg/kg). Thus, the final compost could be applied directly to soil, or metal-polluted zeolites could be separated from the compost prior to application. The different options have been discussed. Copyright © 2010 Elsevier Ltd. All rights reserved.

324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

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

HUMPHREYS, D C

A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Teammore » counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.« less

Corrosion of carbon steel by bacteria from North Sea offshore seawater injection systems: laboratory investigation.

PubMed

Stipanicev, Marko; Turcu, Florin; Esnault, Loïc; Rosas, Omar; Basseguy, Régine; Sztyler, Magdalena; Beech, Iwona B

2014-06-01

Influence of sulfidogenic bacteria, from a North Sea seawater injection system, on the corrosion of S235JR carbon steel was studied in a flow bioreactor; operating anaerobically for 100days with either inoculated or filtrated seawater. Deposits formed on steel placed in reactors contained magnesium and calcium minerals plus iron sulfide. The dominant biofilm-forming organism was an anaerobic bacterium, genus Caminicella, known to produce hydrogen sulfide and carbon dioxide. Open Circuit Potentials (OCP) of steel in the reactors was, for nearly the entire test duration, in the range -80045), suggested pitting on steel samples within the inoculated environment. However, the actual degree of corrosion could neither be directly correlated with the electrochemical data and nor with the steel corrosion in the filtrated seawater environment. Further laboratory tests are thought to clarify the noticed apparent discrepancies. © 2013.

Cure kinetics, morphologies, and mechanical properties of thermoplastic/MWCNT modified multifunctional glassy epoxies prepared via continuous reaction methods

NASA Astrophysics Data System (ADS)

Cheng, Xiaole

The primary goal of this dissertation is to develop a novel continuous reactor method to prepare partially cured epoxy prepolymers for aerospace prepreg applications with the aim of replacing traditional batch reactors. Compared to batch reactors, the continuous reactor is capable of solubilizing and dispersing a broad range of additives including thermoplastic tougheners, stabilizers, nanoparticles and curatives and advancing epoxy molecular weights and viscosities while reducing energy consumption. In order to prove this concept, polyethersulfone (PES) modified 4, 4'-diaminodiphenylsulfone (44DDS)/tetraglycidyl-4, 4'-diaminodiphenylmethane (TGDDM) epoxy prepolymers were firstly prepared using both continuous reactor and batch reactor methods. Kinetic studies confirmed the chain extension reaction in the continuous reactor is similar to the batch reactor, and the molecular weights and viscosities of prepolymers were readily controlled through reaction kinetics. Atomic force microscopy (AFM) confirmed similar cured network morphologies for formulations prepared from batch and continuous reactors. Additionally tensile strength, tensile modulus and fracture toughness analyses concluded mechanical properties of cured epoxy matrices produced from both reactors were equivalent. Effects of multifunctional epoxy compositions on thermoplastics phase-separated morphologies were systematically studied using a combination of AFM with nanomechanical mapping, spectroscopic and calorimetric techniques to provide new insights to tailor cured reaction induced phase separation (CRIPS) in multifunctional epoxy blend networks. Furthermore, how resultant crosslinked glassy polymer network and phase-separated morphologies correlated with mechanical properties are discussed in detail. Multiwall carbon nanotube (MWCNT)/TGDDM epoxy prepolymers were further prepared by combining the successful strategies for advancing epoxy chemistries and dispersing nanotubes using the continuous reactor. Optical microscopy (OM) and scanning electron microscopy (SEM) were used to characterize the MWCNT dispersion states and stabilization in epoxy prepolymer matrix after continuous process and during curing cycles. Additionally, electrical conductivities and mechanical properties of final cured MWCNT/TGDDM composites were measured and discussed in view of their corresponding MWCNT dispersion states. Ternary blends of MWCNT reinforced thermoplastic/epoxy prepolymers were prepared by the continuous reactor. Influence of MWCNT on the CRIPS mechanism and the cured morphologies were systematically investigated using SEM and rheological analysis. Incorporation of MWCNT in thermoplastic/epoxy matrices can lead to a morphological transformation from phase inverted, to co-continuous, and to droplet dispersed morphology. In additional, dynamic mechanical analysis revealed the heterogeneity of MWCNT dispersion in thermoplastic/thermosets systems.

Artificial neural networks modelling the prednisolone nanoprecipitation in microfluidic reactors.

PubMed

Ali, Hany S M; Blagden, Nicholas; York, Peter; Amani, Amir; Brook, Toni

2009-06-28

This study employs artificial neural networks (ANNs) to create a model to identify relationships between variables affecting drug nanoprecipitation using microfluidic reactors. The input variables examined were saturation levels of prednisolone, solvent and antisolvent flow rates, microreactor inlet angles and internal diameters, while particle size was the single output. ANNs software was used to analyse a set of data obtained by random selection of the variables. The developed model was then assessed using a separate set of validation data and provided good agreement with the observed results. The antisolvent flow rate was found to have the dominant role on determining final particle size.

MCNP6 simulated performance of Micro-Pocket Fission Detectors (MPFDs) in the Transient REActor Test (TREAT) Facility

DOE PAGES

Reichenberger, Michael A.; Patel, Vishal K.; Roberts, Jeremy A.; ...

2017-03-03

Here, Micro-Pocket Fission Detectors (MPFDs) are under development for in-core neutron flux measurements at the Transient REActor Test facility (TREAT) and in other experiments at Idaho National Laboratory (INL). The sensitivity of MPFDs to the energy dependent neutron flux at TREAT has been determined for 0.0300-μm thick active material coatings of 242Pu, 232Th, natural uranium, and 93% enriched 235U. Self-shielding effects in the active material of the MPFD was also confirmed to be negligible. Finally, fission fragment energy deposition was found to be in conformance with previously reported results.

Performance testing and analyses of the VSC-17 ventilated concrete cask. Final report

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

McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

1992-05-01

This document details performance test which was conducted on a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask configured for pressurized-water reactor (PWR) spent fuel. The performance test consisted of loading the VSC-17 cask with 17 canisters of consolidated PWR spent fuel from Virginia Power`s Surry and Florida Power & Light Turkey Point reactors. Cask surface, concrete, air channel surfaces, and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in a vertical cask orientation. Data on spent fuel integrity were also obtained.

Diels–Alder reactions of myrcene using intensified continuous-flow reactors

PubMed Central

Álvarez-Diéguez, Miguel Á; Kohl, Thomas M; Tsanaktsidis, John

2017-01-01

This work describes the Diels–Alder reaction of the naturally occurring substituted butadiene, myrcene, with a range of different naturally occurring and synthetic dienophiles. The synthesis of the Diels–Alder adduct from myrcene and acrylic acid, containing surfactant properties, was scaled-up in a plate-type continuous-flow reactor with a volume of 105 mL to a throughput of 2.79 kg of the final product per day. This continuous-flow approach provides a facile alternative scale-up route to conventional batch processing, and it helps to intensify the synthesis protocol by applying higher reaction temperatures and shorter reaction times. PMID:28228853

Characterization of BOR-60 Irradiated 14YWT-NFA1 Tubes

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

Saleh, Tarik A.; Maloy, Stuart Andrew; Aydogan, Eda

2017-02-15

Tubes of FCRD 14YWT-NFA1 Alloy were placed in the BOR-60 reactor and irradiated under a fast flux neutron environment to two conditions: 7 dpa at 360-370 °C and 6 dpa at 385-430 °C. Small sections of the tube were cut and sent to UC Berkeley for nanohardness testing and focused ion beam (FIB) milling of TEM specimens. FIB specimens were sent back to LANL for final FIB milling and TEM imaging. Hardness data and TEM images are presented in this report. This is the first fast reactor neutron irradiated information on the 14YWT-NFA1 alloy.

Selection of alternative central-station technologies for the Satellite Power System (SPS) comparative assessment

NASA Technical Reports Server (NTRS)

Samsa, M.

1980-01-01

An important effort is the Satellite Power System (SPS) comparative Assessment is the selection and characterization of alternative technologies to be compared with the SPS concept. The ground rules, criteria, and screening procedure applied in the selection of those alternative technologies are summarized. The final set of central station alternatives selected for comparison with the SPS concept includes: (1) light water reactor with improved fuel utilization, (2) conventional coal combustion with improved environmental controls, (3) open cycle gas turbine with integral low Btu gasifier, (4) terrestrial photovoltaic, (5) liquid metal fast breeder reactor, and (6) magnetic confinement fusion.

76 FR 5218 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-28

..., October 21, 2010 (74 FR 65038-65039). Thursday, February 10, 2011, Conference Room T2-B1, 11545 Rockville... Evaluation Report Associated with the License Renewal Application for the Palo Verde Nuclear Generating... NRC staff and Arizona Public Service Company regarding the final Safety Evaluation Report associated...

A Hardware-in-the-Loop Simulation Platform for the Verification and Validation of Safety Control Systems

NASA Astrophysics Data System (ADS)

Rankin, Drew J.; Jiang, Jin

2011-04-01

Verification and validation (V&V) of safety control system quality and performance is required prior to installing control system hardware within nuclear power plants (NPPs). Thus, the objective of the hardware-in-the-loop (HIL) platform introduced in this paper is to verify the functionality of these safety control systems. The developed platform provides a flexible simulated testing environment which enables synchronized coupling between the real and simulated world. Within the platform, National Instruments (NI) data acquisition (DAQ) hardware provides an interface between a programmable electronic system under test (SUT) and a simulation computer. Further, NI LabVIEW resides on this remote DAQ workstation for signal conversion and routing between Ethernet and standard industrial signals as well as for user interface. The platform is applied to the testing of a simplified implementation of Canadian Deuterium Uranium (CANDU) shutdown system no. 1 (SDS1) which monitors only the steam generator level of the simulated NPP. CANDU NPP simulation is performed on a Darlington NPP desktop training simulator provided by Ontario Power Generation (OPG). Simplified SDS1 logic is implemented on an Invensys Tricon v9 programmable logic controller (PLC) to test the performance of both the safety controller and the implemented logic. Prior to HIL simulation, platform availability of over 95% is achieved for the configuration used during the V&V of the PLC. Comparison of HIL simulation results to benchmark simulations shows good operational performance of the PLC following a postulated initiating event (PIE).

Trends in the short-term release of fission products and actinides to aqueous solution from used CANDU fuels at elevated temperature

NASA Astrophysics Data System (ADS)

Stroes-Gascoyne, S.

1992-08-01

A large number of short-term leaching experiments has been performed to determine fission product and actinide release from used CANDU (CANada Deuterium Uranium) fuels and to establish which factors affect release. Results are reported after30 ± 10 d leaching at 100-150°C under oxidizing (air) or reducing (Ar-3% H 2 or Ar) conditions, in various synthetic groundwaters. Cesium-137 release (0.007-6%) was positively correlated with increases in fuel power, leachant temperature and ionic strength. Strontium-90 release (0.0003-0.3%) generally increased with ionic strength, higher temperature and redox conditions. Actinide and Tc concentrations were compared to ranges calculated with a thermodynamic equilibrium model, that accounts for the uncertain geochemical parameters of a nuclear waste vault by calculating concentration ranges based on 40000 hypothetical cases. Experimental U concentrations (10 -8.5 to 10 -3 mol/kg) were higher than the model range, probably because of higher redox potentials in the experiments. Measured Pu concentrations (10 -12.5 to 10 -7 mol/kg) were at the low end of the calculated range. Americium and Cm concentrations (10 -12.5 to 10 -7 and 10 -15 to 10 -9 mol/kg, respectively) were highest under oxidizing conditions and higher temperatures. Technetium-99 concentrations (10 -5.5 to 10 -10.5 mol/kg) covered a much narrower range than calculated by the model.

Isotopic Transmutations in Irradiated Beryllium and Their Implications on MARIA Reactor Operation

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

Andrzejewski, Krzysztof J.; Kulikowska, Teresa A

2004-04-15

Beryllium irradiated by neutrons with energies above 0.7 MeV undergoes (n,{alpha}) and (n,2n) reactions. The Be(n,{alpha}) reaction results in subsequent buildup of {sup 6}Li and {sup 3}He isotopes with large thermal neutron absorption cross sections causing poisoning of irradiated beryllium. The amount of the poison isotopes depends on the neutron flux level and spectrum. The high-flux MARIA reactor operated in Poland since 1975 consists of a beryllium matrix with fuel channels in cutouts of beryllium blocks. As the experimental determination of {sup 6}Li, {sup 3}H, and {sup 3}He content in the operational reactor is impossible, a systematic computational study ofmore » the effect of {sup 3}He and {sup 6}Li presence in beryllium blocks on MARIA reactor reactivity and power density distribution has been undertaken. The analysis of equations governing the transmutation has been done for neutron flux parameters typical for MARIA beryllium blocks. Study of the mutual influence of reactor operational parameters and the buildup of {sup 6}Li, {sup 3}H, and {sup 3}He in beryllium blocks has shown the necessity of a detailed spatial solution of transmutation equations in the reactor, taking into account the whole history of its operation. Therefore, fuel management calculations using the REBUS code with included chains for Be(n,{alpha})-initiated reactions have been done for the whole reactor lifetime. The calculated poisoning of beryllium blocks has been verified against the critical experiment of 1993. Finally, the current {sup 6}Li, {sup 3}H, and {sup 3}He contents, averaged for each beryllium block, have been calculated. The reactivity drop caused by this poisoning is {approx}7%.« less

Advanced Instrumentation for Transient Reactor Testing

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

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

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

Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2015-07-01

Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of (135)Xe and (149)Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the (135)Xe and (149)Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The (149)Sm reactivities could be neglected compared to (135)Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4-Al produced the highest (135)Xe reactivity in all the studied fuel group during the reactor operation (0.39 mk) and after the reactor shutdown (0.735 mk), It followed by U3Si-Al (0.34 mk, 0.653 mk), U3Si2-Al (0.33 mk, 0.634 mk), U9Mo-Al (0.3 mk, 0.568 mk) and UO2 (0.24 mk, 0.448 mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest (135)Xe reactivity during the reactor operation and after shutdown. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

NONE

During this reporting period, there were three major thrusts in the WVU portion. First, we started a preliminary investigation on the use of a membrane reactor for HAS. Accordingly, the plug-flow reactor which had been isolated from sulfides was substituted by a membrane reactor. The tubular membrane was first characterized in terms of its permeation properties, i.e., the fluxes, permeances and selectivities of the components. After that, a BASF methanol-synthesis catalyst was tested under different conditions on the membrane reactor. The results will be compared with those from a non-permeable stainless steel tubular reactor under the same conditions. Second, wemore » started a detailed study of one of the catalysts tested during the screening runs. Accordingly, a carbon-supported potassium-doped molybdenum-cobalt catalyst was selected to be run in the Rotoberty reactor. Finally, we have started detailed analyses of reaction products from some earlier screening runs in which non-sulfide molybdenum-based catalysts were employed and much more complicated product distributions were generally observed. These products could not hitherto be analyzed using the gas chromatograph which was then available. A Varian gas chromatograph/mass spectrometer (GC/MS) is being used to characterize these liquid products. At UCC, we completed a screening of an Engelhard support impregnated with copper and cesium. We have met or exceeded three of four catalyst development targets. Oxygenate selectivity is our main hurdle. Further, we tested the effect of replacing stainless-steel reactor preheater tubing and fittings with titanium ones. We had hoped to reduce the yield of hydrocarbons which may have been produced at high temperatures due to Fischer-Tropsch catalysis with the iron and nickel in the preheater tube walls. Results showed that total hydrocarbon space time yield was actually increased with the titanium preheater, while total alcohol space time yield was not significantly affected.« less

Treatment of raw and ozonated oil sands process-affected water under decoupled denitrifying anoxic and nitrifying aerobic conditions: a comparative study.

PubMed

Xue, Jinkai; Zhang, Yanyan; Liu, Yang; Gamal El-Din, Mohamed

2016-11-01

Batch experiments were performed to evaluate biodegradation of raw and ozonated oil sands process-affected water (OSPW) under denitrifying anoxic and nitrifying aerobic conditions for 33 days. The results showed both the anoxic and aerobic conditions are effective in degrading OSPW classical and oxidized naphthenic acids (NAs) with the aerobic conditions demonstrating higher removal efficiency. The reactors under nitrifying aerobic condition reduced the total classical NAs of raw OSPW by 69.1 %, with better efficiency for species of higher hydrophobicity. Compared with conventional aerobic reactor, nitrifying aerobic condition substantially shortened the NA degradation half-life to 16 days. The mild-dose ozonation remarkably accelerated the subsequent aerobic biodegradation of classical NAs within the first 14 days, especially for those with long carbon chains. Moreover, the ozone pretreatment enhanced the biological removal of OSPW classical NAs by leaving a considerably lower final residual concentration of 10.4 mg/L under anoxic conditions, and 5.7 mg/L under aerobic conditions. The combination of ozonation and nitrifying aerobic biodegradation removed total classical NAs by 76.5 % and total oxy-NAs (O3-O6) by 23.6 %. 454 Pyrosequencing revealed that microbial species capable of degrading recalcitrant hydrocarbons were dominant in all reactors. The most abundant genus in the raw and ozonated anoxic reactors was Thauera (~56 % in the raw OSPW anoxic reactor, and ~65 % in the ozonated OSPW anoxic reactor); whereas Rhodanobacter (~40 %) and Pseudomonas (~40 %) dominated the raw and ozonated aerobic reactors, respectively. Therefore, the combination of mild-dose ozone pretreatment and subsequent biological process could be a competent choice for OSPW treatment.

Predictive characterization of aging and degradation of reactor materials in extreme environments. Final report, December 20, 2013 - September 20, 2017

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

Qu, Jianmin

Understanding of reactor material behavior in extreme environments is vital not only to the development of new materials for the next generation nuclear reactors, but also to the extension of the operating lifetimes of the current fleet of nuclear reactors. To this end, this project conducted a suite of unique experimental techniques, augmented by a mesoscale computational framework, to understand and predict the long-term effects of irradiation, temperature, and stress on material microstructures and their macroscopic behavior. The experimental techniques and computational tools were demonstrated on two distinctive types of reactor materials, namely, Zr alloys and high-Cr martensitic steels. Thesemore » materials are chosen as the test beds because they are the archetypes of high-performance reactor materials (cladding, wrappers, ducts, pressure vessel, piping, etc.). To fill the knowledge gaps, and to meet the technology needs, a suite of innovative in situ transmission electron microscopy (TEM) characterization techniques (heating, heavy ion irradiation, He implantation, quantitative small-scale mechanical testing, and various combinations thereof) were developed and used to elucidate and map the fundamental mechanisms of microstructure evolution in both Zr and Cr alloys for a wide range environmental boundary conditions in the thermal-mechanical-irradiation input space. Knowledge gained from the experimental observations of the active mechanisms and the role of local microstructural defects on the response of the material has been incorporated into a mathematically rigorous and comprehensive three-dimensional mesoscale framework capable of accounting for the compositional variation, microstructural evolution and localized deformation (radiation damage) to predict aging and degradation of key reactor materials operating in extreme environments. Predictions from this mesoscale framework were compared with the in situ TEM observations to validate the model.« less

75 FR 28074 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-19

..., 2009, (74 FR 52829-52830). Wednesday, June 9, 2010, Conference Room T2-B1, Two White Flint North... Regulatory Guide (RG) 1.216, ``Containment Structural Integrity Evaluation for Internal Pressure Loadings... representatives of the NRC staff regarding draft final RG 1.216, ``Containment Structural Integrity Evaluation for...

10 CFR 52.98 - Finality of combined licenses; information requests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design... chapter, as applicable. (b) If the combined license does not reference a design certification or a reactor...

10 CFR 52.98 - Finality of combined licenses; information requests.

Code of Federal Regulations, 2010 CFR

2010-01-01

... condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design... chapter, as applicable. (b) If the combined license does not reference a design certification or a reactor...

Synthesis of calculational methods for design and analysis of radiation shields for nuclear rocket systems

NASA Technical Reports Server (NTRS)

Capo, M. A.; Disney, R. K.; Jordan, T. A.; Soltesz, R. G.; Woodsum, H. C.

1969-01-01

Eight computer programs make up a nine volume synthesis containing two design methods for nuclear rocket radiation shields. The first design method is appropriate for parametric and preliminary studies, while the second accomplishes the verification of a final nuclear rocket reactor design.

A COST BENEFIT APPROACH TO REACTOR SIZING AND NUTRIENT SUPPLY FOR BIOTRICKLING FILTERS FOR AIR POLLUTION CONTROL. (R825392)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

75 FR 54961 - Final Supplemental Environmental Impact Statement, Single Nuclear Unit at the Bellefonte Plant...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-09

... facility, a wind farm, a methane- gas cofiring facility, and several small solar photovoltaic facilities... maintenance of select plant systems and other regulatory compliance activities. Major buildings and plant... the plant cooling towers and the reactor, auxiliary, control, turbine, office, and service buildings...

EXPERIMENTAL INVESTIGATION OF MERCURY TRANSFORMATIONS IN PILOT-SCALE COMBUSTION SYSTEMS AND A BENCH-SCALE ENTRAINED FLOW REACTOR. (R827649)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

POLYMER PYROLYSIS AND OXIDATION STUDIES IN A CONTINUOUS FEED AND FLOW REACTOR: CELLULOSE AND POLYSTYRENE. (R824970)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

78 FR 45984 - Yankee Atomic Electric Company, Yankee Nuclear Power Station

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-30

... Electric Company, Yankee Nuclear Power Station AGENCY: Nuclear Regulatory Commission. ACTION: Environmental... (YAEC) is the holder of Possession-Only License DPR-3 for the Yankee Nuclear Power Station (YNPS... on the site of any nuclear power reactor. In its Statement of Considerations (SOC) for the Final Rule...

10 CFR 52.98 - Finality of combined licenses; information requests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design... chapter, as applicable. (b) If the combined license does not reference a design certification or a reactor...

10 CFR 52.98 - Finality of combined licenses; information requests.

Code of Federal Regulations, 2014 CFR

2014-01-01

... condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design... chapter, as applicable. (b) If the combined license does not reference a design certification or a reactor...

10 CFR 52.98 - Finality of combined licenses; information requests.

Code of Federal Regulations, 2013 CFR

2013-01-01

... condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design... chapter, as applicable. (b) If the combined license does not reference a design certification or a reactor...

77 FR 3288 - Receipt of Request for Action

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-23

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0004] Receipt of Request for Action Notice is hereby given... Director of the Office of Nuclear Reactor Regulation (NRR). As provided by 10 CFR 2.206, appropriate action... review to inform its final decision on whether to implement the requested actions. A copy of the petition...

Numerical investigation of flow and heat transfer in a novel configuration multi-tubular fixed bed reactor for propylene to acrolein process

NASA Astrophysics Data System (ADS)

Jiang, Bin; Hao, Li; Zhang, Luhong; Sun, Yongli; Xiao, Xiaoming

2015-01-01

In the present contribution, a numerical study of fluid flow and heat transfer performance in a pilot-scale multi-tubular fixed bed reactor for propylene to acrolein oxidation reaction is presented using computational fluid dynamics (CFD) method. Firstly, a two-dimensional CFD model is developed to simulate flow behaviors, catalytic oxidation reaction, heat and mass transfer adopting porous medium model on tube side to achieve the temperature distribution and investigate the effect of operation parameters on hot spot temperature. Secondly, based on the conclusions of tube-side, a novel configuration multi-tubular fixed-bed reactor comprising 790 tubes design with disk-and-doughnut baffles is proposed by comparing with segmental baffles reactor and their performance of fluid flow and heat transfer is analyzed to ensure the uniformity condition using molten salt as heat carrier medium on shell-side by three-dimensional CFD method. The results reveal that comprehensive performance of the reactor with disk-and-doughnut baffles is better than that of with segmental baffles. Finally, the effects of operating conditions to control the hot spots are investigated. The results show that the flow velocity range about 0.65 m/s is applicable and the co-current cooling system flow direction is better than counter-current flow to control the hottest temperature.

