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Sample records for light-water production reactor

  1. Light water reactor program

    SciTech Connect

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  2. Light water reactor health physics.

    PubMed

    Prince, Robert J; Bradley, Scott E

    2005-06-01

    In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence.

  3. Light water reactor health physics.

    PubMed

    Prince, Robert J; Bradley, Scott E

    2004-11-01

    In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence.

  4. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, William E.; Trapp, Turner J.

    1985-01-01

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  5. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, W.E.; Trapp, T.J.

    1983-06-10

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  6. Radioactive Fission Product Release from Defective Light Water Reactor Fuel Elements

    SciTech Connect

    Konyashov, Vadim V.; Krasnov, Alexander M.

    2002-04-15

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

  7. LIGHT WATER MODERATED NEUTRONIC REACTOR

    DOEpatents

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  8. Light-Water Breeder Reactor

    DOEpatents

    Beaudoin, B. R.; Cohen, J. D.; Jones, D. H.; Marier, Jr, L. J.; Raab, H. F.

    1972-06-20

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  9. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    SciTech Connect

    Mac Donald, Philip Elsworth

    2002-06-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  10. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    SciTech Connect

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  11. Assessment of light water reactor accident management programs and experience

    SciTech Connect

    Hammersley, R.J.

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  12. Commercial Light Water Reactor Tritium Extraction Facility

    SciTech Connect

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  13. Light water reactor lower head failure analysis

    SciTech Connect

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  14. Environmentally assisted cracking in light water reactors

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  15. 60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

    NASA Astrophysics Data System (ADS)

    van Duysen, J. C.; Meric de Bellefon, G.

    2017-02-01

    The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

  16. Fabrication of light water reactor tritium targets

    SciTech Connect

    Pilger, J.P.

    1991-11-01

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

  17. Environmentally assisted cracking in light water reactors.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

    2007-11-06

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the current choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 10{sup 21} n x cm{sup -2}. The crack growth rates (CGRs) of the irradiated steels are a factor of {approx}5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to {approx}3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain >0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature

  18. Light Water Reactor Sustainability Accomplishments Report

    SciTech Connect

    McCarthy, Kathryn A.

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  19. Materials Inventory Database for the Light Water Reactor Sustainability Program

    SciTech Connect

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  20. Towards intrinsically safe light-water reactors

    SciTech Connect

    Hannerz, K

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification.

  1. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Progress Report for Work Through September 2003, 2nd Annual/8th Quarterly Report

    SciTech Connect

    Philip E. MacDonald

    2003-09-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world.

  2. Environmentally assisted cracking in light water reactors

    SciTech Connect

    Shack, W.J.; Kassner, T.F.; Maiya, P.S.; Park, J.Y.; Ruther, W.E.

    1988-10-01

    Research during the past year focused on (1) stress corrosion cracking (SCC) of austentitic stainless steels (SS), (2) fatigue of Type 316NG SS, and (3) SCC of ferritic steels used in reactor piping, pressure vessels, and steam generators. Stress corrosion cracking studies on austentitic SS explored the critical strains required for crack initiation, the effects of crevice conditions on SCC susceptibility, heat-to-heat variations in SCC susceptibility of Type 316NG and modified Type 347 SS, the effect of heat treatment on the susceptibility of Type 347 SS, threshold stress intensity values for crack growth in Type 316NG SS, and the effects of cuprous ion and several organic salts on the SCC of sensitized Type 304 SS. Crevice conditions were observed to strongly promote SCC. Significant heat-to-heat variations were observed in SCC susceptibility of Types 316NG and 347 SS. No correlation was found between SCC behavior and minor variations in chemical composition. A significant effect of heat treatment was observed in Type 347 SS. A heat that was extremely resistant to SCC after heat treatment at 650/degree/C for 24 h was susceptible to transgranular stress corrosion cracking (TGSCC) in the solution-annealed condition. Although there was no sensitization in either condition, the presence or absence of precipitates and differences in precipitate morphology appear to influence the SCC behavior. 20 refs., 20 figs., 11 tabs.

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

    NASA Astrophysics Data System (ADS)

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

    2015-04-01

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

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

    SciTech Connect

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

    2015-04-29

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

  5. Revised accident source terms for light-water reactors

    SciTech Connect

    Soffer, L.

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  6. Light-water breeder reactors: preliminary safety and environmental information document. Volume III

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter.

  7. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report

    SciTech Connect

    Mac Donald, Philip Elsworth

    2002-09-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

  8. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    SciTech Connect

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

    2010-11-01

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

  9. Mechanical design of a light water breeder reactor

    DOEpatents

    Fauth, Jr., William L.; Jones, Daniel S.; Kolsun, George J.; Erbes, John G.; Brennan, John J.; Weissburg, James A.; Sharbaugh, John E.

    1976-01-01

    In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

  10. Multi-Applications Small Light Water Reactor - NERI Final Report

    SciTech Connect

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

  11. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    SciTech Connect

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  12. 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.

  13. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  14. The Consortium for Advanced Simulation of Light Water Reactors

    SciTech Connect

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  15. Final Report on Isotope Ratio Techniques for Light Water Reactors

    SciTech Connect

    Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Mitchell, Mark R.; Meriwether, George H.; Reid, Bruce D.

    2009-07-01

    The Isotope Ratio Method (IRM) is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods.

  16. Fatigue and environmentally assisted cracking in light water reactors

    SciTech Connect

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

    1991-12-01

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas.

  17. Isotopic signature of atmospheric xenon released from light water reactors.

    PubMed

    Kalinowski, Martin B; Pistner, Christoph

    2006-01-01

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

  18. Non-linear analysis in Light Water Reactor design

    SciTech Connect

    Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.

    1980-03-01

    The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.

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

    SciTech Connect

    Hellesen, C.; Grape, S.; Haakanson, A.; Jacobson Svaerd, S.; Jansson, P.

    2013-07-01

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

  20. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    SciTech Connect

    Lewis, M R

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  1. Light-water breeder reactor (LWBR Development Program)

    DOEpatents

    Beaudoin, B.R.; Cohen, J.D.; Jones, D.H.; Marier, L.J. Jr.; Raab, H.F.

    1972-06-20

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

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

    SciTech Connect

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

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

  3. Advanced Light Water Reactor Program: Program management and staff review methodology

    SciTech Connect

    Moran, D.H.

    1986-12-01

    This report summarizes the NRC/EPRI coordinated effort to develop design requirements for a standardized advanced light water reactor (ALWR) and the procedures for screening and applying new generic safety issues to this program. The end-product will be an NRC-approved ALWR Requirements Document for use by the nuclear industry in generating designs of LWRs to be constructed for operation in the 1990s and beyond.

  4. Technologies for Upgrading Light Water Reactor Outlet Temperature

    SciTech Connect

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

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

  5. Thermal-Hydraulic Analysis of Supercritical Pressure Light Water Reactors

    SciTech Connect

    Cheng, X.; Schulenberg, T.; Koshizuka, S.; Oka, Y.; Souyri, A.

    2002-07-01

    In the frame of the European project HPLWR, joined by European research institutions, industrial partners and the University of Tokyo, thermal-hydraulic analysis of supercritical pressure light water reactors has been carried out. A thorough literature survey on heat transfer of supercritical fluids indicates a large deficiency in the prediction of the heat transfer coefficient and the onset of heat transfer deterioration under the reactor condition. A CFD code for analysing the thermal-hydraulic behaviour of supercritical fluids was developed. Numerical results show that the heat transfer coefficient, including the heat transfer deterioration region, can be well predicted using this CFD code, at least for circular tube geometries. Such a CFD code is well suitable for understanding the heat transfer mechanism. Based on the numerical results, a new heat transfer correlation has been proposed. For the thermal-hydraulic design of an HPLWR fuel assembly, the subchannel analysis code STAR-SC has been developed with a high numerical efficiency and a high applicability to different kinds of fuel assembly configurations. The results show clearly that design of a HPLWR fuel assembly is a highly challenging task. At the same time, sub-channel analysis provides some important guidelines for the design of a HPLWR fuel assembly. (authors)

  6. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    McCarthy, Kathryn A.; Busby, Jeremy; Hallbert, Bruce; Bragg-Sitton, Shannon; Smith, Curtis; Barnard, Cathy

    2014-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  7. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  8. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  9. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    SciTech Connect

    Snead, Lance Lewis; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Robb, Kevin R.; Snead, Mary A.