NuLat: A Novel Design for a Reactor Anti-Neutrino Detector

NASA Astrophysics Data System (ADS)

Rountree, S. Derek; NuLat Collaboration

2015-04-01

NuLat is a proposed very-short baseline (3-10m) reactor electron antineutrino (anti-νe) experiment that will probe the current best fit for light sterile neutrino mixing, the 5 MeV excess seen in current short baseline reactor experiments, and serve as a portable surface detector for cooperative (~ 30m baseline) surface monitoring of reactors. The NuLat detector will use an optically segmented 3D Raghavan optical lattice (ROL) detector that channels light via total internal reflection from a scintillation event down the 3 primary axes to the detector faces. The high degree of segmentation allows for each voxel's energy to be determined independently of other voxels, thus providing high temporal and spatial resolution and energy reconstruction independent of position. NuLat detects anti-νe via inverse beta decay (IBD), which produces a positron and a neutron. Most of the time, the positron deposits its kinetic energy into a single voxel allowing superior derivation of the incident anti-νe's energy. The final state neutron is captured via (n, α) on 6 Li or 10 B after a characteristic delay time giving a coincidence tag. This talk will discuss the physics reach of NuLat using a solid loaded scintillator, and the timeline of the NuLat reactor anti-νe program. This research has been funded in part by the National Science Foundation on Award Numbers 1001394 and 1001078.

Series-Bosch Technology for Oxygen Recovery During Lunar or Martian Surface Missions

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew; Rabenberg, Ellen; Stanley, Christine M.; Edmunson, Jennifer; Alleman, James E.; Chen, Kevin; Dumez, Sam

2014-01-01

Long-duration surface missions to the Moon or Mars will require life support systems that maximize resource recovery to minimize resupply from Earth. To address this need, NASA previously proposed a Series-Bosch (S-Bosch) oxygen recovery system, based on the Bosch process, which can theoretically recover 100% of the oxygen from metabolic carbon dioxide. Bosch processes have the added benefits of the potential to recover oxygen from atmospheric carbon dioxide and the use of regolith materials as catalysts, thereby eliminating the need for catalyst resupply from Earth. In 2012, NASA completed an initial design for an S-Bosch development test stand that incorporates two catalytic reactors in series including a Reverse Water-Gas Shift (RWGS) Reactor and a Carbon Formation Reactor (CFR). In 2013, fabrication of system components, with the exception of a CFR, and assembly of the test stand was initiated. Stand-alone testing of the RWGS reactor was completed to compare performance with design models. Continued testing of Lunar and Martian regolith simulants provided sufficient data to design a CFR intended to utilize these materials as catalysts. Finally, a study was conducted to explore the possibility of producing bricks from spent regolith catalysts. The results of initial demonstration testing of the RWGS reactor, results of continued catalyst performance testing of regolith simulants, and results of brick material properties testing are reported. Additionally, design considerations for a regolith-based CFR are discussed.

Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

DOE PAGES

Skavdahl, Isaac; Utgikar, Vivek; Christensen, Richard; ...

2016-05-24

We present an alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T co) and the hot outlet temperature of the intermediate heat exchanger (T ho2) by manipulating the hot-side flow rates of the heat exchangers (F h/F h2) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the controlmore » of the cold outlet temperature of the SHX (T co) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The final option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.« less

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

Series-Bosch Technology for Oxygen Recovery During Lunar or Martian Surface Missions

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, James M.; Stanley, Christine; Edmunson, Jennifer; Dumez, Samuel; Chen, Kevin; Alleman, James E.

2014-01-01

Long-duration surface missions to the Moon or Mars will require life support systems that maximize resource recovery to minimize resupply from Earth. To address this need, NASA previously proposed a Series-Bosch (S-Bosch) oxygen recovery system, based on the Bosch process, which can theoretically recover 100% of the oxygen from metabolic carbon dioxide. Bosch processes have the added benefits of the potential to recover oxygen from atmospheric carbon dioxide and the use of regolith materials as catalysts, thereby eliminating the need for catalyst resupply from Earth. In 2012, NASA completed an initial design for an S-Bosch development test stand that incorporates two catalytic reactors in series including a Reverse Water-Gas Shift (RWGS) Reactor and a Carbon Formation Reactor (CFR). In 2013, fabrication of system components, with the exception of a CFR, and assembly of the test stand was initiated. Stand-alone testing of the RWGS reactor was completed to compare performance with design models. Continued testing of Lunar and Martian regolith simulants provided sufficient data to design a CFR intended to utilize these materials as catalysts. Finally, a study was conducted to explore the possibility of producing bricks from spend regolith catalysts. The results of initial demonstration testing of the RWGS reactor, results of continued catalyst performance testing of regolith simulants, and results of brick material properties testing are reported. Additionally, design considerations for a regolith-based CFR are discussed.

A microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis.

PubMed

Adam, Zachary R

2016-06-01

Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

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

Diamond, D. J.; Varuttamaseni, A.

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portionmore » of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

A reload and startup plan for conversion of the NIST research reactor

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

D. J. Diamond

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reloadmore » portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

Modelling of slaughterhouse solid waste anaerobic digestion: determination of parameters and continuous reactor simulation.

PubMed

López, Iván; Borzacconi, Liliana

2010-10-01

A model based on the work of Angelidaki et al. (1993) was applied to simulate the anaerobic biodegradation of ruminal contents. In this study, two fractions of solids with different biodegradation rates were considered. A first-order kinetic was used for the easily biodegradable fraction and a kinetic expression that is function of the extracellular enzyme concentration was used for the slowly biodegradable fraction. Batch experiments were performed to obtain an accumulated methane curve that was then used to obtain the model parameters. For this determination, a methodology derived from the "multiple-shooting" method was successfully used. Monte Carlo simulations allowed a confidence range to be obtained for each parameter. Simulations of a continuous reactor were performed using the optimal set of model parameters. The final steady-states were determined as functions of the operational conditions (solids load and residence time). The simulations showed that methane flow peaked at a flow rate of 0.5-0.8 Nm(3)/d/m(reactor)(3) at a residence time of 10-20 days. Simulations allow the adequate selection of operating conditions of a continuous reactor. (c) 2010 Elsevier Ltd. All rights reserved.

On the Effectiveness of Wastewater Cylindrical Reactors: an Analysis Through Steiner Symmetrization

NASA Astrophysics Data System (ADS)

Díaz, J. I.; Gómez-Castro, D.

2016-03-01

The mathematical analysis of the shape of chemical reactors is studied in this paper through the research of the optimization of its effectiveness η such as introduced by R. Aris around 1960. Although our main motivation is the consideration of reactors specially designed for the treatment of wastewaters our results are relevant also in more general frameworks. We simplify the modeling by assuming a single chemical reaction with a monotone kinetics leading to a parabolic equation with a non-necessarily differentiable function. In fact we consider here the case of a single, non-reversible catalysis reaction of chemical order q, 00). We assume the chemical reactor of cylindrical shape Ω =G× (0,H) with G and open regular set of {R}2 not necessarily symmetric. We show that among all the sections G with prescribed area the ball is the set of lowest effectiveness η (t,G). The proof uses the notions of Steiner rearrangement. Finally, we show that if the height H is small enough then the effectiveness can be made as close to 1 as desired.

Continuous production of butanol from starch-based packing peanuts.

PubMed

Ezeji, Thaddeus C; Groberg, Marisa; Qureshi, Nasib; Blaschek, Hans P

2003-01-01

Acetone, butanol, ethanol (ABE, or solvents) were produced from starch-based packing peanuts in batch and continuous reactors. In a batch reactor, 18.9 g/L of total ABE was produced from 80 g/L packing peanuts in 110 h of fermentation. The initial and final starch concentrations were 69.6 and 11.1 g/L, respectively. In this fermentation, ABE yield and productivity of 0.32 and 0.17 g/(L h) were obtained, respectively. Compared to the batch fermentation, continuous fermentation of 40 g/L of starchbased packing peanuts in P2 medium resulted in a maximum solvent production of 8.4 g/L at a dilution rate of 0.033 h-1. This resulted in a productivity of 0.27 g/(L h). However, the reactor was not stable and fermentation deteriorated with time. Continuous fermentation of 35 g/L of starch solution resulted in a similar performance. These studies were performed in a vertical column reactor using Clostridium beijerinckii BA101 and P2 medium. It is anticipated that prolonged exposure of culture to acrylamide, which is formed during boiling/autoclaving of starch, affects the fermentation negatively.

Initial development and performance evaluation of a process for formation of dense carbon by pyrolysis of methane

NASA Technical Reports Server (NTRS)

Noyes, G. P.; Cusick, R. J.

1985-01-01

The three steps in pyrolytic carbon formation are: (1) gaseous hydrocarbon polymerization and aromatic formation; (2) gas-phase condensation and surface adsorption/impingement of polyaromatic hydrocarbon; and (3) final dehydration to carbon. The structure of the carbon in the various stages of formation is examined. The apparatuses and experimental procedures for the pyrolysis of methane in a 60 cm long quartz reactor tube at temperatures ranging from 1400-1600 K are described. The percentage of carbon converted and its density are calculated and tabularly presented. The results reveal that dense carbon formation is maximized and soot eliminated by this procedure. It is observed that conversion efficiency depends on the composition of the inlet gas and conversion increases with increasing temperature. Based on the experimental data a three-man carbon reactor subsystem (CRS) is developed; the functions of the Sabatier Methanation Reactor, two carbon formation reactors and fluid handling components of the CRS are analyzed. The CRS forms 16 kg of carbon at a rate of 0.8 kg/day for 20 days in a two percent volume density quartz wool packing at temperature of 1500-1600 K.

Design and scale-up of a semi-industrial downer-reactor for the rounding of irregular polymer particles

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

Sachs, Marius; Schmidt, Jochen; Peukert, Wolfgang

2016-03-09

The recent development of rapid prototyping technologies towards additive manufacturing reveals some major drawbacks of processes such as laser beam melting (LBM). This contribution focuses on the lack of suitable polymer material with a fine particle size and good flowability. Polymer particles obtained by a wet grinding process 1 are treated in a heated downer reactor. This treatment changes the particles’ morphology from a chiselled state towards a spherical form by surface tension forces in a molten state 2 and leads to an improved flowability. To reach the required amount of rounded polymer powder, a downer reactor in semi-industrial scalemore » has been established and will be characterized in this article. For the purpose of particle rounding it is necessary to avoid contact of molten particles with each other and with the hot reactor walls. Furthermore, the heat distribution has been investigated as one of the key parameters of the process. Finally, a proof of concept by rounding wet grinded PBT material was successfully conducted. The product was examined to obtain data about a change in particle size and flowability.« less

Transmutation of Isotopes --- Ecological and Energy Production Aspects

NASA Astrophysics Data System (ADS)

Gudowski, Waclaw

2000-01-01

This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional nuclear power. In this context a lot of hopes and expectations have been expressed for novel systems called Accelerator-Driven Systems, Accelerator-Driven Transmutation of Waste or just Hybrid Reactors. All these names are used for description of the same nuclear system combining a powerful particle accelerator with a subcritical reactor. A careful analysis of possible environmental impact of ATW together with limitation of this technology is presented also in this paper.

Method of handling radioactive alkali metal waste

DOEpatents

Wolson, Raymond D.; McPheeters, Charles C.

1980-01-01

Radioactive alkali metal is mixed with particulate silica in a rotary drum reactor in which the alkali metal is converted to the monoxide during rotation of the reactor to produce particulate silica coated with the alkali metal monoxide suitable as a feed material to make a glass for storing radioactive material. Silica particles, the majority of which pass through a 95 mesh screen or preferably through a 200 mesh screen, are employed in this process, and the preferred weight ratio of silica to alkali metal is 7 to 1 in order to produce a feed material for the final glass product having a silica to alkali metal monoxide ratio of about 5 to 1.

Method of handling radioactive alkali metal waste

DOEpatents

Wolson, R.D.; McPheeters, C.C.

Radioactive alkali metal is mixed with particulate silica in a rotary drum reactor in which the alkali metal is converted to the monoxide during rotation of the reactor to produce particulate silica coated with the alkali metal monoxide suitable as a feed material to make a glass for storing radioactive material. Silica particles, the majority of which pass through a 95 mesh screen or preferably through a 200 mesh screen, are employed in this process, and the preferred weight ratio of silica to alkali metal is 7 to 1 in order to produce a feed material for the final glass product having a silica to alkali metal monoxide ratio of about 5 to 1.

System Modeling of Lunar Oxygen Production: Mass and Power Requirements

NASA Technical Reports Server (NTRS)

Steffen, Christopher J.; Freeh, Joshua E.; Linne, Diane L.; Faykus, Eric W.; Gallo, Christopher A.; Green, Robert D.

2007-01-01

A systems analysis tool for estimating the mass and power requirements for a lunar oxygen production facility is introduced. The individual modeling components involve the chemical processing and cryogenic storage subsystems needed to process a beneficiated regolith stream into liquid oxygen via ilmenite reduction. The power can be supplied from one of six different fission reactor-converter systems. A baseline system analysis, capable of producing 15 metric tons of oxygen per annum, is presented. The influence of reactor-converter choice was seen to have a small but measurable impact on the system configuration and performance. Finally, the mission concept of operations can have a substantial impact upon individual component size and power requirements.

The benefits of a fast reactor closed fuel cycle in the UK

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

Gregg, R.; Hesketh, K.

2013-07-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size,more » so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)« less

Pyrolysis of tyres. Influence of the final temperature of the process on emissions and the calorific value of the products recovered.

PubMed

Díez, C; Martínez, O; Calvo, L F; Cara, J; Morán, A

2004-01-01

A study was made of the pyrolysis of tyre particles, with the aim of determining the possibilities of using the products resulting from the process as fuel. Three final temperatures were used, determined from thermogravimetric data. The design of the experiment was a horizontal oven containing a reactor into which particles of the original tyre were placed. After the process, a solid fraction (char) remained in the reactor, while the gases generated went through a set of scrubbers where most of the condensable fraction (oils) was retained. Finally, once free of this fraction, the gases were collected in glass ampoules. Solid and liquids fractions were subjected to thermogravimetric analyses in order to study their combustibility. The gas fraction was analysed by means of gas chromatography to establish the content of CO, CO2, H2 and hydrocarbons present in the samples (mainly components of gases produced in the pyrolysis process). A special study was made of the sulphur and chlorine content of all the fractions, as the presence of these elements could be problematic if the products are used as fuel. Tyre pyrolysis engenders a solid carbon residue that concentrates sulphur and chorine, with a relatively high calorific value, although not so high as that of the original tyre. The liquid fraction produced by the process has a high calorific value, which rises with the final temperature, up to 40 MJ/kg. The chlorine content of this fraction is negligible. Over 95% of the gas fraction, regardless of the final temperature, is composed of hydrocarbons of a low molecular weight and hydrogen, this fraction also appearing to be free of chlorine.

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

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

Hawari, Ayman; Ougouag, Abderrafi

2014-07-08

This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can bemore » easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.« less

Aeration costs in stirred-tank and bubble column bioreactors

DOE PAGES

Humbird, D.; Davis, R.; McMillan, J. D.

2017-08-10

To overcome knowledge gaps in the economics of large-scale aeration for production of commodity products, Aspen Plus is used to simulate steady-state oxygen delivery in both stirred-tank and bubble column bioreactors, using published engineering correlations for oxygen mass transfer as a function of aeration rate and power input, coupled with new equipment cost estimates developed in Aspen Capital Cost Estimator and validated against vendor quotations. Here, these simulations describe the cost efficiency of oxygen delivery as a function of oxygen uptake rate and vessel size, and show that capital and operating costs for oxygen delivery drop considerably moving from standard-sizemore » (200 m 3) to world-class size (500 m 3) reactors, but only marginally in further scaling up to hypothetically large (1000 m 3) reactors. Finally, this analysis suggests bubble-column reactor systems can reduce overall costs for oxygen delivery by 10-20% relative to stirred tanks at low to moderate oxygen transfer rates up to 150 mmol/L-h.« less

Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

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

J. Benson; J. Cole; J. Jackson

2013-02-01

In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groupsmore » conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.« less

A facile and efficient method of enzyme immobilization on silica particles via Michael acceptor film coatings: immobilized catalase in a plug flow reactor.

PubMed

Bayramoglu, Gulay; Arica, M Yakup; Genc, Aysenur; Ozalp, V Cengiz; Ince, Ahmet; Bicak, Niyazi

2016-06-01

A novel method was developed for facile immobilization of enzymes on silica surfaces. Herein, we describe a single-step strategy for generating of reactive double bonds capable of Michael addition on the surfaces of silica particles. This method was based on reactive thin film generation on the surfaces by heating of impregnated self-curable polymer, alpha-morpholine substituted poly(vinyl methyl ketone) p(VMK). The generated double bonds were demonstrated to be an efficient way for rapid incorporation of enzymes via Michael addition. Catalase was used as model enzyme in order to test the effect of immobilization methodology by the reactive film surface through Michael addition reaction. Finally, a plug flow type immobilized enzyme reactor was employed to estimate decomposition rate of hydrogen peroxide. The highly stable enzyme reactor could operate continuously for 120 h at 30 °C with only a loss of about 36 % of its initial activity.

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

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

Mcwilliams, A. J.

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniquesmore » through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.« less

Atmosphere Processing Module Automation and Catalyst Durability Analysis for Mars ISRU Pathfinder

NASA Technical Reports Server (NTRS)

Petersen, Elspeth M.

2016-01-01

The Mars In-Situ Resource Utilization Pathfinder was designed to create fuel using components found in the planet’s atmosphere and regolith for an ascension vehicle to return a potential sample return or crew return vehicle from Mars. The Atmosphere Processing Module (APM), a subunit of the pathfinder, uses cryocoolers to isolate and collect carbon dioxide from Mars simulant gas. The carbon dioxide is fed with hydrogen into a Sabatier reactor where methane is produced. The APM is currently undergoing the final stages of testing at Kennedy Space Center prior to process integration testing with the other subunits of the pathfinder. The automation software for the APM cryocoolers was tested and found to perform nominally. The catalyst used for the Sabatier reactor was investigated to determine the factors contributing to catalyst failure. The results from the catalyst testing require further analysis, but it appears that the rapid change in temperature during reactor start up or the elevated operating temperature is responsible for the changes observed in the catalyst.

Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

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

Lee, D.D.; Collins, J.L.

One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test usingmore » the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.« less

Nuclear power plant 5,000 to 10,000 kilowatts

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

Not Available

The purpose of this proposal is to present a suggested program for the development of an Aqueous Homogeneous Reactor Power Plant for the production of power in the 5000 to 10,000 kilowatt range under the terms of the Atomic Energy Commission's invitation of September 21, 1955. It envisions a research and development program prior to finalizing fabricating commitments of full scale components for the purpose of proving mechanical and hydraulic operating and chemical processing feasibility with the expectation that such preliminary effort will assure the contruction of the reactor at the lowest cost and successful operation at the earliest date.more » It proposes the construction of a reactor for an eventual net electrical output of ten megawatts but initially in conjunction with a five megawatt turbo-generating unit. This unit would be constructed at the site of the existing Hersey diesel generating plant of the Wolverine Electric Cooperative approximately ten miles north of Big Rapids, Michigan.« less

Aeration costs in stirred-tank and bubble column bioreactors

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

Humbird, D.; Davis, R.; McMillan, J. D.

To overcome knowledge gaps in the economics of large-scale aeration for production of commodity products, Aspen Plus is used to simulate steady-state oxygen delivery in both stirred-tank and bubble column bioreactors, using published engineering correlations for oxygen mass transfer as a function of aeration rate and power input, coupled with new equipment cost estimates developed in Aspen Capital Cost Estimator and validated against vendor quotations. Here, these simulations describe the cost efficiency of oxygen delivery as a function of oxygen uptake rate and vessel size, and show that capital and operating costs for oxygen delivery drop considerably moving from standard-sizemore » (200 m 3) to world-class size (500 m 3) reactors, but only marginally in further scaling up to hypothetically large (1000 m 3) reactors. Finally, this analysis suggests bubble-column reactor systems can reduce overall costs for oxygen delivery by 10-20% relative to stirred tanks at low to moderate oxygen transfer rates up to 150 mmol/L-h.« less

76 FR 17159 - Office of New Reactors; Final Interim Staff Guidance on Standard Review Plan, Section 17.4...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-28

... Interim Staff Guidance on Standard Review Plan, Section 17.4, ``Reliability Assurance Program'' AGENCY... design reliability assurance program (RAP). This ISG updates the guidance provided to the staff in Standard Review Plan (SRP), Section 17.4, ``Reliability Assurance Program,'' of NUREG-0800, ``Standard...

76 FR 23340 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-26

... Thursday, October 21, 2010 (75 FR 65038-65039). Thursday, May 12, 2011, Conference Room T2-B1, 11545... Safety Evaluation Report Associated with the License Renewal Application for the Hope Creek Generating... NRC staff and PSEG Nuclear, LLC regarding the final safety evaluation report (SER) associated with the...

75 FR 81317 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-27

... Thursday, October 21, 2010 (74 FR 65038-65039). Thursday, January 13, 2011, Conference Room T2-B1, Two... Evaluation Report Associated With the Vogtle Units 3 and 4 Combined License Application (Open/Closed)-- The... Nuclear Company, and NuStar Energy regarding the Final Safety Evaluation Report associated with the Vogtle...

78 FR 33120 - Final Interim Staff Guidance LR-ISG-2011-04; Updated Aging Management Criteria for Reactor Vessel...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

..., ``Generic Aging Lessons Learned Report'' (GALL Report), for the aging management of Pressurized Water... communicate insights and lessons learned and to address emergent issues not covered in license renewal... ensure that PWR license renewal applicants will adequately address age-related degradation and aging...

10 CFR 52.157 - Contents of applications; technical information in final safety analysis report.

Code of Federal Regulations, 2010 CFR

2010-01-01

... analysis of the structures, systems, and components of the reactor to be manufactured, with emphasis upon... assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site... structures, systems, and components with the objective of assessing the risk to public health and safety...

Biodrying process: A sustainable technology for treatment of municipal solid waste with high moisture content.

PubMed

Tom, Asha P; Pawels, Renu; Haridas, Ajit

2016-03-01

Municipal solid waste with high moisture content is the major hindrance in the field of waste to energy conversion technologies and here comes the importance of biodrying process. Biodrying is a convective evaporation process, which utilizes the biological heat developed from the aerobic reactions of organic components. The numerous end use possibilities of the output are making the biodrying process versatile, which is possible by achieving the required moisture reduction, volume reduction and bulk density enhancement through the effective utilization of biological heat. In the present case study the detailed research and development of an innovative biodrying reactor has been carried out for the treatment of mixed municipal solid waste with high moisture content. A pilot scale biodrying reactor of capacity 565 cm(3) was designed and set up in the laboratory. The reactor dimensions consisted of an acrylic chamber of 60 cm diameter and 200 cm height, and it was enveloped by an insulation chamber. The insulation chamber was provided to minimise the heat losses through the side walls of the reactor. It simulates the actual condition in scaling up of the reactor, since in bigger scale reactors the heat losses through side walls will be negligible while comparing the volume to surface area ratio. The mixed municipal solid waste with initial moisture content of 61.25% was synthetically prepared in the laboratory and the reactor was fed with 109 kg of this substrate. Aerobic conditions were ensured inside the reactor chamber by providing the air at a constant rate of 40 litre per minute, and the direction of air flow was from the specially designed bottom air chamber to the reactor matrix top. The self heating inside reactor matrix was assumed in the range of 50-60°C during the design stage. Innovative biodrying reactor was found to be efficiently working with the temperature inside the reactor matrix rising to a peak value of 59°C by the fourth day of experiment (the peak observed at a height of 60 cm from the air supply). The process analyses results were promising with a reduction of 56.5% of volume, and an increase of 52% of bulk density of the substrate at the end of 33 days of biodrying. Also the weight of mixed MSW substrate has been reduced by 33.94% in 20 days of reaction and the average moisture reduction of the matrix was 20.81% (reduced from the initial value of 61.25% to final value of 48.5%). The moisture reduction would have been higher, if the condensation of evaporated water at the reactor matrix has been avoided. The non-homogeneous moisture reduction along the height of the reactor is evident and this needs further innovation. The leachate production has been completely eliminated in the innovative biodrying reactor and that is a major achievement in the field of municipal solid waste management technology. Copyright © 2016 Elsevier Ltd. All rights reserved.