    2015-04-01

    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  10. Impact of inflow transport approximation on light water reactor analysis

    NASA Astrophysics Data System (ADS)

    Choi, Sooyoung; Smith, Kord; Lee, Hyun Chul; Lee, Deokjung

    2015-10-01

    The impact of the inflow transport approximation on light water reactor analysis is investigated, and it is verified that the inflow transport approximation significantly improves the accuracy of the transport and transport/diffusion solutions. A methodology for an inflow transport approximation is implemented in order to generate an accurate transport cross section. The inflow transport approximation is compared to the conventional methods, which are the consistent-PN and the outflow transport approximations. The three transport approximations are implemented in the lattice physics code STREAM, and verification is performed for various verification problems in order to investigate their effects and accuracy. From the verification, it is noted that the consistent-PN and the outflow transport approximations cause significant error in calculating the eigenvalue and the power distribution. The inflow transport approximation shows very accurate and precise results for the verification problems. The inflow transport approximation shows significant improvements not only for the high leakage problem but also for practical large core problem analyses.

  11. Thorium-Based Transmuter Fuels for Light Water Reactors

    SciTech Connect

    J. Stephen Herring; P. E. MacDonald; K. Weaver

    2004-04-01

    A light water reactor (LWR) fuel cycle is proposed where the reactor core mainly consists of standard uranium-dioxide (UO2) fuel rods with typical 235U enrichment, along with thoria-urania (ThO2-UO2) or yttria-stablized zirconia fertile-free fuel rods containing the plutonium and minor actinides typical of 30-yr old UO2 fuel in 1/9 to 1/3 of the positions. The goals of this mono-recycling strategy or "twice through fuel cycle" are to transmute the great majority of the long lived actinides in existing LWRs and to discharge a fuel form that is a very robust waste form and whose isotopic content is very proliferation resistant. The incorporation of plutonium into a ThO2 or yttria-stablized zirconia fertile-free matrix results in the consumption of already-separated plutonium without breeding significant additional 239Pu. The minor actinides (i.e., neptunium, americium, curium, berkelium, californium, etc.) are also included in the ThO2 or fertile-free transmuter fuel rods to further reduce the overall long-term radiotoxicity of the fuel cycle. Our analyses have shown that thorium-based or fertile-free fuels can reduce the amount of 239Pu needing further transmutation or going to a repository by ~90%. Also, thorium-based fuels produce a mixture of plutonium isotopes high in 238Pu. Because of the high decay heat and spontaneous neutron generation of 238Pu, this isotope provides intrinsic proliferation resistance.

  12. Thorium-Based Transmuter Fuels for Light Water Reactors

    SciTech Connect

    Herring, J. Stephen; MacDonald, Philip E.; Weaver, Kevan D.

    2004-07-15

    A light water reactor (LWR) fuel cycle is proposed where the reactor core mainly consists of standard uranium-dioxide (UO{sub 2}) fuel rods with typical {sup 235}U enrichment, along with thoria-urania (ThO{sub 2}-UO{sub 2}) or yttria-stablized zirconia fertile-free fuel rods containing the plutonium and minor actinides typical of 30-yr old UO{sub 2} fuel in 1/9 to 1/3 of the positions. The goals of this mono-recycling strategy or 'twice through fuel cycle' are to transmute the great majority of the long lived actinides in existing LWRs and to discharge a fuel form that is a very robust waste form and whose isotopic content is very proliferation resistant. The incorporation of plutonium into a ThO{sub 2} or yttria-stablized zirconia fertile-free matrix results in the consumption of already-separated plutonium without breeding significant additional {sup 239}Pu. The minor actinides (i.e., neptunium, americium, curium, berkelium, californium, etc.) are also included in the ThO{sub 2} or fertile-free transmuter fuel rods to further reduce the overall long-term radiotoxicity of the fuel cycle. Our analyses have shown that thorium-based or fertile-free fuels can reduce the amount of {sup 239}Pu needing further transmutation or going to a repository by {approx}90%. Also, thorium-based fuels produce a mixture of plutonium isotopes high in {sup 238}Pu. Because of the high decay heat and spontaneous neutron generation of {sup 238}Pu, this isotope provides intrinsic proliferation resistance.

  13. Qualification of a Tritium-Producing Target for the light water reactor application

    SciTech Connect

    Apley, W.J.; Beeman, G.H.; Ethridge, J.L.

    1992-06-01

    The Pacific Northwest Laboratory (PNL) currently manages the Light Water Reactor (LWR) Tritium Target Development Project (TTDP) for the Office of New Production Reactors (NP), US Department of Energy. The project`s objective is to demonstrate and qualify a high-temperature LWR tritium target system with fabrication and extraction processes sufficiently confirmed to ensure a deployable system consistent with variable tritium production demands. The project has also examined and reported on technical and institutional issues associated with acquisition and conversion of the yet uncompleted Washington Public Power Supply System Unit I (WNP-1) for tritium production purposes. WNP-1 is a 63% complete, Babcock and Wilcox, 3800 MW thermal, 205 assembly, pressurized water reactor located at Hanford, Washington. WNP-1 has served as the reference LWR plant for the technical evaluation and target development activities. This report discussed the evaluation and development necessary to provide a complete LWR target qualification package.

  14. Multi-Application Small Light Water Reactor Final Report

    SciTech Connect

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as

  15. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-12

    ... Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory Commission... Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES: Submit comments by November...

  16. 78 FR 41436 - Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-10

    ... COMMISSION Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The NRC seeks public...- Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' This area includes a revised...

  17. Characterization of 14C in Swedish light water reactors.

    PubMed

    Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

    2008-08-01

    This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

  18. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    SciTech Connect

    Thomas, Kenneth; Oxstrand, Johanna

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  19. Burnup credit validation of SCALE-4 using light-water-reactor criticals

    SciTech Connect

    Bowman, S.M.; Hermann, O.W.; Brady, M.C.

    1993-03-01

    The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison with LWR core criticals. These benchmarks are relevant because they test a methodology`s ability to predict spent fuel isotopic and to evaluate the reactivity effects of heterogeneities and strong absorbers. Data are available to perform analyses at precise state points. As part of the Burnup Credit Analysis Verification (BCAV) Task, the U.S. Department of Energy has sponsored analysis of selected reactor core critical configurations from commercial pressurized-water-reactors (PWRs). The initial analysis methodology used the SCALE-4 code system to analyze a set of reactor critical configurations from Virginia Power`s Slurry and North Anna reactors. However, the analysis procedure was complex and included the calculation of lumped fission products. The methodology has since been revised to simplify both the data requirements and the calculational procedure for the criticality analyst. This revised methodology is validated here by a comparison with three reactor critical configurations from Tennessee Valley Authority`s Sequoyah Unit 2 Cycle 3 and two from Virginia Power`s Slurry Unit 1 Cycle 2.

  20. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    SciTech Connect

    R. Johansen

    2011-09-01

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  1. Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

    SciTech Connect

    R. M. Ferrer; S. Bays; M. Pope

    2008-04-01

    The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

  2. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  3. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which...

  4. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which...

  5. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which...

  6. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which...

  7. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which...

  8. GENERIC, COMPONENT FAILURE DATA BASE FOR LIGHT WATER AND LIQUID SODIUM REACTOR PRAs

    SciTech Connect

    S. A. Eide; S. V. Chmielewski; T. D. Swantz

    1990-02-01

    A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

  9. Grouping of light water reactors for evaluation of decay heat removal capability

    SciTech Connect

    Karol, R.; Fresco, A.; Perkins, K.R.