Calibration factors for the SNOOPY NP-100 neutron dosimeter

NASA Astrophysics Data System (ADS)

Moscu, D. F.; McNeill, F. E.; Chase, J.

2007-10-01

Within CANDU nuclear power facilities, only a small fraction of workers are exposed to neutron radiation. For these individuals, roughly 4.5% of the total radiation equivalent dose is the result of exposure to neutrons. When this figure is considered across all workers receiving external exposure of any kind, only 0.25% of the total radiation equivalent dose is the result of exposure to neutrons. At many facilities, the NP-100 neutron dosimeter, manufactured by Canberra Industries Incorporated, is employed in both direct and indirect dosimetry methods. Also known as "SNOOPY", these detectors undergo calibration, which results in a calibration factor relating the neutron count rate to the ambient dose equivalent rate, using a standard Am-Be neutron source. Using measurements presented in a technical note, readings from the dosimeter for six different neutron fields in six source-detector orientations were used, to determine a calibration factor for each of these sources. The calibration factor depends on the neutron energy spectrum and the radiation weighting factor to link neutron fluence to equivalent dose. Although the neutron energy spectra measured in the CANDU workplace are quite different than that of the Am-Be calibration source, the calibration factor remains constant - within acceptable limits - regardless of the neutron source used in the calibration; for the specified calibration orientation and current radiation weighting factors. However, changing the value of the radiation weighting factors would result in changes to the calibration factor. In the event of changes to the radiation weighting factors, it will be necessary to assess whether a change to the calibration process or resulting calibration factor is warranted.

Characterization and modelling of the heat transfers in a pilot-scale reactor during composting under forced aeration.

PubMed

de Guardia, A; Petiot, C; Benoist, J C; Druilhe, C

2012-06-01

The paper focused on the modelling of the heat transfers during composting in a pilot-scale reactor under forced aeration. The model took into account the heat production and the transfers by evaporation, convection between material and gas crossing the material, conduction and surface convection between gas and material in bottom and upper parts of the reactor. The model was adjusted thanks to the measurements practised during fifteen composting experiments in which five organic wastes were, each, composted under three constant aeration rates. Heat production was considered proportional to oxygen consumption rate and the enthalpy per mole oxygen consumed was assumed constant. The convective heat transfer coefficients were determined on basis of the continuous measurements of the temperatures of both the lid and the bottom part of the reactor. The model allowed a satisfying prediction of the temperature of the composting material. In most cases, the mean absolute discard between the experimental and the simulated temperatures was inferior to 2.5°C and the peaks of temperature occurred with less than 8h delay. For the half of the experiments the temperature discard between the simulated peak and the experimental one was inferior to 5°C. On basis of the calculation of a stoichiometric production of water through oxidation of the biodegradable organic matter, the simulation of water going out from material as vapour also allowed a rather satisfying prediction of the mass of water in final mixture. The influence of the aeration rate on every type of heat loss was characterized. Finally, the model was used to evaluate the impacts on material temperature caused by the change of the insulation thickness, the ambient temperature, take the lid away, the increase or the decrease of the mass of waste to compost. Copyright © 2012 Elsevier Ltd. All rights reserved.

Characterization and modelling of the heat transfers in a pilot-scale reactor during composting under forced aeration

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

Guardia, A. de, E-mail: amaury.de-guardia@irstea.fr; Universite Europeenne de Bretagne, F-35000 Rennes; Petiot, C.

2012-06-15

The paper focused on the modelling of the heat transfers during composting in a pilot-scale reactor under forced aeration. The model took into account the heat production and the transfers by evaporation, convection between material and gas crossing the material, conduction and surface convection between gas and material in bottom and upper parts of the reactor. The model was adjusted thanks to the measurements practised during fifteen composting experiments in which five organic wastes were, each, composted under three constant aeration rates. Heat production was considered proportional to oxygen consumption rate and the enthalpy per mole oxygen consumed was assumedmore » constant. The convective heat transfer coefficients were determined on basis of the continuous measurements of the temperatures of both the lid and the bottom part of the reactor. The model allowed a satisfying prediction of the temperature of the composting material. In most cases, the mean absolute discard between the experimental and the simulated temperatures was inferior to 2.5 Degree-Sign C and the peaks of temperature occurred with less than 8 h delay. For the half of the experiments the temperature discard between the simulated peak and the experimental one was inferior to 5 Degree-Sign C. On basis of the calculation of a stoichiometric production of water through oxidation of the biodegradable organic matter, the simulation of water going out from material as vapour also allowed a rather satisfying prediction of the mass of water in final mixture. The influence of the aeration rate on every type of heat loss was characterized. Finally, the model was used to evaluate the impacts on material temperature caused by the change of the insulation thickness, the ambient temperature, take the lid away, the increase or the decrease of the mass of waste to compost.« less

Comparative performance of fixed-film biological filters: Application of reactor theory

USGS Publications Warehouse

Watten, B.J.; Sibrell, P.L.

2006-01-01

Nitrification is classified as a two-step consecutive reaction where R1 represents the rate of formation of the intermediate product NO2-N and R2 represents the rate of formation of the final product NO3-N. The relative rates of R1 and R2 are influenced by reactor type characterized hydraulically as plug-flow, plug-flow with dispersion and mixed-flow. We develop substrate conversion models for fixed-film biofilters operating in the first-order kinetic regime based on application of chemical reactor theory. Reactor type, inlet conditions and the biofilm kinetic constants Ki (h-1) are used to predict changes in NH4-N, NO2-N, NO3-N and BOD5. The inhibiting effects of the latter on R1 and R2 were established based on the ?? relation, e.g.:{A formula is presented}where BOD5,max is the concentration that causes nitrification to cease and N is a variable relating Ki to increasing BOD5. Conversion models were incorporated in spreadsheet programs that provided steady-state concentrations of nitrogen and BOD5 at several points in a recirculating aquaculture system operating with input values for fish feed rate, reactor volume, microscreen performance, make-up and recirculating flow rates. When rate constants are standardized, spreadsheet use demonstrates plug-flow reactors provide higher rates of R1 and R2 than mixed-flow reactors thereby reducing volume requirements for target concentrations of NH4-N and NO2-N. The benefit provided by the plug-flow reactor varies with hydraulic residence time t as well as the effective vessel dispersion number, D/??L. Both reactor types are capable of providing net increases in NO2-N during treatment but the rate of decrease in the mixed-flow case falls well behind that predicted for plug-flow operation. We show the potential for a positive net change in NO2-N increases with decreases in the dimensionless ratios K2, (R2 )/K1,( R1 ) and [NO2-N]/[NH4-N] and when the product K1, (R1) t provides low to moderate NH4-N conversions. Maintaining high levels of the latter reduces the effective reactor utilization rate (%) defined here as (RNavg/RNmax)100 where RNavg is the mean reactive nitrogen concentration ([NH4-N] + [NO2-N]) within the reactor, and RNmax represents the feed concentration of the same. Low utilization rates provide a hedge against unexpected increases in substrate loading and reduce water pumping requirements but force use of elevated reactor volumes. Further ?? effects on R1 and R2 can be reduced through use of a tanks-in-series versus a single mixed-flow reactor configuration and by improving the solids removal efficiency of microscreen treatment.

Porous media for catalytic renewable energy conversion

NASA Astrophysics Data System (ADS)

Hotz, Nico

2012-05-01

A novel flow-based method is presented to place catalytic nanoparticles into a reactor by sol-gelation of a porous ceramic consisting of copper-based nanoparticles, silica sand, ceramic binder, and a gelation agent. This method allows for the placement of a liquid precursor containing the catalyst into the final reactor geometry without the need of impregnating or coating of a substrate with the catalytic material. The so generated foam-like porous ceramic shows properties highly appropriate for use as catalytic reactor material, e.g., reasonable pressure drop due to its porosity, high thermal and catalytic stability, and excellent catalytic behavior. The catalytic activity of micro-reactors containing this foam-like ceramic is tested in terms of their ability to convert alcoholic biofuel (e.g. methanol) to a hydrogen-rich gas mixture with low concentrations of carbon monoxide (up to 75% hydrogen content and less than 0.2% CO, for the case of methanol). This gas mixture is subsequently used in a low-temperature fuel cell, converting the hydrogen directly to electricity. A low concentration of CO is crucial to avoid poisoning of the fuel cell catalyst. Since conventional Polymer Electrolyte Membrane (PEM) fuel cells require CO concentrations far below 100 ppm and since most methods to reduce the mole fraction of CO (such as Preferential Oxidation or PROX) have CO conversions of up to 99%, the alcohol fuel reformer has to achieve initial CO mole fractions significantly below 1%. The catalyst and the porous ceramic reactor of the present study can successfully fulfill this requirement.

Evaluation of A Novel Split-Feeding Anaerobic/Oxic Baffled Reactor (A/OBR) For Foodwaste Anaerobic Digestate: Performance, Modeling and Bacterial Community

PubMed Central

Wang, Shaojie; Peng, Liyu; Jiang, Yixin; Gikas, Petros; Zhu, Baoning; Su, Haijia

2016-01-01

To enhance the treatment efficiency from an anaerobic digester, a novel six-compartment anaerobic/oxic baffled reactor (A/OBR) was employed. Two kinds of split-feeding A/OBRs R2 and R3, with influent fed in the 1st, 3rd and 5th compartment of the reactor simultaneously at the respective ratios of 6:3:1 and 6:2:2, were compared with the regular-feeding reactor R1 when all influent was fed in the 1st compartment (control). Three aspects, the COD removal, the hydraulic characteristics and the bacterial community, were systematically investigated, compared and evaluated. The results indicated that R2 and R3 had similar tolerance to loading shock, but the R2 had the highest COD removal of 91.6% with a final effluent of 345 mg/L. The mixing patterns in both split-feeding reactors were intermediate between plug-flow and completely-mixed, with dead spaces between 8.17% and 8.35% compared with a 31.9% dead space in R1. Polymerase chain reaction-denaturing gradient gel electrophoresis (PCR-DGGE) analysis revealed that the split-feeding strategy provided a higher bacterial diversity and more stable bacterial community than that in the regular-feeding strategy. Further analysis indicated that Firmicutes, Bacteroidetes, and Proteobacteria were the dominant bacteria, among which Firmicutes and Bacteroidetes might be responsible for organic matter degradation and Proteobacteria for nitrification and denitrification. PMID:27708368

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

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

Mertyurek, Ugur; Gauld, Ian C.

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE PAGES

Mertyurek, Ugur; Gauld, Ian C.

2015-12-24

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Production of aluminum--silicon alloy and ferrosilicon and commercial purity aluminum by the direct reduction process. First interim technical report, Phase B, September 1--November 30, 1978

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

Bruno, M.J.

1978-12-01

The parameters of charge content, reaction temperatures and residence time were studied in a bench reactor concerning the production of Al--Si and Fe--Si alloys. Results confirmed that minimum final stage reaction temperature is 1950 to 2000/sup 0/C. Residence time varied with initial charge concentration. Fe/sub 2/O/sub 3/ additions to the charge produced a significant increase in metallic yield. A burden preparation procedure was developed for making acceptable agglomerates containing Fe/sub 2/O/sub 3/, bauxite, clay and coke. Particle size distribution of starting materials was correlated with agglomerate strength. A new bench scale reactor was designed and built to facilitate semi-continuous operation,more » using O/sub 2/ injection to burn coke supporting the burden, resulting in burden movement. In a number of runs bridging of the burden material occurred due to condensation of volatilized sub-oxides in the cooler zones of the reactor. The reactor operated smoothly as an iron blast furnace at 1500/sup 0/C, demonstrating the validity of the equipment and test procedures. Initial construction of pilot reactor VSR-1 was completed. Bench scale fractional crystallizer runs were continued to determine the impurity effects of Fe up to 6.9% and Ti up to 1.25% on alloy product concentration and yield. High levels of impurities formed intermetallic complexes with Al and reduced product yield.« less

Ichthyoplankton entrainment study at the SRS Savannah River water intakes for Westinghouse Savannah River Company. Final report

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

Paller, M.

1992-03-26

Cooling water for L and K Reactors and makeup water for Par Pond is pumped from the Savannah River at the 1G, 3G, and 5G pump houses. Ichthyoplankton (drifting fish larvae and eggs) from the river are entrained into the reactor cooling systems with the river water and passed through the reactor`s heat exchangers where temperatures may reach 70{degrees}C during full power operation. Ichthyoplankton mortality under such conditions is assumed to be 100 percent. The number of ichthyoplankton entrained into the cooling system depends on a variety of variables, including time of year, density and distribution of ichthyoplankton in themore » river, discharge levels in the river, and the volume of water withdrawn by the pumps. Entrainment at the 1 G pump house, which is immediately downstream from the confluence of Upper Three Runs Creek and the Savannah River, is also influenced by discharge rates and ichthyoplankton densities in Upper Three Runs Creek. Because of the anticipated restart of several SRS reactors and the growing concern surrounding striped bass and American shad stocks in the Savannah River, the Department of Energy requested that the Environmental Sciences Section (ESS) of the Savannah River Laboratory sample ichthyoplankton at the SRS Savannah River intakes. Dams & Moore, Inc., under a contract with Westinghouse Savannah River Company performed the sampling and data analysis for the ESS.« less

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

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

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE PAGES

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; ...

2017-02-26

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

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

E. Blanford; E. Keldrauk; M. Laufer

2010-09-20

Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement,more » and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using factory prefabricated structural modules, for application to external event shell and base isolated structures.« less

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

NASA Astrophysics Data System (ADS)

Andrianova, E. A.; Tsibul'skiy, V. F.

2017-12-01

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and 238U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with 238U alone should be reprocessed in the usual way.

Noble gas, iodine, and cesium transport in a postulated loss of decay heat removal accident at Browns Ferry

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

Wichner, R.P.; Hodge, S.A.; Weber, C.F.

1984-08-01

This report presents an analysis of the movement of noble gas, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto themore » drywell floor. The analysis of fission product transport presented in this report is based on the no-operator-action sequence and provides an estimate of fission product inventories, as a function of time, within 14 control volumes outside the core, with the atmosphere considered as the final control volume in the transport sequence. As in the case of accident sequences previously studied, we find small barrier for noble gas ejection to air, these gases being effectively purged from the drywell and reactor building by steam and concrete degradation gases. However, significant decay of krypton isotopes occurs during the long delay times involved in this sequence. In contrast, large degrees of holdup for iodine and cesium are projected due to the chemical reactivity of these elements. Only about 2 x 10/sup -4/% of the initial iodine and cesium activity are predicted to be released to the atmosphere. Principal barriers for release are deposition on reactor vessel and containment walls. A significant amount of iodine is captured in the water pool formed in the reactor building basement after actuation of the fire protection system.« less

Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux

NASA Astrophysics Data System (ADS)

Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; Boukhal, H.; Jaï, Otman

2018-06-01

The MNSR is a pool type research reactor, which is difficult to model because of the importance of neutron leakage. The aim of this study is to evaluate a 2-D transport model for the reactor compatible with the latest release of the DRAGON code and 3-D diffusion of the DONJON code. DRAGON code is then used to generate the group macroscopic cross sections needed for full core diffusion calculations. The diffusion DONJON code, is then used to compute the effective multiplication factor (keff), the feedback reactivity coefficients and neutron flux which account for variation in fuel and moderator temperatures as well as the void coefficient have been calculated using the DRAGON and DONJON codes for the MNSR research reactor. The cross sections of all the reactor components at different temperatures were generated using the DRAGON code. These group constants were used then in the DONJON code to calculate the multiplication factor and the neutron spectrum at different water and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. Finally, Good agreements between the calculated and measured have been obtained for every of the feedback reactivity coefficients and neutron flux.

Novel arrangement for an electro-Fenton reactor that does not require addition of iron, acid and a final neutralization stage. Towards the development of a cost-effective technology for the treatment of wastewater.

PubMed

Fernández, Dennys; Robles, Irma; Rodríguez-Valadez, Francisco J; Godínez, Luis A

2018-05-01

A novel arrangement for an electro-Fenton reactor aimed to treat neutral wastewater is presented. The arrangement consists on three-compartments in series, two of them packed with a cation exchange resin and one positioned between these, containing a polarized activated carbon column where the electrochemical generation of the Fenton reagent takes place. While the hydroxyl radicals electrochemically produced in-situ, react with the pollutant species adsorbed on the activated carbon cathode, the resin compartments administrate and collect the iron cation and the hydrated proton species in alternating flow direction cycles. The resulting process is a system that does not require acid or iron chemical addition to the process while at the same time, renders decontaminated water free of iron-dissolved species at neutral pH. The proposed electrochemical reactor arrangement is therefore the basis for the design of commercially viable electro-Fenton reactors in which the addition and subsequent removal of acid and iron chemicals is avoided; two of the currently most limiting features for the development of electro-Fenton technology for treating wastewater. Copyright © 2018 The Authors. Published by Elsevier Ltd.. All rights reserved.

Biocarriers Improve Bioaugmentation Efficiency of a Rapid Sand Filter for the Treatment of 2,6-Dichlorobenzamide-Contaminated Drinking Water.

PubMed

Horemans, Benjamin; Raes, Bart; Vandermaesen, Johanna; Simanjuntak, Yanti; Brocatus, Hannelore; T'Syen, Jeroen; Degryse, Julie; Boonen, Jos; Wittebol, Janneke; Lapanje, Ales; Sørensen, Sebastian R; Springael, Dirk

2017-02-07

Aminobacter sp. MSH1 immobilized in an alginate matrix in porous stones was tested in a pilot system as an alternative inoculation strategy to the use of free suspended cells for biological removal of micropollutant concentrations of 2,6-dichlorobenzamide (BAM) in drinking water treatment plants (DWTPs). BAM removal rates and MSH1 cell numbers were recorded during operation and assessed with specific BAM degradation rates obtained in lab conditions using either freshly grown cells or starved cells to explain reactor performance. Both reactors inoculated with either suspended or immobilized cells showed immediate BAM removal under the threshold of 0.1 μg/L, but the duration of sufficient BAM removal was 2-fold (44 days) longer for immobilized cells. The longer sufficient BAM removal in case of immobilized cells compared to suspended cells was mainly explained by a lower initial loss of MSH1 cells at operational start due to volume replacement and shear. Overall loss of activity in the reactors though was due to starvation, and final removal rates did not differ between reactors inoculated with immobilized and suspended cells. Management of assimilable organic carbon, in addition to cell immobilization, appears crucial for guaranteeing long-term BAM degradation activity of MSH1 in DWTP units.

Summary of ORSphere critical and reactor physics measurements

NASA Astrophysics Data System (ADS)

Marshall, Margaret A.; Bess, John D.

2017-09-01

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

Radiogenic lead as coolant, reflector and moderator in advanced fast reactors

NASA Astrophysics Data System (ADS)

Kulikov, E. G.

2017-01-01

Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

FROM CONCEPT TO REALITY, IN-SITU DECOMMISSIONING OF THE P AND R REACTORS AT THE SAVANNAH RIVER SITE

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

Musall, J.; Blankenship, J.; Griffin, W.

2012-01-09

SRS recently completed an approximately three year effort to decommission two SRS reactors: P-Reactor (Building 105-P) and R-Reactor (Building 105-R). Completed in December 2011, the concurrent decommissionings marked the completion of two relatively complex and difficult facility disposition projects at the SRS. Buildings 105-P and 105-R began operating as production reactors in the early 1950s with the mission of producing weapons material (e.g., tritium and plutonium-239). The 'P' Reactor and was shutdown in 1991 while the 'R' Reactor and was shutdown in 1964. In the intervening period between shutdown and deactivation & decommissioning (D&D), Buildings 105-P and 105-R saw limitedmore » use (e.g., storage of excess heavy water and depleted uranium oxide). For Building 105-P, deactivation was initiated in April 2007 and was essentially complete by June 2010. For Building 105-R, deactivation was initiated in August 2008 and was essentially complete by September 2010. For both buildings, the primary objective of deactivation was to remove/mitigate hazards associated with the remaining hazardous materials, and thus prepare the buildings for in-situ decommissioning. Deactivation removed the following hazardous materials to the extent practical: combustibles/flammables, residual heavy water, acids, friable asbestos (as needed to protect workers performing deactivation and decommissioning), miscellaneous chemicals, lead/brass components, Freon(reg sign), oils, mercury/PCB containing components, mold and some radiologically-contaminated equipment. In addition to the removal of hazardous materials, deactivation included the removal of hazardous energy, exterior metallic components (representing an immediate fall hazard), and historical artifacts along with the evaporation of water from the two Disassembly Basins. Finally, so as to facilitate occupancy during the subsequent in-situ decommissioning, deactivation implemented repairs to the buildings and provided temporary power.« less

U.S.-Russian Cooperation in Science and Technology: A Case Study of the TOPAZ Space-Based Nuclear Reactor International Program

NASA Astrophysics Data System (ADS)

Dabrowski, Richard S.

2014-08-01

The TOPAZ International Program (TIP) was the final name given to a series of projects to purchase and test the TOPAZ-II, a space-based nuclear reactor of a type that had been further developed in the Soviet Union than in the United States. In the changing political situation associated with the break-up of the Soviet Union it became possible for the United States to not just purchase the system, but also to employ Russian scientists, engineers and testing facilities to verify its reliability. The lessons learned from the TIP illuminate some of the institutional and cultural challenges to U.S. - Russian cooperation in technology research which remain true today.

Status and improvement of CLAM for nuclear application

NASA Astrophysics Data System (ADS)

Huang, Qunying

2017-08-01

A program for China low activation martensitic steel (CLAM) development has been underway since 2001 to satisfy the material requirements of the test blanket module (TBM) for ITER, China fusion engineering test reactor and China fusion demonstration reactor. It has been undertaken by the Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences under wide domestic and international collaborations. Extensive work and efforts are being devoted to the R&D of CLAM, such as mechanical property evaluation before and after neutron irradiation, fabrication of scaled TBM by welding and additive manufacturing, improvement of its irradiation resistance as well as high temperature properties by precipitate strengthening to achieve its final successful application in fusion systems. The status and improvement of CLAM are introduced in this paper.

Theoretical and experimental study of the dark signal in CMOS image sensors affected by neutron radiation from a nuclear reactor

NASA Astrophysics Data System (ADS)

Xue, Yuanyuan; Wang, Zujun; He, Baoping; Yao, Zhibin; Liu, Minbo; Ma, Wuying; Sheng, Jiangkun; Dong, Guantao; Jin, Junshan

2017-12-01

The CMOS image sensors (CISs) are irradiated with neutron from a nuclear reactor. The dark signal in CISs affected by neutron radiation is studied theoretically and experimentally. The Primary knock-on atoms (PKA) energy spectra for 1 MeV incident neutrons are simulated by Geant4. And the theoretical models for the mean dark signal, dark signal non-uniformity (DSNU) and dark signal distribution versus neutron fluence are established. The results are found to be in good agreement with the experimental outputs. Finally, the dark signal in the CISs under the different neutron fluence conditions is estimated. This study provides the theoretical and experimental evidence for the displacement damage effects on the dark signal CISs.

Experimental Neutrino Physics: Final Report

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

Lane, Charles E.; Maricic, Jelena

2012-09-05

Experimental studies of neutrino properties, with particular emphasis on neutrino oscillation, mass and mixing parameters. This research was pursued by means of underground detectors for reactor anti-neutrinos, measuring the flux and energy spectra of the neutrinos. More recent investigations have been aimed and developing detector technologies for a long-baseline neutrino experiment (LBNE) using a neutrino beam from Fermilab.

Study, Talk, and Action. A Report of a National Conference on Regionalism and Regionalization in American Postsecondary Education.

ERIC Educational Resources Information Center

Martorana, S. V., Ed.; Nespoli, Lawrence A., Ed.