    1984-06-01

    This grouping report provides a compilation of decay heat removal systems (DHRS) data for operating commercial light water reactors. The reactors have been divided into 12 groups based on similarity of the DHRS and related systems as part of the NRC Task Action Plan on Shutdown Decay Heat Removal Requirements.

  10. REACTOR PRESSURE VESSEL ISSUES FOR THE LIGHT-WATER REACTOR SUSTAINABILITY PROGRAM

    SciTech Connect

    Nanstad, Randy K; Odette, George Robert

    2010-01-01

    The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable operation to at least 80 years. The reactor pressure vessel (RPV) is one of the primary components requiring significant research to enable such long-term operation. There are significant issues that need to be addressed to reduce the uncertainties in regulatory application, such as, 1) high neutron fluence/long irradiation times, and flux effects, 2) material variability, 3) high-nickel materials, 4)specimen size effects and the fracture toughness master curve, etc. The first issue is the highest priority to obtain the data and mechanistic understanding to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs and the LWRSP plans to address those issues.

  11. Materials Degradation in Light Water Reactors: Life After 60,???

    SciTech Connect

    Busby, Jeremy T; Nanstad, Randy K; Stoller, Roger E; Feng, Zhili; Naus, Dan J

    2008-04-01

    Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase

  12. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  13. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  14. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    SciTech Connect

    Pan, Paul Y

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  15. Evolutionary/advanced light water reactor data report

    SciTech Connect

    1996-02-09

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  16. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  17. Fatigue and environmentally assisted cracking in light water reactors

    SciTech Connect

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

    1992-03-01

    Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B steel were determined under small-amplitude cyclic loading in HP water with {approx}300 pbb dissolved oxygen. Radiation-induced segregation and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence also have been investigated. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain-rate tensile tests were conducted on tubular specimens in air and in simulated BWR water at 289{degrees}C.

  18. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    SciTech Connect

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

    2015-09-01

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

  19. Reducing the cobalt inventory in light water reactors

    SciTech Connect

    Ocken, H.

    1985-01-01

    Reducing the cobalt content of materials used in nuclear power plants is one approach to controlling the radiation fields responsible for occupational radiation exposure; corrosion of steam generator tubing is the primary source in pressurized water reactors (PWRs). Wear of the cobalt-base alloys used to hardface valves (especially feedwater regulator valves) and as pins and rollers in control blades are the primary boiling water reactor (BWR) sources. Routine valve maintenance can also be a significant source of cobalt. Wear, mechanical property, and corrosion measurements led to the selection of Nitronic-60/CFA and PH 13-8 Mo/Inconel X-750 as low-cobalt alloys for use as pin/roller combinations. These alloys are currently being tested in two commercial BWRs. Measurements show that Type 440C stainless steel wears less than the cobalt-base alloys in BWR feedwater regulator valves. Sliding wear tests performed at room temperature in simulated PWR water showed that Colmonoy 74 and 84, Deloro 40, and Vertx 4776 are attractive low-cobalt hardfacing alloys if the applied loads are less than or equal to103 MPa. The cobalt-base alloys performed best at high loads (207 MPa). Ongoing laboratory studies address the development and evaluation of cobalt-free iron-base hardfacing alloys and seek to improve the wear resistance of cobalt-base alloys by using lasers. Reducing cobalt impurity levels in core components that are periodically discharged should also help reduce radiation fields and disposal costs.

  20. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    SciTech Connect

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  1. Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Blau, P. J.; Qu, J.; Lu, R.

    2016-11-01

    Fretting wear damage to fuel cladding from flow-induced vibrations can be a significant concern in the operation of light water nuclear reactors. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. The multi-stage model accounts for oxide layers and wear rate transitions. This paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

  2. Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

    SciTech Connect

    Blau, P. J.; Qu, J.; Lu, R.

    2016-09-21

    One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

  3. Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

    DOE PAGES

    Blau, P. J.; Qu, J.; Lu, R.

    2016-09-21

    One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes themore » basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.« less

  4. Radiological characteristics of light-water reactor spent fuel: A literature survey of experimental data. [82 references

    SciTech Connect

    Roddy, J.W.; Mailen, J.C.

    1987-12-01

    This survey brings together the experimentally determined light-water reactor spent fuel data comprising radionuclide composition, decay heat, and photon and neutron generation rates as identified in a literature survey. Many citations compare these data with values calculated using a radionuclide generation and depletion computer code, ORIGEN, and these comparisons have been included. ORIGEN is a widely recognized method for estimating the actinide, fission product, and activation product contents of irradiated reactor fuel, as well as the resulting heat generation and radiation levels. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of spent fuel. 82 refs., 4 figs., 17 tabs.

  5. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    SciTech Connect

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  6. Recent performance experience with US light water reactor self-actuating safety and relief valves

    SciTech Connect

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  7. Light Water Reactor Sustainability Program: Digital Architecture Project Plan

    SciTech Connect

    Thomas, Ken

    2014-09-01

    There are many technologies available to the nuclear power industry to improve efficiency in plant work activities. These range from new control room technologies to those for mobile field workers. They can make a positive impact on a wide range of performance objectives – increase in productivity, human error reduction, validation of results, accurate transfer of data, and elimination of repetitive tasks. It is expected that the industry will more and more turn to these technologies to achieve these operational efficiencies to lower costs. At the same time, this will help utilities manage a looming staffing problem as the inevitable retirement wave of the more seasoned workers affects both staffing levels and knowledge retention. A barrier to this wide-scale implementation of new technologies for operational efficiency is the lack of a comprehensive digital architecture that can support the real-time information exchanges needed to achieve the desired operational efficiencies. This project will define an advanced digital architecture that will accommodate the entire range of system, process, and plant worker activity to enable the highest degree of integration, thereby creating maximum efficiency and productivity. This pilot project will consider a range of open standards that are suitable for the various data and communication requirements of a seamless digital environment. It will map these standards into an overall architecture to support the II&C developments of this research program.

  8. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    SciTech Connect

    Holbrook, Mark; Kinsey, Jim

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  9. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL)

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Franceschini, Fausto; Evans, Thomas M.; Gehin, Jess C.

    2016-02-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a first-principles approach. In order to pursue these goals, CASL has participation from laboratory, academic, and industry partners. These partners are pursuing the solution of ten major "Challenge Problems" in order to advance the state-of-the-art in reactor design and analysis to permit power uprates, higher burnup, life extension, and increased safety. At present, the problems being addressed by CASL are primarily reactor physics-oriented; however, this paper is intended to introduce CASL to the reactor dosimetry community because of the importance of reactor physics modelling and nuclear data to define the source term for that community and the applicability and extensibility of the transport methods being developed.

  10. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    SciTech Connect

    K. A. McCarthy; D. L. Williams; R. Reister

    2012-05-01

    The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

  11. The burnup dependence of light water reactor spent fuel oxidation

    SciTech Connect

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies

  12. Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011

    SciTech Connect

    Not Listed

    2011-11-01

    The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

  13. International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

    NASA Astrophysics Data System (ADS)

    Geraskin, N. I.; Glebov, V. B.

    2017-01-01

    The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

  14. Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community

    SciTech Connect

    Nancy J. Lybeck; Magdy S. Tawfik; Binh T. Pham

    2011-08-01

    Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.

  15. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

    SciTech Connect

    Philip E. MacDonald

    2005-01-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was

  16. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    SciTech Connect

    Corradini, M. L.

    2015-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

  17. Survey of light-water-reactor designs to be offered in the United States

    SciTech Connect

    Spiewak, I.