This report of a National Conference on Regionalism and Regionalization in American Postsecondary Education contains an overview and summary of the final project report, a keynote address, four papers on the implications of regionalism, some reactor comments, an essay on leadership, and four descriptive accounts of operational regionalization…

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2010 CFR

2010-01-01

... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF... Work Authorizations Early Partial Decisions on Site Suitability-Combined License Under 10 Cfr Part 52... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or...

75 FR 5356 - Office of New Reactors; Final Interim Staff Guidance on Post-Combined License Commitments

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-02

... the guidance provided to the NRC staff in Section 1.0, ``Introduction and Interfaces,'' of NUREG-0800... (COL) applications. In addition, this ISG supplements the guidance provided in Section C.III.4 of... Accession No. ML091671355) to solicit public and industry comment. The NRC staff received comments (ADAMS...

BOILING NUCLEAR SUPERHEATER (BONUS) POWER STATION. Final Summary Design Report

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

None

1962-05-01

The design and construction of the Boiling Nuclear Superheater (BONUS) Power Station at Punta Higuera on the seacoast at the westernmost tip of Puerto Rico are described. The reactor has an output of 17.5 Mw(e). This report will serve as a source of information for personnel engaged in management, evaluation, and training. (N.W.R.)

Revised FINAL–REPORT NO. 2: INDEPENDENT CONFIRMATORY SURVEY SUMMARY AND RESULTS FOR THE ENRICO FERMI ATOMIC POWER PLANT, UNIT 1, NEWPORT, MICHIGAN (DOCKET NO. 50 16; RFTA 10-004) 2018-SR-02-1

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

Erika Bailey

2011-10-27

The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor power with the first core loading (Core A) was 200 MW. The primary system was filled with sodium in Decembermore » 1960 and criticality was achieved in August 1963. The reactor was tested at low power during the first couple years of operation. Power ascension testing above 1 MW commenced in December 1965 immediately following the receipt of a high-power operating license. In October 1966 during power ascension, zirconium plates at the bottom of the reactor vessel became loose and blocked sodium coolant flow to some fuel subassemblies. Two subassemblies started to melt and the reactor was manually shut down. No abnormal releases to the environment occurred. Forty-two months later after the cause had been determined, cleanup completed, and the fuel replaced, Fermi 1 was restarted. However, in November 1972, PRDC made the decision to decommission Fermi 1 as the core was approaching its burn-up limit. The fuel and blanket subassemblies were shipped off-site in 1973. Following that, the secondary sodium system was drained and sent off-site. The radioactive primary sodium was stored on-site in storage tanks and 55 gallon (gal) drums until it was shipped off-site in 1984. The initial decommissioning of Fermi 1 was completed in 1975. Effective January 23, 1976, DPR-9 was transferred to the Detroit Edison Company (DTE) as a 'possession only' license (DTE 2010a). This report details the confirmatory activities performed during the second Oak Ridge Institute for Science and Education (ORISE) site visit to Fermi 1 in November 2010. The survey was strategically planned during a Unit 2 (Fermi 2) outage to take advantage of decreased radiation levels that were observed and attributed to Fermi 2 from the operating unit during the first site visit. However, during the second visit there were elevated radiation levels observed and attributed to the partially dismantled Fermi 1 reactor vessel and a waste storage box located on the 3rd floor of the Fermi 1 Turbine Building. Confirmatory surveys (unshielded) performed directly in the line of sight of these areas were affected. The objective of the confirmatory survey was to verify that the final radiological conditions were accurately and adequately described in Final Status Survey (FSS) documentation, relative to the established release criteria. This objective was achieved by performing document reviews, as well as independent measurements and sampling. Specifically, documentation of the planning, implementation, and results of the FSS were evaluated; side-by-side FSS measurement and source comparisons were performed; site areas were evaluated relative to appropriate FSS classification; and areas were assessed for residual, undocumented contamination.« less

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

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

Davenport, Michael; Petti, D. A.; Palmer, Joe

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experimentsmore » are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.« less

Developpement d'une methode de Monte Carlo dependante du temps et application au reacteur de type CANDU-6

NASA Astrophysics Data System (ADS)

Mahjoub, Mehdi

La resolution de l'equation de Boltzmann demeure une etape importante dans la prediction du comportement d'un reacteur nucleaire. Malheureusement, la resolution de cette equation presente toujours un defi pour une geometrie complexe (reacteur) tout comme pour une geometrie simple (cellule). Ainsi, pour predire le comportement d'un reacteur nucleaire,un schema de calcul a deux etapes est necessaire. La premiere etape consiste a obtenir les parametres nucleaires d'une cellule du reacteur apres une etape d'homogeneisation et condensation. La deuxieme etape consiste en un calcul de diffusion pour tout le reacteur en utilisant les resultats de la premiere etape tout en simplifiant la geometrie du reacteur a un ensemble de cellules homogenes le tout entoure de reflecteur. Lors des transitoires (accident), ces deux etapes sont insuffisantes pour pouvoir predire le comportement du reacteur. Comme la resolution de l'equation de Boltzmann dans sa forme dependante du temps presente toujours un defi de taille pour tous types de geometries,un autre schema de calcul est necessaire. Afin de contourner cette difficulte, l'hypothese adiabatique est utilisee. Elle se concretise en un schema de calcul a quatre etapes. La premiere et deuxieme etapes demeurent les memes pour des conditions nominales du reacteur. La troisieme etape se resume a obtenir les nouvelles proprietes nucleaires de la cellule a la suite de la perturbation pour les utiliser, au niveau de la quatrieme etape, dans un nouveau calcul de reacteur et obtenir l'effet de la perturbation sur le reacteur. Ce projet vise a verifier cette hypothese. Ainsi, un nouveau schema de calcul a ete defini. La premiere etape de ce projet a ete de creer un nouveau logiciel capable de resoudre l'equation de Boltzmann dependante du temps par la methode stochastique Monte Carlo dans le but d'obtenir des sections efficaces qui evoluent dans le temps. Ce code a ete utilise pour simuler un accident LOCA dans un reacteur nucleaire de type CANDU-6. Les sections efficaces dependantes du temps ont ete par la suite utilisees dans un calcul de diffusion espace-temps pour un reacteur CANDU-6 subissant un accident de type LOCA affectant la moitie du coeur afin d'observer son comportement durant toutes les phases de la perturbation. Dans la phase de developpement, nous avons choisi de demarrer avec le code OpenMC, developpe au MIT,comme plateforme initiale de developpement. L'introduction et le traitement des neutrons retardes durant la simulation ont presente un grand defi a surmonter. Il est important de noter que le code developpe utilisant la methode Monte Carlo peut etre utilise a grande echelle pour la simulation de tous les types des reacteurs nucleaires si les supports informatiques sont disponibles.

Composting in small laboratory pilots: Performance and reproducibility

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

Lashermes, G.; Barriuso, E.; Le Villio-Poitrenaud, M.

2012-02-15

Highlights: Black-Right-Pointing-Pointer We design an innovative small-scale composting device including six 4-l reactors. Black-Right-Pointing-Pointer We investigate the performance and reproducibility of composting on a small scale. Black-Right-Pointing-Pointer Thermophilic conditions are established by self-heating in all replicates. Black-Right-Pointing-Pointer Biochemical transformations, organic matter losses and stabilisation are realistic. Black-Right-Pointing-Pointer The organic matter evolution exhibits good reproducibility for all six replicates. - Abstract: Small-scale reactors (<10 l) have been employed in composting research, but few attempts have assessed the performance of composting considering the transformations of organic matter. Moreover, composting at small scales is often performed by imposing a fixed temperature, thus creatingmore » artificial conditions, and the reproducibility of composting has rarely been reported. The objectives of this study are to design an innovative small-scale composting device safeguarding self-heating to drive the composting process and to assess the performance and reproducibility of composting in small-scale pilots. The experimental setup included six 4-l reactors used for composting a mixture of sewage sludge and green wastes. The performance of the process was assessed by monitoring the temperature, O{sub 2} consumption and CO{sub 2} emissions, and characterising the biochemical evolution of organic matter. A good reproducibility was found for the six replicates with coefficients of variation for all parameters generally lower than 19%. An intense self-heating ensured the existence of a spontaneous thermophilic phase in all reactors. The average loss of total organic matter (TOM) was 46% of the initial content. Compared to the initial mixture, the hot water soluble fraction decreased by 62%, the hemicellulose-like fraction by 68%, the cellulose-like fraction by 50% and the lignin-like fractions by 12% in the final compost. The TOM losses, compost stabilisation and evolution of the biochemical fractions were similar to observed in large reactors or on-site experiments, excluding the lignin degradation, which was less important than in full-scale systems. The reproducibility of the process and the quality of the final compost make it possible to propose the use of this experimental device for research requiring a mass reduction of the initial composted waste mixtures.« less

Investigation of Dispersed and Dispersed Annular (rivulet or Thin Film) Flow Phase Separation in Tees.

NASA Astrophysics Data System (ADS)

McCreery, Glenn Ernest

An experimental and analytical investigation of dispersed and dispersed-annular (rivulet or thin film) flow phase separation in tees has been successfully completed. The research was directed at, but is not specific to, determining flow conditions, following a loss of coolant accident, in the large rectangular passageways leading to vacuum buildings in the containment envelope of some CANDU nuclear reactors. The primary objectives of the research were to: (1) obtain experimental data to help formulate and test mechanistic analytical models of phase separation, and (2) develop the analytical models in computer programs which predict phase separation from upstream flow and pressure conditions and downstream and side branch pressure boundary conditions. To meet these objectives an air-water experimental apparatus was constructed, and consists of large air blowers attached to a long rectangular duct leading to a tee in the horizontal plane. A variety of phenomena was investigated including, for comparison with computer predictions, air streamlines and eddy boundary geometry, drop size spectra, macroscopic mass balances, liquid rivulet pathlines, and trajectories of drops of known size and velocity. Four separate computer programs were developed to analyze phase separation. Three of the programs are used sequentially to calculate dispersed mist phase separation in a tee. The fourth is used to calculate rivulet or thin film pathlines. Macroscopic mass balances are calculated from a summation of mass balances for drops with representative sizes (and masses) spaced across the drop size spectrum. The programs are tested against experimental data, and accurately predict gas flow fields, drop trajectories, rivulet pathlines and macroscopic mass balances. In addition to development of the computer programs, analysis was performed to specify the scaling of dispersed mist and rivulet or thin film flow, to investigate pressure losses in tees, and the inter-relationship of loss coefficients, contraction coefficients, and eddy geometry. The important transient effects of liquid storage in eddies were also analyzed.

Microstructural evolution of CANDU spacer material Inconel X-750 under in situ ion irradiation

NASA Astrophysics Data System (ADS)

Zhang, He Ken; Yao, Zhongwen; Judge, Colin; Griffiths, Malcolm

2013-11-01

Work on Inconel®Inconel® is a registered trademark of Special Metals Corporation that refers to a family of austenitic nickel-chromium-based superalloys.1 X-750 spacers removed from CANDU®CANDU® is a registered trademark of Atomic Energy of Canada Limited standing for ''CANada Deuterium Uranium''.2 reactors has shown that they become embrittled and there is development of many small cavities within the metal matrix and along grain boundaries. In order to emulate the neutron irradiation induced microstructural changes, heavy ion irradiations (1 MeV Kr2+ ions) were performed while observing the damage evolution using an intermediate voltage electron microscope (IVEM) operating at 200 kV. The irradiations were carried out at various temperatures 60-400 °C. The principal strengthening phase, γ‧, was disordered at low doses (˜0.06 dpa) during the irradiation. M23C6 carbides were found to be stable up to 5.4 dpa. Lattice defects consisted mostly of stacking fault tetrahedras (SFTs), 1/2<1 1 0> perfect loops and small 1/3<1 1 1> faulted Frank loops. The ratio of SFT number density to loop number density for each irradiation condition was found to be neither temperature nor dose dependent. Under the operation of the ion beam the SFT production was very rapid, with no evidence for further growth once formed, indicating that they probably formed as a result of cascade collapse in a single cascade. The number density of the defects was found to saturate at low dose (˜0.68 dpa). No cavities were observed regardless of the irradiation temperature between 60 °C and 400 °C for doses up to 5.4 dpa. In contrast, cavities have been observed after neutron irradiation in the same material at similar doses and temperatures indicating that helium, produce during neutron irradiation, may be essential for the nucleation and growth of cavities.

Human Health and the Biological Effects of Tritium in Drinking Water: Prudent Policy Through Science – Addressing the ODWAC New Recommendation

PubMed Central

Dingwall, S.; Mills, C.E.; Phan, N.; Taylor, K.; Boreham, D.R.

2011-01-01

Tritium is a radioactive form of hydrogen and is a by-product of energy production in Canadian Deuterium Uranium (CANDU) reactors. The release of this radioisotope into the environment is carefully managed at CANDU facilities in order to minimize radiation exposure to the public. However, under some circumstances, small accidental releases to the environment can occur. The radiation doses to humans and non-human biota from these releases are low and orders of magnitude less than doses received from naturally occurring radioisotopes or from manmade activities, such as medical imaging and air travel. There is however a renewed interest in the biological consequences of low dose tritium exposures and a new limit for tritium levels in Ontario drinking water has been proposed. The Ontario Drinking Water Advisory Council (ODWAC) issued a formal report in May 2009 in response to a request by the Minister of the Environment, concluding that the Ontario Drinking Water Quality Standard for tritium should be revised from the current 7,000 Bq/L level to a new, lower 20 Bq/L level. In response to this recommendation, an international scientific symposium was held at McMaster University to address the issues surrounding this change in direction and the validity of a new policy. Scientists, regulators, government officials, and industrial stakeholders were present to discuss the potential health risks associated with low level radiation exposure from tritium. The regulatory, economic, and social implications of the new proposed limit were also considered. The new recommendation assumed a linear-no-threshold model to calculate carcinogenic risk associated with tritium exposure, and considered tritium as a non-threshold chemical carcinogen. Both of these assumptions are highly controversial given that recent research suggests that low dose exposures have thresholds below which there are no observable detrimental effects. Furthermore, mutagenic and carcinogenic risk calculated from tritium exposure at 20 Bq/L would be orders of magnitude less than that from exposure to natural background sources of radiation. The new proposed standard would set the radiation dose limit for drinking water to 0.0003 mSv/year, which is equivalent to approximately three times the dose from naturally occurring tritium in drinking water. This new standard is incongruent with national and international standards for safe levels of radiation exposure, currently set at 1 mSv/year for the general public. Scientific research from leading authorities on the carcinogenic health effects of tritium exposure supports the notion that the current standard of 7,000 Bq/L (annual dose of 0.1 mSv) is a safe standard for human health. Policy-making for the purpose of regulating tritium levels in drinking water is a dynamic multi-stage process that is influenced by more than science alone. Ethics, economics, and public perception also play important roles in policy development; however, these factors sometimes undermine the scientific evidence that should form the basis of informed decision making. Consequently, implementing a new standard without a scientific basis may lead the public to perceive that risks from tritium have been historically underestimated. It was concluded that the new recommendation is not supported by any new scientific insight regarding negative consequences of low dose effects, and may be contrary to new data on the potential benefits of low dose effects. Given the lack of cost versus benefit analysis, this type of dramatic policy change could have detrimental effects to society from an ethical, economical, and public perception perspective. PMID:21431084

Human Health and the Biological Effects of Tritium in Drinking Water: Prudent Policy Through Science - Addressing the ODWAC New Recommendation.

PubMed

Dingwall, S; Mills, C E; Phan, N; Taylor, K; Boreham, D R

2011-02-22

Tritium is a radioactive form of hydrogen and is a by-product of energy production in Canadian Deuterium Uranium (CANDU) reactors. The release of this radioisotope into the environment is carefully managed at CANDU facilities in order to minimize radiation exposure to the public. However, under some circumstances, small accidental releases to the environment can occur. The radiation doses to humans and non-human biota from these releases are low and orders of magnitude less than doses received from naturally occurring radioisotopes or from manmade activities, such as medical imaging and air travel. There is however a renewed interest in the biological consequences of low dose tritium exposures and a new limit for tritium levels in Ontario drinking water has been proposed. The Ontario Drinking Water Advisory Council (ODWAC) issued a formal report in May 2009 in response to a request by the Minister of the Environment, concluding that the Ontario Drinking Water Quality Standard for tritium should be revised from the current 7,000 Bq/L level to a new, lower 20 Bq/L level. In response to this recommendation, an international scientific symposium was held at McMaster University to address the issues surrounding this change in direction and the validity of a new policy. Scientists, regulators, government officials, and industrial stakeholders were present to discuss the potential health risks associated with low level radiation exposure from tritium. The regulatory, economic, and social implications of the new proposed limit were also considered.The new recommendation assumed a linear-no-threshold model to calculate carcinogenic risk associated with tritium exposure, and considered tritium as a non-threshold chemical carcinogen. Both of these assumptions are highly controversial given that recent research suggests that low dose exposures have thresholds below which there are no observable detrimental effects. Furthermore, mutagenic and carcinogenic risk calculated from tritium exposure at 20 Bq/L would be orders of magnitude less than that from exposure to natural background sources of radiation. The new proposed standard would set the radiation dose limit for drinking water to 0.0003 mSv/year, which is equivalent to approximately three times the dose from naturally occurring tritium in drinking water. This new standard is incongruent with national and international standards for safe levels of radiation exposure, currently set at 1 mSv/year for the general public. Scientific research from leading authorities on the carcinogenic health effects of tritium exposure supports the notion that the current standard of 7,000 Bq/L (annual dose of 0.1 mSv) is a safe standard for human health.Policy-making for the purpose of regulating tritium levels in drinking water is a dynamic multi-stage process that is influenced by more than science alone. Ethics, economics, and public perception also play important roles in policy development; however, these factors sometimes undermine the scientific evidence that should form the basis of informed decision making. Consequently, implementing a new standard without a scientific basis may lead the public to perceive that risks from tritium have been historically underestimated. It was concluded that the new recommendation is not supported by any new scientific insight regarding negative consequences of low dose effects, and may be contrary to new data on the potential benefits of low dose effects. Given the lack of cost versus benefit analysis, this type of dramatic policy change could have detrimental effects to society from an ethical, economical, and public perception perspective.

Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

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

Moe, Wayne L.

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory importance of key DOE reactor research initiatives should be assessed early in the technology development process. Quality assurance requirements supportive of later licensing activities must also be attached to important research activities to ensure resulting data is usable in that context. Early regulatory analysis and licensing approach planning thus provides a significant benefit to the formulation of research plans and also enables the planning and development of a compatible AdvSMR licensing framework, should significant modification be required.« less

Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

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

Moe, Wayne Leland

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory importance of key DOE reactor research initiatives should be assessed early in the technology development process. Quality assurance requirements supportive of later licensing activities must also be attached to important research activities to ensure resulting data is usable in that context. Early regulatory analysis and licensing approach planning thus provides a significant benefit to the formulation of research plans and also enables the planning and development of a compatible AdvSMR licensing framework, should significant modification be required.« less

Rapid solar-thermal decarbonization of methane

NASA Astrophysics Data System (ADS)

Dahl, Jaimee Kristen

Due to the ever-increasing demand for energy and the concern over the environmental impact of continuing to produce energy using current methods, there is interest in developing a hydrogen economy. Hydrogen is a desirable energy source because it is abundant in nature and burns cleanly. One method for producing hydrogen is to utilize a renewable energy source to obtain high enough temperatures to decompose a fossil fuel into its elements. This thesis work is directed at developing a solar-thermal aerosol flow reactor to dissociate methane to carbon black and hydrogen. The technology is intended as a "bridge" between current hydrogen production methods, such as conventional steam-methane reformers, and future "zero emission" technology for producing hydrogen, such as dissociating water using a renewable heating source. A solar furnace is used to heat a reactor to temperatures in excess of 2000 K. The final reactor design studied consists of three concentric vertical tubes---an outer quartz protection tube, a middle solid graphite heating tube, and an inner porous graphite reaction tube. A "fluid-wall" is created on the inside wall of the porous reaction tube in order to prevent deposition of the carbon black co-product on the reactor tube wall. The amorphous carbon black produced aids in heating the gas stream by absorbing radiation from the reactor wall. Conversions of 90% are obtained at a reactor wall temperature of 2100 K and an average residence time of 0.01 s. Computer modeling is also performed to study the gas flow and temperature profiles in the reactor as well as the kinetics of the methane dissociation reaction. The simulations indicate that there is little flow of the fluid-wall gas through the porous wall in the hot zone region, but this can be remedied by increasing the inlet temperature of the fluid-wall gas and/or increasing the tube permeability only in the hot zone region of the wall. The following expression describes the kinetics of methane dissociation in a solar-thermal fluid-wall reactor: dXdt=5.8x108 exp-155,600RT 1-X 7.2s-1. The experimental and theoretical work reported in this thesis is the groundwork that will be utilized in scaling up the reactor to produce hydrogen in distributed or centralized facilities.

Removal of highly elevated nitrate from drinking water by pH-heterogenized heterotrophic denitrification facilitated with ferrous sulfide-based autotrophic denitrification.

PubMed

Huang, Bin; Chi, Guangyu; Chen, Xin; Shi, Yi

2011-11-01

The performance of acetic acid-supported pH-heterogenized heterotrophic denitrification (HD) facilitated with ferrous sulfide-based autotrophic denitrification (AD) was investigated in upflow activated carbon-packed column reactors for reliable removal of highly elevated nitrate (42 mg NO(3)-Nl(-1)) in drinking water. The use of acetic acid as substrate provided sufficient internal carbon dioxide to completely eliminate the need of external pH adjustment for HD, but simultaneously created vertically heterogenized pH varying from 4.8 to 7.8 in the HD reactor. After 5-week acclimation, the HD reactor developed a moderate nitrate removal capacity with about one third of nitrate removal occurring in the acidic zone (pH 4.8-6.2). To increase the treatment reliability, acetic acid-supported HD was operated under 10% carbon limitation to remove >85% of nitrate, and ferrous sulfide-based AD was supplementally operated to remove residual nitrate and formed nitrite without excess of soluble organic carbon, nitrite or sulfate in the final effluent. Copyright © 2011 Elsevier Ltd. All rights reserved.

Development of tritium permeation barriers on Al base in Europe

NASA Astrophysics Data System (ADS)

Benamati, G.; Chabrol, C.; Perujo, A.; Rigal, E.; Glasbrenner, H.

The development of the water cooled lithium lead (WCLL) DEMO fusion reactor requires the production of a material capable of acting as a tritium permeation barrier (TPB). In the DEMO blanket reactor permeation barriers on the structural material are required to reduce the tritium permeation from the Pb-17Li or the plasma into the cooling water to acceptable levels (<1 g/d). Because of experimental work previously performed, one of the most promising TPB candidates is A1 base coatings. Within the EU a large R&D programme is in progress to develop a TPB fabrication technique, compatible with the structural materials requirements and capable of producing coatings with acceptable performances. The research is focused on chemical vapour deposition (CVD), hot dipping, hot isostatic pressing (HIP) technology and spray (this one developed also for repair) deposition techniques. The final goal is to select a reference technique to be used in the blanket of the DEMO reactor and in the ITER test module fabrication. The activities performed in four European laboratories are summarised here.

A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation

DOE PAGES

Lai, Shigang; Shi, Li; Fok, Alex; ...

2017-01-01

Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC) is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancingmore » crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method.« less

A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation

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

Lai, Shigang; Shi, Li; Fok, Alex

Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC) is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancingmore » crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method.« less

Continuous immobilized yeast reactor system for complete beer fermentation using spent grains and corncobs as carrier materials.

PubMed

Brányik, Tomás; Silva, Daniel P; Vicente, António A; Lehnert, Radek; e Silva, João B Almeida; Dostálek, Pavel; Teixeira, José A

2006-12-01

Despite extensive research carried out in the last few decades, continuous beer fermentation has not yet managed to outperform the traditional batch technology. An industrial breakthrough in favour of continuous brewing using immobilized yeast could be expected only on achievement of the following process characteristics: simple design, low investment costs, flexible operation, effective process control and good product quality. The application of cheap carrier materials of by-product origin could significantly lower the investment costs of continuous fermentation systems. This work deals with a complete continuous beer fermentation system consisting of a main fermentation reactor (gas-lift) and a maturation reactor (packed-bed) containing yeast immobilized on spent grains and corncobs, respectively. The suitability of cheap carrier materials for long-term continuous brewing was proved. It was found that by fine tuning of process parameters (residence time, aeration) it was possible to adjust the flavour profile of the final product. Consumers considered the continuously fermented beer to be of a regular quality. Analytical and sensorial profiles of both continuously and batch fermented beers were compared.

Dynamic control of nutrient-removal from industrial wastewater in a sequencing batch reactor, using common and low-cost online sensors.

PubMed

Dries, Jan

2016-01-01

On-line control of the biological treatment process is an innovative tool to cope with variable concentrations of chemical oxygen demand and nutrients in industrial wastewater. In the present study we implemented a simple dynamic control strategy for nutrient-removal in a sequencing batch reactor (SBR) treating variable tank truck cleaning wastewater. The control system was based on derived signals from two low-cost and robust sensors that are very common in activated sludge plants, i.e. oxidation reduction potential (ORP) and dissolved oxygen. The amount of wastewater fed during anoxic filling phases, and the number of filling phases in the SBR cycle, were determined by the appearance of the 'nitrate knee' in the profile of the ORP. The phase length of the subsequent aerobic phases was controlled by the oxygen uptake rate measured online in the reactor. As a result, the sludge loading rate (F/M ratio), the volume exchange rate and the SBR cycle length adapted dynamically to the activity of the activated sludge and the actual characteristics of the wastewater, without affecting the final effluent quality.

In-situ biogas upgrading in thermophilic granular UASB reactor: key factors affecting the hydrogen mass transfer rate.

PubMed

Bassani, Ilaria; Kougias, Panagiotis G; Angelidaki, Irini

2016-12-01

Biological biogas upgrading coupling CO 2 with external H 2 to form biomethane opens new avenues for sustainable biofuel production. For developing this technology, efficient H 2 to liquid transfer is fundamental. This study proposes an innovative setup for in-situ biogas upgrading converting the CO 2 in the biogas into CH 4 , via hydrogenotrophic methanogenesis. The setup consisted of a granular reactor connected to a separate chamber, where H 2 was injected. Different packing materials (rashig rings and alumina ceramic sponge) were tested to increase gas-liquid mass transfer. This aspect was optimized by liquid and gas recirculation and chamber configuration. It was shown that by distributing H 2 through a metallic diffuser followed by ceramic sponge in a separate chamber, having a volume of 25% of the reactor, and by applying a mild gas recirculation, CO 2 content in the biogas dropped from 42 to 10% and the final biogas was upgraded from 58 to 82% CH 4 content. Copyright © 2016 Elsevier Ltd. All rights reserved.

Copper recovery and gold enrichment from waste printed circuit boards by mediated electrochemical oxidation.

PubMed

Fogarasi, Szabolcs; Imre-Lucaci, Florica; Imre-Lucaci, Arpád; Ilea, Petru

2014-05-30

The present study aims to develop an eco-friendly chemical-electrochemical process for the simultaneous recovery of copper and separation of a gold rich residue from waste printed circuit boards (WPCBs). The process was carried out by employing two different types of reactors coupled in series: a leaching reactor with a perforated rotating drum, for the dissolution of base metals and a divided electrochemical reactor for the regeneration of the leaching solution with the parallel electrowinning of copper. The process performances were evaluated on the basis of the dissolution efficiency, current efficiency and specific energy consumptions. Finally a process scale up was realized taking into consideration the optimal values of the operating parameters. The laboratory scale leaching plant allowed the recovery of a high purity copper deposit (99.04wt.%) at a current efficiency of 63.84% and specific energy consumption of 1.75kWh/kg cooper. The gold concentration in the remained solid residue was 25 times higher than the gold concentration in the initial WPCB samples. Copyright © 2014 Elsevier B.V. All rights reserved.

Qualification of heavy water based irradiation device in the JSI TRIGA reactor for irradiations of FT-TIMS samples for nuclear safeguards

NASA Astrophysics Data System (ADS)

Radulović, Vladimir; Kolšek, Aljaž; Fauré, Anne-Laure; Pottin, Anne-Claire; Pointurier, Fabien; Snoj, Luka

2018-03-01

The Fission Track Thermal Ionization Mass Spectrometry (FT-TIMS) method is considered as the reference method for particle analysis in the field of nuclear Safeguards for measurements of isotopic compositions (fissile material enrichment levels) in micrometer-sized uranium particles collected in nuclear facilities. An integral phase in the method is the irradiation of samples in a very well thermalized neutron spectrum. A bilateral collaboration project was carried out between the Jožef Stefan Institute (JSI, Slovenia) and the Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA, France) to determine whether the JSI TRIGA reactor could be used for irradiations of samples for the FT-TIMS method. This paper describes Monte Carlo simulations, experimental activation measurements and test irradiations performed in the JSI TRIGA reactor, firstly to determine the feasibility, and secondly to design and qualify a purpose-built heavy water based irradiation device for FT-TIMS samples. The final device design has been shown experimentally to meet all the required performance specifications.

Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells

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

E.T. Robinson; John Sirman; Prasad Apte

2005-05-01

This final report summarizes work accomplished in the Program from January 1, 2001 through December 31, 2004. Most of the key technical objectives for this program were achieved. A breakthrough material system has lead to the development of an OTM (oxygen transport membrane) compact planar reactor design capable of producing either syngas or hydrogen. The planar reactor shows significant advantages in thermal efficiency and a step change reduction in costs compared to either autothermal reforming or steam methane reforming with CO{sub 2} recovery. Syngas derived ultra-clean transportation fuels were tested in the Nuvera fuel cell modular pressurized reactor and inmore » International Truck and Engine single cylinder test engines. The studies compared emission and engine performance of conventional base fuels to various formulations of ultra-clean gasoline or diesel fuels. A proprietary BP oxygenate showed significant advantage in both applications for reducing emissions with minimal impact on performance. In addition, a study to evaluate new fuel formulations for an HCCI engine was completed.« less

SLSF in-reactor local fault safety experiment P4. Final report

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

Thompson, D. H.; Holland, J. W.; Braid, T. H.

The Sodium Loop Safety Facility (SLSF), a major facility in the US fast-reactor safety program, has been used to simulate a variety of sodium-cooled fast reactor accidents. SLSF experiment P4 was conducted to investigate the behavior of a "worse-than-case" local fault configuration. Objectives of this experiment were to eject molten fuel into a 37-pin bundle of full-length Fast-Test-Reactor-type fuel pins form heat-generating fuel canisters, to characterize the severity of any molten fuel-coolant interaction, and to demonstrate that any resulting blockage could either be tolerated during continued power operation or detected by global monitors to prevent fuel failure propagation. The designmore » goal for molten fuel release was 10 to 30 g. Explusion of molten fuel from fuel canisters caused failure of adjacent pins and a partial flow channel blockage in the fuel bundle during full-power operation. Molten fuel and fuel debris also lodged against the inner surface of the test subassembly hex-can wall. The total fuel disruption of 310 g evaluated from posttest examination data was in excellent agreement with results from the SLSF delayed neutron detection system, but exceeded the target molten fuel release by an order of magnitude. This report contains a summary description of the SLSF in-reactor loop and support systems and the experiment operations. results of the detailed macro- and microexamination of disrupted fuel and metal and results from the analysis of the on-line experimental data are described, as are the interpretations and conclusions drawn from the posttest evaluations. 60 refs., 74 figs.« less

Manufacturing demonstration of microbially mediated zinc sulfide nanoparticles in pilot-plant scale reactors

DOE PAGES

Moon, Ji-Won; Phelps, Tommy J.; Fitzgerald Jr, Curtis L.; ...

2016-04-27

The thermophilic anaerobic metal-reducing bacterium Thermoanaerobacter sp. X513 efficiently produces zinc sulfide (ZnS) nanoparticles (NPs) in laboratory-scale ( ≤24-L) reactors. To determine whether this process can be up-scaled and adapted for pilot-plant production while maintaining NP yield and quality, a series of meso-scale experiments were performed using 100-l and 900-l reactors. Pasteurization and N 2-sparging replaced autoclaving and boiling for deoxygenating media in the transition from small-scale to pilot-plant reactors. Consecutive 100-L batches using new or recycled media produced ZnS NPs with highly reproducible ~2 nm average crystallite size (ACS) and yields of ~0.5g L -1, similar to small-scale batches.more » The 900-L pilot plant reactor produced ~ 320 g ZnS without process optimization or replacement of used medium; this quantity would be sufficient to form a ZnS thin film with ~120 nm thickness over 0.5 m width 13 km length. At all scales, the bacteria produced significant amounts of acetic, lactic and formic acids, which could be neutralized by the controlled addition of sodium hydroxide without the use of an organic pH buffer, eliminating 98% of the buffer chemical costs. In conclusion, the final NP products were characterized using XRD, ICP-OES, FTIR, DLS, and C/N analyses, which confirmed the growth medium without organic buffer enhanced the ZnS NP properties by reducing carbon and nitrogen surface coatings and supporting better dispersivity with similar ACS.« less

Anaerobic reactor/high rate pond combined technology for sewage treatment in the Mediterranean area.

PubMed

El Hafiane, F; El Hamouri, B

2005-01-01

Two high-rate, anaerobic/aerobic units were used to treat the sewage of the Institut Agronomique st Vétérinaire Hassan II (Morocco) campus in a 1,100 m2-plant designed for 1,500 e.p. and receiving 63 m3 per day. The anaerobic pre-treatment consisted of a two-step up-flow anaerobic reactor (TSUAR) comprising two reactors and one external settler all in series. The aerobic line, or post-treatment, consisted of a high-rate algal pond (HRAP) and one maturation pond in series. The system totalized a hydraulic retention time (HRT) of 9 days. A gravel filter (GF) was constructed behind the TSUAR to trap low-density particles. The TSUAR removed 80% of COD and 90% of SS within 48 h. Solids retention time in the reactors averaged 32 d with a specific sludge production of 0.28 g SS g(-1) COD removed. Almost 93% of the sludge evacuated from the settler was stabilized. Specific biogas production from both reactors was 0.25m3 kg(-1) COD removed. Used in this configuration, the HRAP lost its BOD removal activity and increased its nutrients and pathogens removal capabilities (tertiary treatment). Results showed that 85% of total nitrogen and 48% of total phosphorus were removed by the HRAP. Land area requirement of this combination was less than 1 m2 per capita and filtered final effluent was of excellent quality (COD, 82 mg/l; TKN, 8.3 mg/l; total P, 2.7 mg/l, faecal coliforms, 2.4 10(3)/100 ml and zero helminths eggs).

Technicians Manufacture a Nozzle for the Kiwi B-1-B Engine

NASA Image and Video Library

1964-05-21

Technicians manufacture a nozzle for the Kiwi B-1-B nuclear rocket engine in the Fabrication Shop’s vacuum oven at the National Aeronautics and Space Administration (NASA) Lewis Research Center. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test basic nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The final phase of the program, called Reactor-In-Flight-Test, would be an actual launch test. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The turbopump, which pumped the fuels from the storage tanks to the engine, was the primary tool for restarting the engine. The NERVA had to be able to restart in space on its own using a safe preprogrammed startup system. Lewis researchers endeavored to design and test this system. This non-nuclear Kiwi engine, seen here, was being prepared for tests at Lewis’ High Energy Rocket Engine Research Facility (B-1) located at Plum Brook Station. The tests were designed to start an unfueled Kiwi B-1-B reactor and its Aerojet Mark IX turbopump without any external power.

Manufacturing demonstration of microbially mediated zinc sulfide nanoparticles in pilot-plant scale reactors.

PubMed

Moon, Ji-Won; Phelps, Tommy J; Fitzgerald, Curtis L; Lind, Randall F; Elkins, James G; Jang, Gyoung Gug; Joshi, Pooran C; Kidder, Michelle; Armstrong, Beth L; Watkins, Thomas R; Ivanov, Ilia N; Graham, David E

2016-09-01

The thermophilic anaerobic metal-reducing bacterium Thermoanaerobacter sp. X513 efficiently produces zinc sulfide (ZnS) nanoparticles (NPs) in laboratory-scale (≤ 24-L) reactors. To determine whether this process can be up-scaled and adapted for pilot-plant production while maintaining NP yield and quality, a series of pilot-plant scale experiments were performed using 100-L and 900-L reactors. Pasteurization and N2-sparging replaced autoclaving and boiling for deoxygenating media in the transition from small-scale to pilot plant reactors. Consecutive 100-L batches using new or recycled media produced ZnS NPs with highly reproducible ~2-nm average crystallite size (ACS) and yields of ~0.5 g L(-1), similar to the small-scale batches. The 900-L pilot plant reactor produced ~320 g ZnS without process optimization or replacement of used medium; this quantity would be sufficient to form a ZnS thin film with ~120 nm thickness over 0.5 m width × 13 km length. At all scales, the bacteria produced significant amounts of acetic, lactic, and formic acids, which could be neutralized by the controlled addition of sodium hydroxide without the use of an organic pH buffer, eliminating 98 % of the buffer chemical costs. The final NP products were characterized using XRD, ICP-OES, TEM, FTIR, PL, DLS, HPLC, and C/N analyses, which confirmed that the growth medium without organic buffer enhanced the ZnS NP properties by reducing carbon and nitrogen surface coatings and supporting better dispersivity with similar ACS.

The Bosch Process-Performance of a Developmental Reactor and Experimental Evaluation of Alternative Catalysts

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew

2010-01-01

Bosch-based reactors have been in development at NASA since the 1960's. Traditional operation involves the reduction of carbon dioxide with hydrogen over a steel wool catalyst to produce water and solid carbon. While the system is capable of completely closing the loop on oxygen and hydrogen for Atmosphere Revitalization, steel wool requires a reaction temperature of 650C or higher for optimum performance. The single pass efficiency of the reaction over steel wool has been shown to be less than 10% resulting in a high recycle stream. Finally, the formation of solid carbon on steel wool ultimately fouls the catalyst necessitating catalyst resupply. These factors result in high mass, volume and power demands for a Bosch system. Interplanetary transportation and surface exploration missions of the moon, Mars, and near-earth objects will require higher levels of loop closure than current technology cannot provide. A Bosch system can provide the level of loop closure necessary for these long-term missions if mass, volume, and power can be kept low. The keys to improving the Bosch system lie in reactor and catalyst development. In 2009, the National Aeronautics and Space Administration refurbished a circa 1980's developmental Bosch reactor and built a sub-scale Bosch Catalyst Test Stand for the purpose of reactor and catalyst development. This paper describes the baseline performance of two commercially available steel wool catalysts as compared to performance reported in the 1960's and 80's. Additionally, the results of sub-scale testing of alternative Bosch catalysts, including nickel- and cobalt-based catalysts, are discussed.

76 FR 6638 - Virginia Electric and Power Company D/B/A/ Dominion Virginia Power and Old Dominion Electric...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-02-07

... Company D/B/A/ Dominion Virginia Power and Old Dominion Electric Cooperative, North Anna Power Station... combined license (COL) application to build and operate a new reactor at its North Anna Power Station (NAPS... Combined License (COL) for North Anna Power Station, Unit 3.'' A notice of availability of the final...

9 CFR 147.6 - Procedure for determining the status of flocks reacting to tests for Mycoplasma gallisepticum...

Code of Federal Regulations, 2010 CFR

2010-01-01

... Agency may examine reactors by the in vivo bio-assay, PCR-based procedures, and/or culture procedures before final determination of the flock status is made. (13) If the in vivo bio-assay, PCR-based... classification for which it was tested. (14) If the in vivo bio-assay, PCR-based procedures, or culture...

Scientific and Technical Manpower Requirements of Selected Segments of the Atomic Energy Field. Final Report.

ERIC Educational Resources Information Center

Voight, Keith L.

The primary purpose of the study was to develop a supply/demand ratio for nuclear degree scientists and engineers from July 1969 through 1973. The need by private industry and electric utilities for scientists and engineers with degrees in disciplines other than nuclear science or engineering, as well as for technicians, nuclear reactor operators,…

Browns Ferry Unit-3 cavity neutron spectral analysis. Final report

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

Martin, G.C.

1981-08-01

This report describes neutron dosimetry measurements performed in the Browns Ferry Unit-3 reactor cavity using multiple dosimeter and spectrum unfolding techniques to assess radiation-induced degradation of nuclear plant pressure vessels. Test results and conclusions indicating the feasibility of determining neutron flux spectra and the densities in the pressure vessel cavity region via dosimetric measurements are presented.

Progress on RERTR activities in Argentina

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

Balart, S.; Calzetta, O.; Cristini, P.

2008-07-15

Since last RERTR meeting, several tasks involving RERTR activities continued deploying in Argentina: through an agreement between CNEA and US-DoE final steps in the RA-6 reactor core conversion from HEU to LEU are taking place; by means of a return campaign of 42 US origin SNF in the frame of the US-SNF FRR program; an effective minimization of HEU inventory is close to be accomplished; development of a LEU dispersed U-Mo fuel prototype, to be irradiated in a high flux reactor in the frame of the ARG/4/092 IAEA's Technical Cooperation project is progressing; very high density monolithic U-Mo miniplates andmore » plates using MEU and LEU fuel with Zry-4 cladding were developed to be irradiated as a part of the RERTR program irradiation experiment; atomistic modeling prediction (BFS techniques and first principles) enabled to find some trends on the interaction phases; diffusion couples tests under X-ray synchrotron analysis allowed the characterization of several phases involving U-Mo(-Zr) / Al(-Si); finally CNEA continued spreading high quality LEU technology for fission RI production by means of agreements with different producers interested on HEU-LEU conversion. (author)« less

Transmit-receive eddy current probes for defect detection and sizing in steam generator tubes

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

Obrutsky, L.S.; Cecco, V.S.; Sullivan, S.P.

1997-02-01

Inspection of steam generator tubes in aging Nuclear Generating Stations is increasingly important. Defect detection and sizing, especially in defect prone areas such as the tubesheet, support plates and U-bend regions, are required to assess the fitness-for-service of the steam generators. Information about defect morphology is required to address operational integrity issues, i.e., risk of tube rupture, number of tubes at risk, consequential leakage. A major challenge continues to be the detection and sizing of circumferential cracks. Utilities around the world have experienced this type of tube failure. Conventional in-service inspection, performed with eddy current bobbin probes, is ineffectual inmore » detecting circumferential cracks in tubing. It has been demonstrated in CANDU steam generators, with deformation, magnetite and copper deposits that multi-channel probes with transmit-receive eddy current coils are superior to those using surface impedance coils. Transmit-receive probes have strong directional properties, permitting probe optimization according to crack orientation. They are less sensitive to lift-off noise and magnetite deposits and possess good discrimination to internal defects. A single pass C3 array transmit-receive probe developed by AECL can detect and size circumferential stress corrosion cracks as shallow as 40% through-wall. Since its first trial in 1992, it has been used routinely for steam generator in-service inspection of four CANDU plants, preventing unscheduled shutdowns due to leaking steam generator tubes. More recently, a need has surfaced for simultaneous detection of both circumferential and axial cracks. The C5 probe was designed to address this concern. It combines transmit-receive array probe technology for equal sensitivity to axial and circumferential cracks with a bobbin probe for historical reference. This paper will discuss the operating principles of transmit-receive probes, along with inspection results.« less

Development of Molten Corium Using An Exothermic Chemical Reaction for the Molten- Fuel Moderator-Interaction Studies at Chalk River Laboratories

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

Nitheanandan, T.; Sanderson, D.B.; Kyle, G.

2004-07-01

Atomic Energy of Canada Limited (AECL) has partnered with Argonne National Laboratory to develop a corium thermite prototypical of Candu material and test the concept of ejecting {approx}25 kg of the molten material from a pressure tube with a driving pressure of 10 MPa. This development program has been completed and the technology transferred to AECL. Preparation for the molten-fuel moderator-interaction tests at AECL's Chalk River Laboratories is well underway. A mixture of 0.582 U/0.077 U{sub 3}O{sub 8}/0.151 Zr/0.19 CrO{sub 3} (wt%) as reactant chemicals has been demonstrated to produce a corium consisting of 0.73 UO{sub 2}/0.11 Zr/0.06 ZrO{sub 2}/0.10more » Cr (wt%) at {approx}2400 deg. C. This is comparable to the target Candu specific corium of 0.9 UO{sub 2}/0.1 Zr (wt%), with limited oxidation. The peak melt temperature was confirmed from small-scale thermitic reaction tests. Several small-scale tests were completed to qualify the thermite to ensure operational safety and a quantifiable experimental outcome. The proposed molten-fuel moderator-interaction experiments at Chalk River Laboratories will consist of heating the thermite mixture inside a 1.14-m long insulated pressure tube. Once the molten material has reached the desired temperature of {approx}2400 deg. C, the pressure inside the tube will be raised to about 10 MPa, and the pressure tube will fail at a pre-machined flaw, ejecting the molten material into the surrounding tank of water. The test apparatus, instrumentation, data acquisition and control systems have been assembled, and a series of successful commissioning tests have been completed. (authors)« less

RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

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

Kelly Cummins; Igor Bolshinsky; Ken Allen

2009-07-01

In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required tomore » complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.« less

Aging Management Guideline for commercial nuclear power plants: Motor control centers; Final report

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

Toman, G.; Gazdzinski, R.; O`Hearn, E.

1994-02-01

This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specificmore » aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.« less

Development of a proton-exchange membrane electrochemical reclaimed water post-treatment system

NASA Technical Reports Server (NTRS)

Kaba, Lamine; Verostko, Charles E.; Hitchens, G. D.; Murphy, Oliver J.

1991-01-01

A single-cell electrochemical reactor that utilizes a proton exchange membrane (PEM) as a solid electrolyte is being investigated for posttreatment of reclaimed waste waters with low or negligible electrolyte content. Posttreatment is a final 'polishing' of reclaimed waste waters prior to reuse, and involves removing organic impurities at levels as high as 100 ppm to below 500 ppb total organic carbon (TOC) content to provide disinfection. The system does not utilize or produce either expendable hardware components or chemicals and has no moving parts. Test data and kinetic analysis are presented. The feasibility and application for water reclamation processes in controlled ecological environments (e.g., lunar/Mars habitats) are also presented. Test results show that the electrochemical single cell reactor provides effective posttreatment.

Grid-to-rod flow-induced impact study for PWR fuel in reactor

DOE PAGES

Jiang, Hao; Qu, Jun; Lu, Roger Y.; ...

2016-06-10

The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effectivemore » way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE PAGES

Durham, J. M.; Poulson, D.; Bacon, J.; ...

2018-04-10

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Hydrothermal corrosion of silicon carbide joints without radiation

DOE PAGES

Koyanagi, Takaaki; Katoh, Yutai; Terrani, Kurt A.; ...

2016-09-28

In this paper, hydrothermal corrosion of four types of the silicon carbide (SiC) to SiC plate joints were investigated under pressurized water reactor and boiling water reactor relevant chemical conditions without irradiation. The joints were formed by metal diffusion bonding using molybdenum or titanium interlayer, reaction sintering using Ti—Si—C system, and SiC nanopowder sintering. Most of the joints withstood the corrosion tests for five weeks. The recession of the SiC substrates was limited. Based on the recession of the bonding layers, it was concluded that all the joints except for the molybdenum diffusion bond are promising under the reducing environmentsmore » without radiation. Finally, the SiC nanopowder sintered joint was the most corrosion tolerant under the oxidizing environment among the four joints.« less

Method for producing components with internal architectures, such as micro-channel reactors, via diffusion bonding sheets

DOEpatents

Alman, David E [Corvallis, OR; Wilson, Rick D [Corvallis, OR; Davis, Daniel L [Albany, OR

2011-03-08

This invention relates to a method for producing components with internal architectures, and more particularly, this invention relates to a method for producing structures with microchannels via the use of diffusion bonding of stacked laminates. Specifically, the method involves weakly bonding a stack of laminates forming internal voids and channels with a first generally low uniaxial pressure and first temperature such that bonding at least between the asperites of opposing laminates occurs and pores are isolated in interfacial contact areas, followed by a second generally higher isostatic pressure and second temperature for final bonding. The method thereby allows fabrication of micro-channel devices such as heat exchangers, recuperators, heat-pumps, chemical separators, chemical reactors, fuel processing units, and combustors without limitation on the fin aspect ratio.

Measuring Theta_13 at Daya Bay

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

Lau, Kwong

2014-03-14

We measured the neutrino mixing angle, theta13, presumably related to the preponderance of matter over antimatter in our universe with high precision. We determined theta13 by measuring the disappearance of neutrinos from a group of six nuclear reactors. The target, located inside a mountain at about 2 km from the reactors, is 80 tons of liquid scintillator doped with trace amount of Gadolinium to increase its neutron detection efficiency. The neutrino flux is measured by the inverse beta-decay reaction where the final-state particles are detected by the liquid scintillator. The measured value of theta13, based on data collected over 3more » years, is large, around 8 degrees, rendering the measurement of the parameter related to matter-antimatter asymmetry in future long baseline neutrino experiments easier.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

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

Durham, J. M.; Poulson, D.; Bacon, J.

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Catalytic biomass conversion methods, catalysts, and methods of making the same

DOEpatents

Delgass, William Nicholas; Agrawal, Rakesh; Ribeiro, Fabio Henrique; Saha, Basudeb; Yohe, Sara Lynn; Abu-Omar, Mahdi M; Parsell, Trenton; Dietrich, Paul James; Klein, Ian Michael

2017-10-10

Described herein are processes for one-step delignification and hydrodeoxygenation of lignin fraction a biomass feedstock. The lignin feedstock is derived from by-products of paper production and biorefineries. Additionally described is a process for converting biomass-derived oxygenates to lower oxygen-content compounds and/or hydrocarbons in the liquid or vapor phase in a reactor system containing hydrogen and a catalyst comprised of a hydrogenation function and/or an oxophilic function and/or an acid function. Finally, also described herein is a process for converting biomass-derived oxygenates to lower oxygen-content compounds and/or hydrocarbons in the liquid or vapor phase in a reactor system containing hydrogen and a catalyst comprised of a hydrogenation function and/or an oxophilic function and/or an acid function.

Screening study for evaluation of the potential for system 80+ to consume excess plutonium - Volume 1. Final report

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

Not Available

1994-04-30

As part of the U.S. effort to evaluate technologies offering solutions for the safe disposal or utilization of surplus nuclear materials, the fiscal year 1993 Energy and Water Appropriations legislation provided the Department of Energy (DOE) the necessary funds to conduct multi-phased studies to determine the technical feasibility of using reactor technologies for the triple mission of burning weapons grade plutonium, producing tritium for the existing smaller weapons stockpile, and generating commercial electricity. DOE limited the studies to five advanced reactor designs. Among the technologies selected is the ABB-Combustion Engineering (ABB-CE) System 80+. The DOE study, currently in Phase ID,more » is proceeding with a more detailed evaluation of the design`s capability for plutonium disposition.« less

OPERATIONAL CHARACTERISTICS OF THE ARMOUR FISSION GAS GAMMA FACILITY

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

Terrell, C.W.; McElroy, W.N.

1958-10-31

As the reactor power level is changed frequently, the radiation levels in the gamma facility fluctuate. Data are presented to show the power dependency of the gamma dose rate and the manner of growth and decay. Additional data show the dependercy of the equilibrium gamma activity on the foel temperature and total system pressure. The final phase of the work is directed toward determining an average gamma energy by attenuation measurements with various thicknesses of several materials. The neutrou flux associated with the gas phase activity is determined by foil measurement. From the measurements of dose rate and average gammamore » energy, calculations to determine the number of curies of gas phase decay gamma activity per watt of reactor power are presented. (auth)« less

Fukushima Accident: Sequence of Events and Lessons Learned

NASA Astrophysics Data System (ADS)

Morse, Edward C.

2011-10-01

The Fukushima Dai-Ichi nuclear power station suffered a devastating Richter 9.0 earthquake followed by a 14.0 m tsunami on 11 March 2011. The subsequent loss of power for emergency core cooling systems resulted in damage to the fuel in the cores of three reactors. The relief of pressure from the containment in these three reactors led to sufficient hydrogen gas release to cause explosions in the buildings housing the reactors. There was probably subsequent damage to a spent fuel pool of a fourth reactor caused by debris from one of these explosions. Resultant releases of fission product isotopes in air were significant and have been estimated to be in the 3 . 7 --> 6 . 3 ×1017 Bq range (~10 MCi) for 131I and 137Cs combined, or approximately one tenth that of the Chernobyl accident. A synopsis of the sequence of events leading up to this large release of radioactivity will be presented, along with likely scenarios for stabilization and site cleanup in the future. Some aspects of the isotope monitoring programs, both locally and at large, will also be discussed. An assessment of radiological health risk for the plant workers as well as the general public will also be presented. Finally, the impact of this accident on design and deployment of nuclear generating stations in the future will be discussed.

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

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

Daniel, G.; Rudisill, T.; Almond, P.

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Sitemore » (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.« less

Impact of paper and cardboard suppression on OFMSW anaerobic digestion.

PubMed

Fonoll, X; Astals, S; Dosta, J; Mata-Alvarez, J

2016-10-01

Mechanical-biological treatment plants treat municipal solid waste to recover recyclable materials, nutrients and energy. Waste paper and cardboard (WP), the second main compound in municipal solid waste (∼30% in weight basis), is typically used for biogas generation. However, its recovery is gaining attention as it can be used to produce add-value products like bioethanol and residual derived fuel. Nevertheless, WP suppression or replacement will impact anaerobic digestion in terms of biogas production, process stability and digestate management. Two lab-scale reactors were used to assess the impact of WP in anaerobic digestion performance. A control reactor was only fed with biowaste (BioW), while a second reactor was fed with two different mixtures of BioW and WP, i.e. 85/15% and 70/30% (weight basis). Results indicate that either replacing half of the WP by BioW or removing half of the WP has little impact on the methane production. When removing half of the WP, methane production could be sustained by a larger waste biodegradability. The replacement of all WP by BioW increased the reactor methane production (∼37%), while removing all WP would have reduced the methane production about 15%. Finally, replacing WP loading rate by BioW led to a system less tolerant to instability periods and with poorer digestate quality. Copyright © 2016 Elsevier Ltd. All rights reserved.

Summary of ORSphere Critical and Reactor Physics Measurements

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

Marshall, Margaret A.; Bess, John D.

In the early 1970s Dr. John T. Mihalczo (team leader), J. J. Lynn, and J. R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVAmore » I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.« less

Pressurized diesel fuel processing using hydrogen peroxide for the fuel cell power unit in low-oxygen environments

NASA Astrophysics Data System (ADS)

Lee, Kwangho; Han, Gwangwoo; Cho, Sungbaek; Bae, Joongmyeon

2018-03-01

A novel concept for diesel fuel processing utilizing H2O2 is suggested to obtain the high-purity H2 required for air-independent propulsion using polymer electrolyte membrane fuel cells for use in submarines and unmanned underwater vehicles. The core components include 1) a diesel-H2O2 autothermal reforming (ATR) reactor to produce H2-rich gas, 2) a water-gas shift (WGS) reactor to convert CO to H2, and 3) a H2 separation membrane to separate only high-purity H2. Diesel and H2O2 can easily be pressurized as they are liquids. The application of the H2 separation membrane without a compressor in the middle of the process is thus advantageous. In this paper, the characteristics of pressurized ATR and WGS reactions are investigated according to the operating conditions. In both reactors, the methanation reaction is enhanced as the pressure increases. Then, permeation experiments with a H2 separation membrane are performed while varying the temperature, pressure difference, and inlet gas composition. In particular, approximately 90% of the H2 is recovered when the steam-separated rear gas of the WGS reactor is used in the H2 separation membrane. Finally, based on the experimental results, design points are suggested for maximizing the efficiency of the diesel-H2O2 fuel processor.

Artemisinin production by plant hairy root cultures in gas- and liquid-phase bioreactors.

PubMed

Patra, Nivedita; Srivastava, Ashok K

2016-01-01

Alternative biotechnological protocol for large-scale artemisinin production was established. It featured enhanced growth and artemisinin production by cultivation of hairy roots in nutrient mist bioreactor (NMB) coupled with novel cultivation strategies. Artemisinin is used for the treatment of cerebral malaria. Presently, its main source is from seasonal plant Artemisia annua. This study featured investigation of growth and artemisinin production by A. annua hairy roots (induced by Agrobacterium rhizogenes-mediated genetic transformation of explants) in three bioreactor configurations-bubble column reactor, NMB and modified NMB particularly to establish their suitability for commercial production. It was observed that cultivation of hairy roots in a non-stirred bubble column reactor exhibited a biomass accumulation of 5.68 g/l only while batch cultivation in a custom-made NMB exhibited a higher biomass concentration of 8.52 g/l but relatively lower artemisinin accumulation of 0.22 mg/g was observed in this reactor. A mixture of submerged liquid-phase growth (for 5 days) followed by gas-phase cultivation in nutrient mist reactor operation strategy (for next 15 days) was adopted for hairy root cultivation in this investigation. Reasonably, high (23.02 g/l) final dry weight along with the artemisinin accumulation (1.12 mg/g, equivalent to 25.78 mg/l artemisinin) was obtained in this bioreactor, which is the highest reported artemisinin yield in the gas-phase NMB cultivation.

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

NASA Astrophysics Data System (ADS)

Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

2018-02-01

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

DOE PAGES

Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; ...

2016-08-15

Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representativemore » of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.« less

Comprehensive characterization of measurement data gathered by the pressure tube to calandria tube gap probe

NASA Astrophysics Data System (ADS)

Shokralla, Shaddy Samir Zaki

Multi-frequency eddy current measurements are employed in estimating pressure tube (PT) to calandria tube (CT) gap in CANDU fuel channels, a critical inspection activity required to ensure fitness for service of fuel channels. In this thesis, a comprehensive characterization of eddy current gap data is laid out, in order to extract further information on fuel channel condition, and to identify generalized applications for multi-frequency eddy current data. A surface profiling technique, generalizable to multiple probe and conductive material configurations has been developed. This technique has allowed for identification of various pressure tube artefacts, has been independently validated (using ultrasonic measurements), and has been deployed and commissioned at Ontario Power Generation. Dodd and Deeds solutions to the electromagnetic boundary value problem associated with the PT to CT gap probe configuration were experimentally validated for amplitude response to changes in gap. Using the validated Dodd and Deeds solutions, principal components analysis (PCA) has been employed to identify independence and redundancies in multi-frequency eddy current data. This has allowed for an enhanced visualization of factors affecting gap measurement. Results of the PCA of simulation data are consistent with the skin depth equation, and are validated against PCA of physical experiments. Finally, compressed data acquisition has been realized, allowing faster data acquisition for multi-frequency eddy current systems with hardware limitations, and is generalizable to other applications where real time acquisition of large data sets is prohibitive.

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

Kreitman, Paul J.; Sirianni, Steve R.; Pillard, Mark M.

Entergy recently performed an Extended Power Up-rate (EPU) on their Grand Gulf Nuclear Station, near Port Gibson, Mississippi. To support the EPU, a new Steam Dryer Assembly was installed during the last refueling outage. Due to limited access into the containment, the large Replacement Steam Dryer (RSD) had to be brought into the containment in pieces and then final assembly was completed on the refueling floor before installation into the reactor. Likewise, the highly contaminated Original Steam Dryer (OSD) had to be segmented into manageable sections, loaded into specially designed shielded containers, and rigged out of containment where they willmore » be safely stored until final disposal is accomplished at an acceptable waste repository. Westinghouse Nuclear Services was contracted by Entergy to segment, package and remove the OSD from containment. This work was performed on critical path during the most recent refueling outage. The segmentation was performed underwater to minimize radiation exposure to the workers. Special hydraulic saws were developed for the cutting operations based on Westinghouse designs previously used in Sweden to segment ABB Reactor Internals. The mechanical cutting method was selected because of its proven reliability and the minimal cutting debris that is generated by the process. Maintaining stability of the large OSD sections during cutting was accomplished using a custom built support stand that was installed into the Moisture Separator Pool after the Moisture Separator was installed back in the reactor vessel. The OSD was then moved from the Steam Dryer Pool to the Moisture Separator Pool for segmentation. This scenario resolved the logistical challenge of having two steam dryers and a moisture separator in containment simultaneously. A water filtration/vacuum unit was supplied to maintain water clarity during the cutting and handling operations and to collect the cutting chips. (authors)« less

High power ring methods and accelerator driven subcritical reactor application

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

Tahar, Malek Haj

2016-08-07

High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g.,more » PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field imperfections on the transverse beam dynamics. The results obtained allow to develop a correction scheme to minimize the tune variations of the FFAG. This is the cornerstone of a new fixed tune non-scaling FFAG that represents a potential candidate for high power applications. As part of the developments towards high power at the KURRI FFAG, beam dynamics studies have to account for space charge effects. In that framework, models have been installed in the tracking code ZGOUBI to account for the self-interaction of the particles in the accelerator. Application to the FFAG studies is shown. Finally, one focused on the ADSR concept as a candidate to solve the problem of nuclear waste. In order to establish the accelerator requirements, one compared the performance of ADSR with other conventional critical reactors by means of the levelized cost of energy. A general comparison between the different accelerator technologies that can satisfy these requirements is finally presented. In summary, the main drawback of the ADSR technology is the high Levelized Cost Of Energy compared to other advanced reactor concepts that do not employ an accelerator. Nowadays, this is a show-stopper for any industrial application aiming at producing energy (without dealing with the waste problem). Besides, the reactor is not intrinsically safer than critical reactor concepts, given the complexity of managing the target interface between the accelerator and the reactor core.« less

75 FR 13 - Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-04

...The Nuclear Regulatory Commission (NRC) is amending its regulations to provide alternate fracture toughness requirements for protection against pressurized thermal shock (PTS) events for pressurized water reactor (PWR) pressure vessels. This final rule provides alternate PTS requirements based on updated analysis methods. This action is desirable because the existing requirements are based on unnecessarily conservative probabilistic fracture mechanics analyses. This action reduces regulatory burden for those PWR licensees who expect to exceed the existing requirements before the expiration of their licenses, while maintaining adequate safety, and may choose to comply with the final rule as an alternative to complying with the existing requirements.

The Outlook for Some Fission Products Utilization with the Aim to Immobilize Long-Lived Radionuclides

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

Pokhitonov, Y.A.

2008-07-01

The prospects for development of nuclear power are intimately associated with solving the problem of safe management and removal from the biosphere of generated radioactive wastes. The most suitable material for fission products and actinides immobilization is the crystalline ceramics. By now numerous literature data are available concerning the synthesis of a large range of various materials with zirconium-based products. It worth mentioning that zirconium is only one of fission products accumulated in the fuel in large amounts. The development of new materials intended for HLW immobilization will allow increasing of radionuclides concentration in solidified product so providing costs reductionmore » at the stage of subsequent storage. At the same time the idea to use for synthesis of compounds, suitable as materials for long-term storage or final disposal of rad-wastes some fission products occurring in spent fuel in considerable amount and capable to form insoluble substances seems to be rather attractive. In authors opinion in the nearest future one can expect the occurrence of publications proposing the techniques allowing the use of 'reactor's zirconium, molybdenum or, perhaps, technetium as well, with the aim of preparing materials suitable for long-lived radionuclides storage or final disposal. The other element, which is generated in the reactor and worth mentioning, is palladium. The prospects for using palladium are defined not only by its higher generation in the reactor, but by a number of its chemical properties as well. It is evident that the use of natural palladium with the purpose of radionuclides immobilization is impossible due to its high cost and deficiency). In author's opinion such materials could be used as targets for long-lived radionuclides transmutation as well. The object of present work was the study on methods that could allow to use 'reactor' palladium with the aim of long-lived radionuclides such as I-129 and TUE immobilization. In the paper the results of experiments on synthesis of matrices with TUE oxides and PdI{sub 2} on palladium base are presented. (authors)« less

Northrop Triga facility decommissioning plan versus actual results

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

Gardner, F.W.

1986-01-01

This paper compares the Triga facility decontamination and decommissioning plan to the actual results and discusses key areas where operational activities were impacted upon by the final US Nuclear Regulatory Commission (NRC)-approved decontamination and decommissioning plan. Total exposures for fuel transfer were a factor of 4 less than planned. The design of the Triga reactor components allowed the majority of the components to be unconditionally released.

The Optimized Integration of the Decontamination Plan and the Radwaste Management Plan into Decommissioning Plan to the VVR-S Research Reactor from Romania

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

Barariu, G.

2008-07-01

The paper presents the progress of the Decontamination Plan and Radioactive Waste Management Plan which accompanies the Decommissioning Plan for research reactor VVR-S located in Magurele, Ilfov, near Bucharest, Romania. The new variant of the Decommissioning Plan was elaborated taking into account the IAEA recommendation concerning radioactive waste management. A new feasibility study for VVR-S decommissioning was also elaborated. The preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as rehabilitation of the existing Radioactive Waste Treatment Plant and the upgrade of the Radioactive Waste Disposal Facilitymore » at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and reusing of cleaned reactor building is envisaged. An inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the IAEA assistance. Environmental concerns are a part of the radioactive waste management strategy. In conclusion: The current version 8 of the Draft Decommissioning Plan which include the Integrated concept of Decontamination and Decommissioning and Radwaste Management, reflects the substantial work that has been incorporated by IFIN-HH in collaboration with SITON, which has resulted in substantial improvement in document The decommissioning strategy must take into account costs for VVR-S Reactor decommissioning, as well as costs for much needed refurbishments to the radioactive waste treatment plant and the Baita-Bihor waste disposal repository. Several improvements to the Baita-Bihor repository and IFIN-HH waste treatment facility were proposed. The quantities and composition of the radioactive waste generated by VVR-S Reactor dismantling were again estimated by streams and the best demonstrated practicable processing solution was proposed. The estimated quantities of materials to be managed in the near future raise some issues that need to be solved swiftly, such as treatment of aluminum and lead and graphite management. It is envisaged that these materials to be treated to Subsidiary for Nuclear Research (SCN) Pitesti. (authors)« less

Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

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

M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department ofmore » Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor use, and the high-level waste from the processing will be stabilized and stored for less than 20 years before being sent back to the Czech Republic for final disposition. Finally, the paper contains a section for the summary and conclusions.« less

Combined process of electrocoagulation and photocatalytic degradation for the treatment of olive washing wastewater.

PubMed

Ates, Hasan; Dizge, Nadir; Yatmaz, H Cengiz

2017-01-01

In this study, an electrocoagulation reactor (ECR) and photocatalytic reactor (PCR) were tested to understand the performance of combined electrocoagulation and photocatalytic-degradation of olive washing wastewater (OWW). The effects of initial pH (6.0, 6.9, 8.0, 9.0), applied voltage (10.0, 12.5, 15.0 V), and operating time (30, 60, 90, 120 min) were investigated in the electrocoagulation reactor when aluminum electrodes were used as both anode and cathode. The pH, conductivity, color, chemical oxygen demand (COD), and phenol were measured versus time to determine the efficiency of the ECR and PCR process. It was observed that electrocoagulation as a single treatment process supplied the COD removal of 62.5%, color removal of 98.1%, and total phenol removal of 87% at optimum conditions as pH 6.9, applied voltage of 12.5 V, and operating time of 120 min. Moreover, final pH and conductivity were 7.7 and 980 μS/cm, respectively. On the other hand, the effect of semiconductor catalyst type (TiO 2 and ZnO) and loading (1, 2, 3 g/L) were tested using PCR as a stand-alone technique. It was found that photocatalytic degradation as a single treatment process when using 1 g/L ZnO achieved the COD removal of 46%, color removal of 99% with a total phenol removal of 41% at optimum conditions. Final pH and conductivity were 6.2 and 915 μS/cm, respectively. Among semiconductor catalysts, TiO 2 and ZnO performed identical efficiencies for both COD and total phenol removal. Moreover, combination in which electrochemical degradation was employed as a pre-treatment to the photocatalytic degradation process obtained high COD removal of 88% and total phenol, as well as color removal of 100% for the OWW. The electrochemical treatment alone was not effective, but in combination with the photocatalytic process, led to a high-quality effluent. Finally, sludge collected from the electrocoagulation process was characterized by attenuated total reflection Fourier transform infrared and X-ray powder diffraction analyses.

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

2017-05-01

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

Biodegradation of 2,4,6-trichlorophenol in a packed-bed biofilm reactor equipped with an internal net draft tube riser for aeration and liquid circulation.

PubMed

Jesús, A Gómez-De; Romano-Baez, F J; Leyva-Amezcua, L; Juárez-Ramírez, C; Ruiz-Ordaz, N; Galíndez-Mayer, J

2009-01-30

For the aerobic biodegradation of the fungicide and defoliant 2,4,6-trichlorophenol (2,4,6-TCP), a bench-scale packed-bed bioreactor equipped with a net draft tube riser for liquid circulation and oxygenation (PB-ALR) was constructed. To obtain a high packed-bed volume relative to the whole bioreactor volume, a high A(D)/A(R) ratio was used. Reactor's downcomer was packed with a porous support of volcanic stone fragments. PB-ALR hydrodynamics and oxygen mass transfer behavior was evaluated and compared to the observed behavior of the unpacked reactor operating as an internal airlift reactor (ALR). Overall gas holdup values epsilon(G), and zonal oxygen mass transfer coefficients determined at various airflow rates in the PB-ALR, were higher than those obtained with the ALR. When comparing mixing time values obtained in both cases, a slight increment in mixing time was observed when reactor was operated as a PB-ALR. By using a mixed microbial community, the biofilm reactor was used to evaluate the aerobic biodegradation of 2,4,6-TCP. Three bacterial strains identified as Burkholderia sp., Burkholderia kururiensis and Stenotrophomonas sp. constituted the microbial consortium able to cometabolically degrade the 2,4,6-TCP, using phenol as primary substrate. This consortium removed 100% of phenol and near 99% of 2,4,6-TCP. Mineralization and dehalogenation of 2,4,6-TCP was evidenced by high COD removal efficiencies ( approximately 95%), and by the stoichiometric release of chloride ions from the halogenated compound ( approximately 80%). Finally, it was observed that the microbial consortium was also capable to metabolize 2,4,6-TCP without phenol as primary substrate, with high removal efficiencies (near 100% for 2,4,6-TCP, 92% for COD and 88% for chloride ions).

Nuclear modules for space electric propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1998-01-01

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

Characterization of reactive CaCO 3 crystallization in a fluidized bed reactor as a central process of direct air capture

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

Burhenne, Luisa; Giacomin, Caroline; Follett, Trevor

A laboratory-scale, fluidized-bed pellet reactor (BPR) was used to investigate a CaCO 3 crystallization process for the recovery of CO 2 in a Direct Air Capture (DAC) process. The BPR performance was validated against data from a pilot-scale unit. Subsequently, the pellet growth under process-relevant conditions was studied over a period of 144 h. The experimental results with the BPR, containing a bed of pellets sized between 0.65 and 0.84 mm, have shown that a calcium retention of 80% can be achieved at a fluidization velocity of 60 m h -1 and a calcium loading rate of 3 mol hmore » -1. This result is consistent with calcium retention observed at pilot scale operation and hence, results from the BPR are considered representative for the pilot scale unit. Starting with a bed of pellets sized between 0.15 and 0.5 mm, the average pellet growth rate, G, at the reactor bottom increased from 8.1E-10 to 11E–10 m s -1 at the onset and decreased to 4.9E–10 m s -1 over the course of a 144 h test. The calcium retention over the course the test showed the same trend (initial increase and final decrease) as the pellet growth rate. A theoretical bed growth model was developed and validated against data from the pilot scale and benchtop pellet reactors. The model was used to calculate the bed porosity and total pellet surface area in each control volume. Lastly, the pellet surface area growth at the bottom of the reactor reproduced the pellet growth and retention data trends.« less

A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors

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

Grabaskas, Dave; Brunett, Acacia J.; Bucknor, Matthew

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release ofmore » radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.« less

Catalyst and process development for synthesis gas conversion to isobutylene. Final report, September 1, 1990--January 31, 1994

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

Anthony, R.G.; Akgerman, A.

1994-05-06

Previous work on isosynthesis (conversion of synthesis gas to isobutane and isobutylene) was performed at very low conversions or extreme process conditions. The objectives of this research were (1) determine the optimum process conditions for isosynthesis; (2) determine the optimum catalyst preparation method and catalyst composition/properties for isosynthesis; (3) determine the kinetics for the best catalyst; (4) develop reactor models for trickle bed, slurry, and fixed bed reactors; and (5) simulate the performance of fixed bed trickle flow reactors, slurry flow reactors, and fixed bed gas phase reactors for isosynthesis. More improvement in catalyst activity and selectivity is needed beforemore » isosynthesis can become a commercially feasible (stand-alone) process. Catalysts prepared by the precipitation method show the most promise for future development as compared with those prepared hydrothermally, by calcining zirconyl nitrate, or by a modified sol-gel method. For current catalysts the high temperatures (>673 K) required for activity also cause the production of methane (because of thermodynamics). A catalyst with higher activity at lower temperatures would magnify the unique selectivity of zirconia for isobutylene. Perhaps with a more active catalyst and acidification, oxygenate production could be limited at lower temperatures. Pressures above 50 atm cause an undesirable shift in product distribution toward heavier hydrocarbons. A model was developed that can predict carbon monoxide conversion an product distribution. The rate equation for carbon monoxide conversion contains only a rate constant and an adsorption equilibrium constant. The product distribution was predicted using a simple ratio of the rate of CO conversion. This report is divided into Introduction, Experimental, and Results and Discussion sections.« less

On the use of a molten salt fast reactor to apply an idealized transmutation scenario for the nuclear phase out.

PubMed

Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara; Litskevich, Dzianis; Scholl, Susanne

2014-01-01

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations--a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described.

On the Use of a Molten Salt Fast Reactor to Apply an Idealized Transmutation Scenario for the Nuclear Phase Out

PubMed Central

Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara; Litskevich, Dzianis; Scholl, Susanne

2014-01-01

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations – a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described. PMID:24690768

Hybrid finite-volume/transported PDF method for the simulation of turbulent reactive flows

NASA Astrophysics Data System (ADS)

Raman, Venkatramanan

A novel computational scheme is formulated for simulating turbulent reactive flows in complex geometries with detailed chemical kinetics. A Probability Density Function (PDF) based method that handles the scalar transport equation is coupled with an existing Finite Volume (FV) Reynolds-Averaged Navier-Stokes (RANS) flow solver. The PDF formulation leads to closed chemical source terms and facilitates the use of detailed chemical mechanisms without approximations. The particle-based PDF scheme is modified to handle complex geometries and grid structures. Grid-independent particle evolution schemes that scale linearly with the problem size are implemented in the Monte-Carlo PDF solver. A novel algorithm, in situ adaptive tabulation (ISAT) is employed to ensure tractability of complex chemistry involving a multitude of species. Several non-reacting test cases are performed to ascertain the efficiency and accuracy of the method. Simulation results from a turbulent jet-diffusion flame case are compared against experimental data. The effect of micromixing model, turbulence model and reaction scheme on flame predictions are discussed extensively. Finally, the method is used to analyze the Dow Chlorination Reactor. Detailed kinetics involving 37 species and 158 reactions as well as a reduced form with 16 species and 21 reactions are used. The effect of inlet configuration on reactor behavior and product distribution is analyzed. Plant-scale reactors exhibit quenching phenomena that cannot be reproduced by conventional simulation methods. The FV-PDF method predicts quenching accurately and provides insight into the dynamics of the reactor near extinction. The accuracy of the fractional time-stepping technique in discussed in the context of apparent multiple-steady states observed in a non-premixed feed configuration of the chlorination reactor.

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

Characterization of reactive CaCO 3 crystallization in a fluidized bed reactor as a central process of direct air capture

DOE PAGES

Burhenne, Luisa; Giacomin, Caroline; Follett, Trevor; ...

2017-10-25

A laboratory-scale, fluidized-bed pellet reactor (BPR) was used to investigate a CaCO 3 crystallization process for the recovery of CO 2 in a Direct Air Capture (DAC) process. The BPR performance was validated against data from a pilot-scale unit. Subsequently, the pellet growth under process-relevant conditions was studied over a period of 144 h. The experimental results with the BPR, containing a bed of pellets sized between 0.65 and 0.84 mm, have shown that a calcium retention of 80% can be achieved at a fluidization velocity of 60 m h -1 and a calcium loading rate of 3 mol hmore » -1. This result is consistent with calcium retention observed at pilot scale operation and hence, results from the BPR are considered representative for the pilot scale unit. Starting with a bed of pellets sized between 0.15 and 0.5 mm, the average pellet growth rate, G, at the reactor bottom increased from 8.1E-10 to 11E–10 m s -1 at the onset and decreased to 4.9E–10 m s -1 over the course of a 144 h test. The calcium retention over the course the test showed the same trend (initial increase and final decrease) as the pellet growth rate. A theoretical bed growth model was developed and validated against data from the pilot scale and benchtop pellet reactors. The model was used to calculate the bed porosity and total pellet surface area in each control volume. Lastly, the pellet surface area growth at the bottom of the reactor reproduced the pellet growth and retention data trends.« less

New reactor cavity cooling system having passive safety features using novel shape for HTGRs and VHTRs

DOE PAGES

Takamatsu, Kuniyoshi; Hu, Rui

2014-11-27

A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features without a requirement for electricity and mechanical drive is proposed for high temperature gas cooled reactors (HTGRs) and very high temperature reactors (VHTRs). The RCCS design consists of continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 (°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV with radiation and natural convection. Employing the air as the working fluid and the ambient airmore » as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting due to overheating the fuels. The simulation results from a commercial CFD code, STAR-CCM+, show that the temperature distribution of the RCCS is within the temperature limits of the structures, such as the maximum operating temperature of the RPV, 713.15 (K) = 440 (°C), and the heat released from the RPV could be removed safely, even during a loss of coolant accident (LOCA). Finally, when the RCCS can remove 600 (kW) of the rated nominal state even during LOCA, the safety review for building the HTTR could confirm that the temperature distribution of the HTTR is within the temperature limits of the structures to secure structures and fuels after the shutdown because the large heat capacity of the graphite core can absorb heat from the fuel in a short period. Therefore, the capacity of the new RCCS design would be sufficient for decay heat removal.« 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

Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

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

Pan, Dongqing; Chien Jen, Tien; Li, Tao

2014-01-15

This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domainmore » with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired.« less

Pratt & Whitney ESCORT derivative for mars surface power

NASA Astrophysics Data System (ADS)

Feller, Gerald J.; Joyner, Russell

1999-01-01

The purpose of this paper is to address the applicability of a common reactor system design from the Pratt & Whitney ESCORT nuclear thermal rocket engine concept to support current NASA mars surface-based power requirements. The ESCORT is a bimodal engine capable of supporting a wide range of propulsive thermal and vehicle electrical power requirements. The ESCORT engine is powered by a fast-spectrum beryllium-reflected CERMET-fueled nuclear reactor. In addition to an expander cycle propulsive mode, the ESCORT is capable of operating in an electrical power mode. In this mode, the reactor is used to heat a mixture of helium and xenon to drive a closed-loop Brayton cycle in order to generate electrical energy. Recent Design Reference Mission requirements (DRM) from NASA Johnson Space Center and NASA Lewis Research Center studies in 1997 and 1998 have detailed upgraded requirements for potential mars transfer missions. The current NASA DRM requires a nuclear thermal propulsion system capable of delivering total mission requirements of 200170 N (45000 lbf) thrust and 50 kWe of spacecraft electrical power. Additionally, these requirements detailed a surface power system capable of providing approximately 160 kW of electrical energy over an approximate 10 year period within a given weight and volume envelope. Current NASA studies use a SP-100 reactor (0.8 MT) and a NERVA derivative (1.6 MT) as baseline systems. A mobile power cart of approximate dimensions 1.7 m×4.5 m×4.4 m has been conceptualized to transport the reactor power system on the Mars Surface. The 63.25 cm diameter and 80.25 cm height of the ESCORT and its 1.3 MT of weight fit well within the current weight and volume target range of the NASA DRM requirements. The modifications required to the ESCORT reactor system to support this upgraded electrical power requirements along with operation in the Martian atmospheric conditions are addressed in this paper. Sufficient excess reactivity and burnup capability were designed into the ESCORT reactor system to support these upgraded requirements. Only slight modifications to reactor hardware were required to address any environmental considerations. These modifications involved sealing any refractory metal alloy components from the CO2 in the Martian Atmosphere. Also, the Brayton cycle Power Conversion Unit (PCU) hardware was modified to support the upgraded requirements. This paper discusses the design analysis performed and provides information on the final common reactor concept to be used on the Mars surface to support manned missions.

Westinghouse Small Modular Reactor nuclear steam supply system design

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

Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development andmore » integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)« less

Vectorized Monte Carlo methods for reactor lattice analysis

NASA Technical Reports Server (NTRS)

Brown, F. B.

1984-01-01

Some of the new computational methods and equivalent mathematical representations of physics models used in the MCV code, a vectorized continuous-enery Monte Carlo code for use on the CYBER-205 computer are discussed. While the principal application of MCV is the neutronics analysis of repeating reactor lattices, the new methods used in MCV should be generally useful for vectorizing Monte Carlo for other applications. For background, a brief overview of the vector processing features of the CYBER-205 is included, followed by a discussion of the fundamentals of Monte Carlo vectorization. The physics models used in the MCV vectorized Monte Carlo code are then summarized. The new methods used in scattering analysis are presented along with details of several key, highly specialized computational routines. Finally, speedups relative to CDC-7600 scalar Monte Carlo are discussed.

Characterization of metal adsorption kinetic properties in batch and fixed-bed reactors.

PubMed

Chen, J Paul; Wang, Lin

2004-01-01

Copper adsorption kinetic properties in batch and fixed-bed reactors were studied in this paper. The isothermal adsorption experiments showed that the copper adsorption capacity of a granular activated carbon (Filtrasorb 200) increased when ionic strength was higher. The presence of EDTA diminished the adsorption. An intraparticle diffusion model and a fixed-bed model were successfully used to describe the batch kinetic and fixed-bed operation behaviors. The kinetics became faster when the solution pH was not controlled, implying that the surface precipitation caused some metal uptake. The external mass transfer coefficient, the diffusivity and the dispersion coefficient were obtained from the modeling. It was found that both external mass transfer and dispersion coefficients increased when the flow rate was higher. Finally effects of kinetic parameters on simulation of fixed-bed operation were conducted.

Metallurgical failure analysis of MH-1A reactor core hold-down bolts. Final report

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

Hawthorne, J.R.; Watson, H.E.

1976-11-01

The Naval Research Laboratory has performed a failure analysis on two MH-1A reactor core hold-down bolts that broke in service. Adherence to fabrication specifications, post-service properties and possible causes of bolt failure were investigated. The bolt material was verified as 17-4PH precipitation hardening stainless steel. Measured bolt dimensions also were in accordance with fabrication drawing specifications. Bolt failure occurred in the region of a locking pin hole which reduced the bolt net section by 47 percent. The failure analysis indicates that the probable cause of failure was net section overloading resulting from a lateral bending force on the bolt. Themore » analysis indicates that net section overloading could also have resulted from combined tensile stresses (bolt preloading plus differential thermal expansion). Recommendations are made for improved bolting.« less

A Comparison of Brayton and Stirling Space Nuclear Power Systems for Power Levels from 1 Kilowatt to 10 Megawatts

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2000-01-01

An analytical study was conducted to assess the performance and mass of Brayton and Stirling nuclear power systems for a wide range of future NASA space exploration missions. The power levels and design concepts were based on three different mission classes. Isotope systems, with power levels from 1 to 10 kW, were considered for planetary surface rovers and robotic science. Reactor power systems for planetary surface outposts and bases were evaluated from 10 to 500 kW. Finally, reactor power systems in the range from 100 kW to 10 mW were assessed for advanced propulsion applications. The analysis also examined the effect of advanced component technology on system performance. The advanced technologies included high temperature materials, lightweight radiators, and high voltage power management and distribution.

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

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

Jacobsen, George; Back, Christina

2015-10-30

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called themore » endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.« less

Rover/NERVA-derived near-term nuclear propulsion

NASA Technical Reports Server (NTRS)

1993-01-01

FY-92 accomplishments centered on conceptual design and analyses for 25, 50, and 75 K engines with emphasis on the 50 K engine. During the first period of performance, flow and energy balances were prepared for each of these configurations and thrust-to-weight values were estimated. A review of fuel technology and key data from the Rover/NERVA program established a baseline for proven reactor performance and areas of enhancement to meet near-term goals. Studies were performed of the criticality and temperature profiles for probable fuel and moderator loadings for the three engine sizes, with a more detailed analysis of the 50 K size. During the second period of performance, analyses of the 50 K engine continued. A chamber/nozzle contour was selected and heat transfer and fatigue analyses were performed for likely construction materials. Reactor analyses were performed to determine component radiation heating rates, reactor radiation fields, water immersion poisoning requirements, temperature limits for restartability, and a tie-tube thermal analysis. Finally, a brief assessment of key enabling technologies was made, with a view toward identifying development issues and identification of the critical path toward achieving engine qualification within 10 years.

Efficient removal of antibiotics in a fluidized bed reactor by facile fabricated magnetic powdered activated carbon.

PubMed

Ma, Jianqing; Yang, Qunfeng; Xu, Dongmei; Zeng, Xiaomei; Wen, Yuezhong; Liu, Weiping

2017-02-01

Powdered activated carbons (PACs) with micrometer size are showing great potential for enabling and improving technologies in water treatment. The critical problem in achieving practical application of PAC involves simple, effective fabrication of magnetic PAC and the design of a feasible reactor that can remove pollutants and recover the adsorbent efficiently. Herein, we show that such materials can be fabricated by the combination of PAC and magnetic Fe 3 O 4 with chitosan-Fe hydrogel through a simple co-precipitation method. According to the characterization results, CS-Fe/Fe 3 O 4 /PAC with different micrometers in size exhibited excellent magnetic properties. The adsorption of tetracycline was fast and efficient, and 99.9% removal was achieved in 30 min. It also possesses good usability and stability to co-existing ions, organics, and different pH values due to its dispersive interaction nature. Finally, the prepared CS-Fe/Fe 3 O 4 /PAC also performed well in the fluidized bed reactor with electromagnetic separation function. It could be easily separated by applying a magnetic field and was effectively in situ regenerated, indicating a potential of practical application for the removal of pollutants from water.

[Studies on photo-electron-chemical catalytic degradation of the malachite green].

PubMed

Li, Ming-yu; Diao, Zeng-hui; Song, Lin; Wang, Xin-le; Zhang, Yuan-ming

2010-07-01

A novel two-compartment photo-electro-chemical catalytic reactor was designed. The TiO2/Ti thin film electrode thermally formed was used as photo-anode, and graphite as cathode and a saturated calomel electrode (SCE) as the reference electrode in the reactor. The anode compartment and cathode compartment were connected with the ionic exchange membrane in this reactor. Effects of initial pH, initial concentration of malachite green and connective modes between the anode compartment and cathode compartment on the decolorization efficiency of malachite green were investigated. The degradation dynamics of malachite green was studied. Based on the change of UV-visible light spectrum, the degradation process of malachite green was discussed. The experimental results showed that, during the time of 120 min, the decolouring ratio of the malachite green was 97.7% when initial concentration of malachite green is 30 mg x L(-1) and initial pH is 3.0. The catalytic degradation of malachite green was a pseudo-first order reaction. In the degradation process of malachite green the azo bond cleavage and the conjugated system of malachite green were attacked by hydroxyl radical. Simultaneity, the aromatic ring was oxidized. Finally, malachite green was degraded into other small molecular compounds.

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

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

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings withoutmore » and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.« less

Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

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

Fast, Ivan; Bosbach, Dirk; Aksyutina, Yuliya

A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculationsmore » accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)« less

ASME Material Challenges for Advanced Reactor Concepts

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

Piyush Sabharwall; Ali Siahpush

2013-07-01

This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at highermore » temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.« less

Captives of Their Fantasies: The German Atomic Bomb Scientists

NASA Astrophysics Data System (ADS)

Klotz, Irving M.

1997-02-01

When the Nazi government collapsed in May, 1945, an Allied intelligence mission took into custody nine of the German scientists who played key roles in the German atomic bomb project. Under great secrecy these men were confined in a large country house, Farm Hall, near Cambridge (England), and their conversations were recorded surreptitiously by hidden microphones in every room. The transcripts were kept TOP SECRET for 47 years and were finally released recently. They give fascinating insights into the personalities of the guests and invaluable information on what the Germans really understood about the physics and chemistry of a nuclear reactor and an atomic bomb. The Farm Hall transcripts clearly establish that (a) the Germans on August 6, 1945 did not believe that the Allies had exploded an atomic bomb over Hiroshima that day; (b) they never succeeded in constructing a self-sustaining nuclear reactor; (c) they were confused about the differences between an atomic bomb and a reactor; (d) they did not know how to correctly calculate the critical mass of a bomb; (e) they thought that "plutonium" was probably element 91. The Farm Hall transcripts contradict the self-serving and sensationalist writings about German efforts that have appeared during the past fifty years.

Synthesis of biodiesel from waste cooking oil using sonochemical reactors.

PubMed

Hingu, Shishir M; Gogate, Parag R; Rathod, Virendra K

2010-06-01

Investigation into newer routes of biodiesel synthesis is a key research area especially due to the fluctuations in the conventional fuel prices and the environmental advantages of biodiesel. The present work illustrates the use of sonochemical reactors for the synthesis of biodiesel from waste cooking oil. Transesterification of used frying oil with methanol, in the presence of potassium hydroxide as a catalyst has been investigated using low frequency ultrasonic reactor (20 kHz). Effect of different operating parameters such as alcohol-oil molar ratio, catalyst concentration, temperature, power, pulse and horn position on the extent of conversion of oil have been investigated. The optimum conditions for the transesterification process have been obtained as molar ratio of alcohol to oil as 6:1, catalyst concentration of 1 wt.%, temperature as 45 degrees C and ultrasound power as 200 W with an irradiation time of 40 min. The efficacy of using ultrasound has been compared with the conventional stirring approach based on the use of a six blade turbine with diameter of 1.5 cm operating at 1000 rpm. Also the purification aspects of the final product have been investigated. (c) 2010 Elsevier B.V. All rights reserved.

High-strength wastewater treatment in a pure oxygen thermophilic process: 11-year operation and monitoring of different plant configurations.

PubMed

Collivignarelli, M C; Bertanza, G; Sordi, M; Pedrazzani, R

2015-01-01

This research was carried out on a full-scale pure oxygen thermophilic plant, operated and monitored throughout a period of 11 years. The plant treats 60,000 t y⁻¹ (year 2013) of high-strength industrial wastewaters deriving mainly from pharmaceuticals and detergents production and landfill leachate. Three different plant configurations were consecutively adopted: (1) biological reactor + final clarifier and sludge recirculation (2002-2005); (2) biological reactor + ultrafiltration: membrane biological reactor (MBR) (2006); and (3) MBR + nanofiltration (since 2007). Progressive plant upgrading yielded a performance improvement chemical oxygen demand (COD) removal efficiency was enhanced by 17% and 12% after the first and second plant modification, respectively. Moreover, COD abatement efficiency exhibited a greater stability, notwithstanding high variability of the influent load. In addition, the following relevant outcomes appeared from the plant monitoring (present configuration): up to 96% removal of nitrate and nitrite, due to denitrification; low-specific biomass production (0.092 kgVSS kgCODremoved⁻¹), and biological treatability of residual COD under mesophilic conditions (BOD5/COD ratio = 0.25-0.50), thus showing the complementarity of the two biological processes.

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

DOE PAGES

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.; ...

2016-03-16

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings withoutmore » and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.« less

In situ bioremediation in Europe

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

Porta, A.; Young, J.K.; Molton, P.M.

1993-06-01

Site remediation activity in Europe is increasing, even if not at the forced pace of the US. Although there is a better understanding of the benefits of bioremediation than of other approaches, especially about in situ bioremediation of contaminated soils, relatively few projects have been carried out full-scale in Europe or in the US. Some engineering companies and large industrial companies in Europe are investigating bioremediation and biotreatment technologies, in some cases to solve their internal waste problems. Technologies related to the application of microorganisms to the soil, release of nutrients into the soil, and enhancement of microbial decontamination aremore » being tested through various additives such as surfactants, ion exchange resins, limestone, or dolomite. New equipment has been developed for crushing and mixing or injecting and sparging the microorganisms, as have new reactor technologies (e.g., rotating aerator reactors, biometal sludge reactors, and special mobile containers for simultaneous storage, transportation, and biodegradation of contaminated soil). Some work has also been done with immobilized enzymes to support and restore enzymatic activities related to partial or total xenobiotic decontamination. Finally, some major programs funded by public and private institutions confirm that increasing numbers of firms have a working interest in bioremediation.« less

Thermo-physics technical note No. 37: SNAP- 10A, Stainless Steel-316 vessel wall ablation. Final report

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

Montgomery, L.D.

1964-12-07

The altitudes and times of ablation have been determined for the SNAP-10A, SS-316 vessel wall reentering under various conditions. The results are confined to one typical location on the reactor and to one typical reentry trajectory. The location is the side wall of the vessel and the trajectory is the one used in NAA-SR-8303.

Nuclear thermal propulsion transportation systems for lunar/Mars exploration

NASA Technical Reports Server (NTRS)

Clark, John S.; Borowski, Stanley K.; Mcilwain, Melvin C.; Pellaccio, Dennis G.

1992-01-01

Nuclear thermal propulsion technology development is underway at NASA and DoE for Space Exploration Initiative (SEI) missions to Mars, with initial near-earth flights to validate flight readiness. Several reactor concepts are being considered for these missions, and important selection criteria will be evaluated before final selection of a system. These criteria include: safety and reliability, technical risk, cost, and performance, in that order. Of the concepts evaluated to date, the Nuclear Engine for Rocket Vehicle Applications (NERVA) derivative (NDR) is the only concept that has demonstrated full power, life, and performance in actual reactor tests. Other concepts will require significant design work and must demonstrate proof-of-concept. Technical risk, and hence, development cost should therefore be lowest for the concept, and the NDR concept is currently being considered for the initial SEI missions. As lighter weight, higher performance systems are developed and validated, including appropriate safety and astronaut-rating requirements, they will be considered to support future SEI application. A space transportation system using a modular nuclear thermal rocket (NTR) system for lunar and Mars missions is expected to result in significant life cycle cost savings. Finally, several key issues remain for NTR's, including public acceptance and operational issues. Nonetheless, NTR's are believed to be the 'next generation' of space propulsion systems - the key to space exploration.

Impact of the proposed energy tax on nuclear electric generating technologies

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

Edmunds, T.A.; Lamont, A.D.; Pasternak, A.D.

1993-05-01

The President`s new economic initiatives include an energy tax that will affect the costs of power from most electric generating technologies. The tax on nuclear power could be applied in a number of different ways at several different points in the fuel cycle. These different approaches could have different effects on the generation costs and benefits of advanced reactors. The Office of Nuclear Energy has developed models for assessing the costs and benefits of advanced reactor cycles which must be updated to take into account the impacts of the proposed tax. This report has been prepared to assess the spectrummore » of impacts of the energy tax on nuclear power and can be used in updating the Office`s economic models. This study was conducted in the following steps. First, the most authoritative statement of the proposed tax available at this time was obtained. Then the impacts of the proposed tax on the costs of nuclear and fossil fueled generation were compared. Finally several other possible approaches to taxing nuclear energy were evaluated. The cost impact on several advanced nuclear technologies and a current light water technology were computed. Finally, the rationale for the energy tax as applied to various electric generating methods was examined.« less

Evaluation of Neutron-induced Cross Sections and their Related Covariances with Physical Constraints

NASA Astrophysics Data System (ADS)

De Saint Jean, C.; Archier, P.; Privas, E.; Noguère, G.; Habert, B.; Tamagno, P.

2018-02-01

Nuclear data, along with numerical methods and the associated calculation schemes, continue to play a key role in reactor design, reactor core operating parameters calculations, fuel cycle management and criticality safety calculations. Due to the intensive use of Monte-Carlo calculations reducing numerical biases, the final accuracy of neutronic calculations increasingly depends on the quality of nuclear data used. This paper gives a broad picture of all ingredients treated by nuclear data evaluators during their analyses. After giving an introduction to nuclear data evaluation, we present implications of using the Bayesian inference to obtain evaluated cross sections and related uncertainties. In particular, a focus is made on systematic uncertainties appearing in the analysis of differential measurements as well as advantages and drawbacks one may encounter by analyzing integral experiments. The evaluation work is in general done independently in the resonance and in the continuum energy ranges giving rise to inconsistencies in evaluated files. For future evaluations on the whole energy range, we call attention to two innovative methods used to analyze several nuclear reaction models and impose constraints. Finally, we discuss suggestions for possible improvements in the evaluation process to master the quantification of uncertainties. These are associated with experiments (microscopic and integral), nuclear reaction theories and the Bayesian inference.

Development of Macroscale Models of UO 2 Fuel Sintering and Densification using Multiscale Modeling and Simulation

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

Greenquist, Ian; Tonks, Michael

2016-10-01

Light water reactor fuel pellets are fabricated using sintering to final densities of 95% or greater. During reactor operation, the porosity remaining in the fuel after fabrication decreases further due to irradiation-assisted densification. While empirical models have been developed to describe this densification process, a mechanistic model is needed as part of the ongoing work by the NEAMS program to develop a more predictive fuel performance code. In this work we will develop a phase field model of sintering of UO 2 in the MARMOT code, and validate it by comparing to published sintering data. We will then add themore » capability to capture irradiation effects into the model, and use it to develop a mechanistic model of densification that will go into the BISON code and add another essential piece to the microstructure-based materials models. The final step will be to add the effects of applied fields, to model field-assisted sintering of UO 2. The results of the phase field model will be validated by comparing to data from field-assisted sintering. Tasks over three years: 1. Develop a sintering model for UO 2 in MARMOT 2. Expand model to account for irradiation effects 3. Develop a mechanistic macroscale model of densification for BISON« less

Development of process control capability through the Browns Ferry Integrated Computer System using Reactor Water Clanup System as an example. Final report

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

Smith, J.; Mowrey, J.

1995-12-01

This report describes the design, development and testing of process controls for selected system operations in the Browns Ferry Nuclear Plant (BFNP) Reactor Water Cleanup System (RWCU) using a Computer Simulation Platform which simulates the RWCU System and the BFNP Integrated Computer System (ICS). This system was designed to demonstrate the feasibility of the soft control (video touch screen) of nuclear plant systems through an operator console. The BFNP Integrated Computer System, which has recently. been installed at BFNP Unit 2, was simulated to allow for operator control functions of the modeled RWCU system. The BFNP Unit 2 RWCU systemmore » was simulated using the RELAP5 Thermal/Hydraulic Simulation Model, which provided the steady-state and transient RWCU process variables and simulated the response of the system to control system inputs. Descriptions of the hardware and software developed are also included in this report. The testing and acceptance program and results are also detailed in this report. A discussion of potential installation of an actual RWCU process control system in BFNP Unit 2 is included. Finally, this report contains a section on industry issues associated with installation of process control systems in nuclear power plants.« less

Decontamination and decommissioning of the Mayaguez (Puerto Rico) facility

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

Jackson, P.K.; Freemerman, R.L.

1989-11-01

On February 6, 1987 the US Department of Energy (DOE) awarded the final phase of the decontamination and decommissioning of the nuclear and reactor facilities at the Center for Energy and Environmental Research (CEER), in Mayaguez, Puerto Rico. Bechtel National, Inc., was made the decontamination and decommissioning (D and D) contractor. The goal of the project was to enable DOE to proceed with release of the CEER facility for use by the University of Puerto Rico, who was the operator. This presentation describes that project and lesson learned during its progress. The CEER facility was established in 1957 as themore » Puerto Rico Nuclear Center, a part of the Atoms for Peace Program. It was a nuclear training and research institution with emphasis on the needs of Latin America. It originally consisted of a 1-megawatt Materials Testing Reactor (MTR), support facilities and research laboratories. After eleven years of operation the MTR was shutdown and defueled. A 2-megawatt TRIGA reactor was installed in 1972 and operated until 1976, when it woo was shutdown. Other radioactive facilities at the center included a 10-watt homogeneous L-77 training reactor, a natural uranium graphite-moderated subcritical assembly, a 200KV particle accelerator, and a 15,000 Ci Co-60 irradiation facility. Support facilities included radiochemistry laboratories, counting rooms and two hot cells. As the emphasis shifted to non-nuclear energy technology a name change resulted in the CEER designation, and plans were started for the decontamination and decommissioning effort.« less

The extraction of bitumen from western oil sands: Volume 2. Final report

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

Oblad, A.G.; Dahlstrom, D.A.; Deo, M.D.

1997-11-26

The program is composed of 20 projects, of which 17 are laboratory bench or laboratory pilot scale processes or computer process simulations that are performed in existing facilities on the University of Utah campus in north-east Salt Lake City. These tasks are: (1) coupled fluidized-bed bitumen recovery and coked sand combustion; (2) water-based recovery of bitumen; (3) oil sand pyrolysis in a continuous rotary kiln reactor; (4) oil sand pyrolysis in a large diameter fluidized bed reactor; (5) oil sand pyrolysis in a small diameter fluidized bed reactor; (6) combustion of spent sand in a transport reactor; (7) recovery andmore » upgrading of oil sand bitumen using solvent extraction methods; (8) fixed-bed hydrotreating of Uinta Basin bitumens and bitumen-derived hydrocarbon liquids; (9) ebullieted bed hydrotreating of bitumen and bitumen derived liquids; (10) bitumen upgrading by hydropyrolysis; (11) evaluation of Utah`s major oil sand deposits for the production of asphalt, high-energy jet fuels and other specialty products; (12) characterization of the bitumens and reservoir rocks from the Uinta Basin oil sand deposits; (13) bitumen upgrading pilot plant recommendations; (14) liquid-solid separation and fine tailings thickening; (15) in-situ production of heavy oil from Uinta Basin oil sand deposits; (16) oil sand research and development group analytical facility; and (17) process economics. This volume contains reports on nine of these projects, references, and a bibliography. 351 refs., 192 figs., 65 tabs.« less

Regional synchrony in full-scale activated sludge bioreactors due to deterministic microbial community assembly

PubMed Central

Griffin, James S; Wells, George F

2017-01-01

Seasonal community structure and regionally synchronous population dynamics have been observed in natural microbial ecosystems, but have not been well documented in wastewater treatment bioreactors. Few studies of community dynamics in full-scale activated sludge systems facing similar meteorological conditions have been done to compare the importance of deterministic and neutral community assembly mechanisms. We subjected weekly activated sludge samples from six regional full-scale bioreactors at four wastewater treatment plants obtained over 1 year to Illumina sequencing of 16S ribosomal RNA genes, resulting in a library of over 17 million sequences. All samples derived from reactors treating primarily municipal wastewater. Despite variation in operational characteristics and location, communities displayed temporal synchrony at the individual operational taxonomic unit (OTU), broad phylogenetic affiliation and community-wide scale. Bioreactor communities were dominated by 134 abundant and highly regionally synchronized OTU populations that accounted for over 50% of the total reads. Non-core OTUs displayed abundance-dependent population synchrony. Alpha diversity varied by reactor, but showed a highly reproducible and synchronous seasonal fluctuation. Community similarity was dominated by seasonal changes, but individual reactors maintained minor stable differences after 1 year. Finally, the impacts of mass migration driven by direct biomass transfers between reactors was investigated, but had no significant effect on community similarity or diversity in the sink community. Our results show that population dynamics in activated sludge bioreactors are consistent with niche-driven assembly guided by seasonal temperature fluctuations. PMID:27996980

Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

NASA Astrophysics Data System (ADS)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

2016-10-01

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

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

Dawson, D.M.; Guerra, G.; Neider, T.

1995-12-01

This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity ormore » other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose.« less

The detector system of the Daya Bay reactor neutrino experiment

DOE PAGES

An, F. P.

2015-12-15

The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin 22θ 13 and the effective mass splitting Δm 2 ee. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrummore » due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors’ baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This study describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.« less

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

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

NONE

1995-08-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized intomore » six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.« less

High efficiency chlorine removal from polyvinyl chloride (PVC) pyrolysis with a gas-liquid fluidized bed reactor.

PubMed

Yuan, G; Chen, D; Yin, L; Wang, Z; Zhao, L; Wang, J Y

2014-06-01

In this research a gas-liquid fluidized bed reactor was developed for removing chlorine (Cl) from polyvinyl chloride (PVC) to favor its pyrolysis treatment. In order to efficiently remove Cl within a limited time before extensive generation of hydrocarbon products, the gas-liquid fluidized bed reactor was running at 280-320 °C, where hot N2 was used as fluidizing gas to fluidize the molten polymer, letting the molten polymer contact well with N2 to release Cl in form of HCl. Experimental results showed that dechlorination efficiency is mainly temperature dependent and 300 °C is a proper reaction temperature for efficient dechlorination within a limited time duration and for prevention of extensive pyrolysis; under this temperature 99.5% of Cl removal efficiency can be obtained within reaction time around 1 min after melting is completed as the flow rate of N2 gas was set around 0.47-0.85 Nm(3) kg(-1) for the molten PVC. Larger N2 flow rate and additives in PVC would enhance HCl release but did not change the final dechlorination efficiency; and excessive N2 flow rate should be avoided for prevention of polymer entrainment. HCl is emitted from PVC granules or scraps at the mean time they started to melt and the melting stage should be taken into consideration when design the gas-liquid fluidized bed reactor for dechlorination. Copyright © 2013 Elsevier Ltd. All rights reserved.

Economical Production of Pu-238: NIAC Phase I Final Report

NASA Technical Reports Server (NTRS)

Howe, Steven D.; Crawford, Douglas; Navarro, Jorge; O'Brien, Robert C.; Katalenich, Jeff; Ring, Terry

2016-01-01

All space exploration missions traveling beyond Jupiter must use radioisotopic power sources for electrical power. The best isotope to power these sources is plutonium-238 (Pu-238). The US supply of Pu-238 is almost exhausted and will be gone within the next decade. The Department of Energy has initiated a production program with a $10M allocation from NASA but the cost is estimated at over $100M to get to production levels. The Center for Space Nuclear Research (CSNR) has conceived of a potentially better process to produce Pu-238 earlier and for significantly less cost. Potentially, the front end capital costs could be provided by private industry such that the government only had to pay for the product produced. In the Phase I NIAC (NASA Innovative Advanced Concepts) grant, the CSNR has evaluated the feasibility of using a low power, commercially available nuclear reactor to produce 1.5 kg of Pu-238 per year. The impact on the neutronics of the reactor have been assessed, the amount of Neptunium target material estimated, and the production rates calculated. In addition, the size of the post-irradiation processing facility has been established. Finally, as the study progressed, a new method for fabricating the Pu-238 product into the form used for power sources has been identified to reduce the cost of the final product. In short, the concept appears to be viable, can produce the amount of Pu-238 needed to support the NASA missions, can be available within a few years, and will cost significantly less than the current DOE program.

The diversity and unit of reactor noise theory

NASA Astrophysics Data System (ADS)

Kuang, Zhifeng

The study of reactor noise theory concerns questions about cause and effect relationships, and utilisation of random noise in nuclear reactor systems. The diversity of reactor noise theory arises from the variety of noise sources, the various mathematical treatments applied and various practical purposes. The neutron noise in zero- energy systems arises from the fluctuations in the number of neutrons per fission, the time between nuclear events, and the type of reactions. It can be used to evaluate system parameters. The mathematical treatment is based on the master equation of stochastic branching processes. The noise in power reactor systems is given rise by random processes of technological origin such as vibration of mechanical parts, boiling of the coolant, fluctuations of temperature and pressure. It can be used to monitor reactor behaviour with the possibility of detecting malfunctions at an early stage. The mathematical treatment is based on the Langevin equation. The unity of reactor noise theory arises from the fact that useful information from noise is embedded in the second moments of random variables, which lends the possibility of building up a unified mathematical description and analysis of the various reactor noise sources. Exploring such possibilities is the main subject among the three major topics reported in this thesis. The first subject is within the zero power noise in steady media, and we reported on the extension of the existing theory to more general cases. In Paper I, by use of the master equation approach, we have derived the most general Feynman- and Rossi-alpha formulae so far by taking the full joint statistics of the prompt and all the six groups of delayed neutron precursors, and a multiple emission source into account. The involved problems are solved with a combination of effective analytical techniques and symbolic algebra codes (Mathematica). Paper II gives a numerical evaluation of these formulae. An assessment of the contribution of the terms that are novel as compared to the traditional formulae has been made. The second subject treats a problem in power reactor noise with the Langevin formalism. With a very few exceptions, in all previous work the diffusion approximation was used. In order to extend the treatment to transport theory, in Paper III, we introduced a novel method, i.e. Padé approximation via Lanczos algorithm to calculate the transfer function of a finite slab reactor described by one-group transport equation. It was found that the local-global decomposition of the neutron noise, formerly only reproduced in at least 2- group theory, can be reconstructed. We have also showed the existence of a boundary layer of the neutron noise close to the boundary. Finally, we have explored the possibility of building up a unified theory to account for the coexistence of zero power and power reactor noise in a system. In Paper IV, a unified description of the neutron noise is given by the use of backward master equations in a model where the cross section fluctuations are given as a simple binary pseudorandom process. The general solution contains both the zero power and power reactor noise concurrently, and they can be extracted individually as limiting cases of the general solution. It justified the separate treatments of zero power and power reactor noise. The result was extended to the case including one group of delayed neutron precursors in Paper V.

Final Report for the “WSU Neutron Capture Therapy Facility Support”

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

Gerald E. Tripard; Keith G. Fox

2006-08-24

The objective for the cooperative research program for which this report has been written was to provide separate NCT facility user support for the students, faculty and scientists who would be doing the U.S. Department of Energy Office (DOE) of Science supported advanced radiotargeted research at the WSU 1 megawatt TRIGA reactor. The participants were the Idaho National laboratory (INL, P.I., Dave Nigg), the Veterinary Medical Research Center of Washington State University (WSU, Janean Fidel and Patrick Gavin), and the Washington State University Nuclear Radiation Center (WSU, P.I., Gerald Tripard). A significant number of DOE supported modifications were made tomore » the WSU reactor in order to create an epithermal neutron beam while at the same time maintaining the other activities of the 1 MW reactor. These modifications were: (1) Removal of the old thermal column. (2) Construction and insertion of a new epithermal filter, collimator and shield. (3) Construction of a shielded room that could accommodate the very high radiation field created by an intense neutron beam. (4) Removal of the previous reactor core fuel cluster arrangement. (5) Design and loading of the new reactor core fuel cluster arrangement in order to optimize the neutron flux entering the epithermal neutron filter. (6) The integration of the shielded rooms interlocks and radiological controls into the SCRAM chain and operating electronics of the reactor. (7) Construction of a motorized mechanism for moving and remotely controlling the position of the entire reactor bridge. (8) The integration of the reactor bridge control electronics into the SCRAM chain and operating electronics of the reactor. (9) The design, construction and attachment to the support structure of the reactor of an irradiation box that could be inserted into position next to the face of the reactor. (Necessitated by the previously mentioned core rearrangement). All of the above modifications were successfully completed and tested. The resulting epithermal beam of 1 x 10{sup 9} n/sec-cm{sup 2} was measured by Idaho National Laboratory with assistance from WSU's Neutron Activation Analysis Group. The beam is as good as our initial proposals for the project had predicted. In addition to all of the design, construction and insertion of the hardware, shielding, electronics and radiation monitoring systems there was considerable manpower and effort put into changes in the Technical Specifications of the reactor and implementing procedures for use of the new facility. This staff involvement is one of the reasons we requested special facility support from the DOE. Once the facility was competed and all of the recalibrations and measurements made to characterize the differences between this reactor core and the previous core we began to assist INL in making their beam measurements with foils and phantoms. Although we proposed support for only one additional staff position to support this new NCT facility the staff support provided by the WSU Nuclear Radiation Center was greater than had been anticipated by our initial proposal. INL was also assisted in the testing of a heavy water (deuterated water) bladder that can be inserted into the collimator in order to produce an intense, external thermal neutron beam. The external epithermal and/or thermal neutron beam capability remains available for use, if funding becomes available for future research projects.« less

Methods and strategies for future reactor safety goals

NASA Astrophysics Data System (ADS)

Arndt, Steven Andrew

There have been significant discussions over the past few years by the United States Nuclear Regulatory Commission (NRC), the Advisory Committee on Reactor Safeguards (ACRS), and others as to the adequacy of the NRC safety goals for use with the next generation of nuclear power reactors to be built in the United States. The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the United States. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have been used as surrogates for the QHOs. In its review of the new plant licensing policy the ACRS has looked at the safety goals, as has the NRC. A number of issues have been raised including what the Commission had in mind when it drafted the safety goals and QHOs, how risk from multiple reactors at a site should be combined for evaluation, how the combination of a new and old reactor at the same site should be evaluated, what the criteria for evaluating new reactors should be, and whether new reactors should be required to be safer than current generation reactors. As part of the development and application of the NRC safety goal policy statement the Commissioners laid out the expectations for the safety of a nuclear power plant but did not address the risk associated with current multi-unit sites, potential modular reactor sites, and hybrid sites that could contain current generation reactors, new passive reactors, and/or modular reactors. The NRC safety goals and the QHOs refer to a "nuclear power plant," but do not discuss whether a "plant" refers to only a single unit or all of the units on a site. There has been much discussion on this issue recently due to the development of modular reactors. Additionally, the risk of multiple reactor accidents on the same site has been largely ignored in the probabilistic risk assessments (PRAs) done to date, and in most risk-informed analyses and discussions. This dissertation examines potential approaches to updating the safety goals that include the establishment of new quantitative safety goal associated with the comparative risk of generating electricity by viable competing technologies and modifications of the goals to account for multi-plant reactor sites, and issues associated with the use of safety goals in both initial licensing and operational decision making. This research develops a new quantitative health objective that uses a comparable benefit risk metric based on the life-cycle risk of the construction, operation and decommissioning of a comparable non-nuclear electric generation facility, as well as the risks associated with mining and transportation. This dissertation also evaluates the effects of using various methods for aggregating site risk as a safety metric, as opposed to using single plant safety goals. Additionally, a number of important assumptions inherent in the current safety goals, including the effect of other potential negative societal effects such as the generation of greenhouse gases (e.g., carbon dioxide) have on the risk of electric power production and their effects on the setting of safety goals, is explored. Finally, the role risk perception should play in establishing safety goals has been explored. To complete this evaluation, a new method to analytically compare alternative technologies of generating electricity was developed, including development of a new way to evaluate risk perception, and a new method was developed for evaluating the risk at multiple units on a single site. To test these modifications to the safety goals a number of possible reactor designs and configurations were evaluated using these new proposed safety goals to determine the goals' usefulness and utility. The results of the analysis showed that the modifications provide measures that more closely evaluate the potential risk to the public from the operation of nuclear power plants than the current safety goals, while still providing a straight-forward process for assessment of reactor design and operation.

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

Kiselyov, V.A.; Sokov, L.M.

The LBB regulatory approach adopted in Russia in 1993 as an extra safety barrier is described for advanced WWER 1000 reactor steamline. The application of LBB concept requires the following additional protections. First, the steamline should be a highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that it is not subjected to any disqualifying failure mechanism. Second, a deterministic fracture mechanics analysis and leak rate evaluation have been performed to demonstrate that postulated through-wall crack that yields 95 1/min at normal operation conditions is stable even under seismic loads. Finally, it hasmore » been verified that the leak detection systems are sufficiently reliable, diverse and sensitive, and that adequate margins exist to detect a through wall crack smaller than the critical size. The obtained results are encouraging and show the possibility of the application of the LBB case to the steamline of advanced WWER 1000 reactor.« less

Predicting Large Deflections of Multiplate Fuel Elements Using a Monolithic FSI Approach

DOE PAGES

Curtis, Franklin G.; Freels, James D.; Ekici, Kivanc

2017-10-26

As part of the Global Threat Reduction Initiative, the Oak Ridge National Laboratory is evaluating conversion of fuel for the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium. Currently, multiphysics simulations that model fluid-structure interaction phenomena are being performed to ensure the safety of the reactor with the new fuel type. A monolithic solver that fully couples fluid and structural dynamics is used to model deflections in the new design. A classical experiment is chosen to validate the capabilities of the current solver and the method. Here, a single-plate simulation with various boundary conditions as well asmore » a five-plate simulation are presented. Finally, use of the monolithic solver provides stable solutions for the large deflections and the tight coupling of the fluid and structure and the maximum deflections are captured accurately.« less

Dedicated nuclear facilities for electrolytic hydrogen production

NASA Technical Reports Server (NTRS)

Foh, S. E.; Escher, W. J. D.; Donakowski, T. D.

1979-01-01

An advanced technology, fully dedicated nuclear-electrolytic hydrogen production facility is presented. This plant will produce hydrogen and oxygen only and no electrical power will be generated for off-plant use. The conceptual design was based on hydrogen production to fill a pipeline at 1000 psi and a 3000 MW nuclear base, and the base-line facility nuclear-to-shaftpower and shaftpower-to-electricity subsystems, the water treatment subsystem, electricity-to-hydrogen subsystem, hydrogen compression, efficiency, and hydrogen production cost are discussed. The final conceptual design integrates a 3000 MWth high-temperature gas-cooled reactor operating at 980 C helium reactor-out temperature, direct dc electricity generation via acyclic generators, and high-current density, high-pressure electrolyzers based on the solid polymer electrolyte approach. All subsystems are close-coupled and optimally interfaced and pipeline hydrogen is produced at 1000 psi. Hydrogen costs were about half of the conventional nuclear electrolysis process.

Biofilm thickness measurement using an ultrasound method in a liquid phase.

PubMed

Maurício, R; Dias, C J; Jubilado, N; Santana, F

2013-10-01

In this report, the development of an online, noninvasive, measurement method of the biofilm thickness in a liquid phase is presented. The method is based in the analysis of the ultrasound wave pulse-echo behavior in a liquid phase reproducing the real reactor conditions. It does not imply the removal of the biomass from the support or any kind of intervention in the support (pipes) to detect and perform the measurements (non-invasiveness). The developed method allows for its sensor to be easily and quickly mounted and unmounted in any location along a pipe or reactor wall. Finally, this method is an important innovation because it allows the thickness measurement of a biofilm, in liquid phase conditions that can be used in monitoring programs, to help in scheduling cleaning actions to remove the unwanted biofilm, in several application areas, namely in potable water supply pipes.

Radiation Resistant Electrical Insulation Materials for Nuclear Reactors: Final Report

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

Duckworth, Robert C.; Aytug, Tolga; Paranthaman, M. Parans

The instrument and control cables in future nuclear reactors will be exposed to temperatures, dose rates, and accumulated doses exceeding those originally anticipated for the 40-year operational life of the nuclear power plant fleet. The use of nanocomposite dielectrics as insulating material for such cables has been considered a route to performance improvement. In this project, nanoparticles were developed and successfully included in three separate material systems [cross-linked polyvinyl alcohol (PVA/XLPVA), cross-linked polyethylene (PE/XLPE), and polyimide (PI)], and the chemical, electrical, and mechanical performance of each was analyzed as a function of environmental exposure and composition. Improvements were found inmore » each material system; however, refinement of each processing pathway is needed, and the consequences of these refinements in the context of thermal, radiation, and moisture exposures should be evaluated before transferring knowledge to industry.« less