    1986-03-01

    ORNL has conducted a Nuclear Power Options Viability Study for the Department of Energy. That study is primarily concerned with new technology which could be developed for initial operation in the 2000 to 2010 time frame. Such technology would have to compete not only with coal options but with incrementally improved commercial light-water-reactors. This survey reported here was undertaken to gain an understanding of the nuclear commercial technology likely to be offered in the late 1980s and perhaps beyond. The three US vendors actively marketing NSSSs are each developing a product for the future which they expect to be more reliable, more maintainable, more economical, and safer than the present plants. These are all essentially 3800-MW(t) designs, although all are studying smaller plants. They apparently will be off offered as standard prelicensed designs with much larger scope than earlier NSSS offerings, with the possibility of firm prices. Westinghouse with Mitsubishi Heavy Industries is developing a completely new design (APWR) to be built initially in Japan, hopefully for operation by the mid-1990s. Westinghouse is making a strong effort to have the APWR licensed in the US as a standard plant. Combustion Engineering (C-E) is evaluating potential improvements to the System-80 standard design (CESSAR) that has already received final design approval by the NRC. General Electric (GE), with Hitachi and Toshiba, is developing a new design (ABWR) that incorporates advanced features which have been proven by the worldwide BWR suppliers. The ABWR is to be built initially in Japan, but the design could be adapted to the United States. Westinghouse, C-E, and GE have done some conceptual evaluation of reactors in the 600-MW(e) class. The Westinghouse concept is a two-loop plant intended for factory assembly in a shipyard and delivery to a site by barge. The GE concept is a modification of the ABWR with some additional passive safety features. 16 figs.

  18. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    SciTech Connect

    Not Available

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  19. Burnup credit validation of SCALE-4 using light water reactor criticals

    SciTech Connect

    Bowman, S.M.; Hermann, O.W. ); Brady, M.C. )

    1993-01-01

    The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water-reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison to LWR core criticals. These are relevant benchmarks because they test a methodology's ability to predict spent fuel isotopics and to evaluate the reactivity effects of heterogeneities and strong absorbers. Data are available to perform analyses at precise state points. The US Department of Energy Burnup Credit Program has sponsored analysis of selected reactor core critical configurations from commercial pressurized-water-reactors (PWRs) in order to validate an appropriate analysis methodology. The initial methodology used the SCALE-4 code system to analyze a set of reactor critical configurations from Virginia Power's Surry and North Anna reactors. The methodology has since been revised to simplify both the data requirements and the calculational procedure for the criticality analyst. This revised methodology is validated here by comparison to three reactor critical configurations from Tennessee Valley Authority's Sequoyah Unit 2 Cycle 3 and two from Virginia Power's Surry Unit 1 Cycle 2.

  20. Neutron collar calibration for assay of LWR (light-water reactor) fuel assemblies

    SciTech Connect

    Menlove, H.O.; Pieper, J.E.

    1987-03-01

    The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the /sup 235/U content, and the /sup 238/U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities.

  1. Passive and inherent safety technologies for light-water nuclear reactors

    SciTech Connect

    Forsberg, C.W.

    1990-07-01

    Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs.

  2. Conceptual design of a pressure tube light water reactor with variable moderator control

    SciTech Connect

    Rachamin, R.; Fridman, E.; Galperin, A.

    2012-07-01

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  3. Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

    SciTech Connect

    Shannon M. Bragg-Sitton

    2013-09-01

    Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

  4. Utility leadership in reopening the nuclear option with advanced light water reactors

    SciTech Connect

    Marston, T.U.; Layman, W.H. )

    1992-01-01

    Since 1981, the Electric Power Research Institute (EPRI) has been pursing the development of the advanced light water reactor (ALWR). The ALWR Program is comprised of five phases and are described in the paper. In order to meet the anticipated baseline power generation requirements in the US, the Nuclear Power Oversight Committee (NPOC) has developed a strategic plan for ALWR implementation in order to regain the nuclear option in the United States. The paper also covers the policies behind the utility requirements, the status of ALWR developments in the United States, the electricity demands during the period 1990-2010, and some of the innovative features of the passive plants presently under design.

  5. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    SciTech Connect

    Blanchard, James; Butt, Darryl; Meyer, Mitchell; Xu, Peng

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  6. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    SciTech Connect

    Budd, W.A.

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  7. Design and optimization of a back-flow limiter for the high performance light water reactor

    SciTech Connect

    Fischer, Kai; Laurien, Eckart; Claas, Andreas G.; Schulenberg, Thomas

    2007-07-01

    Design and Analysis of a back-flow limiter are presented, which is implemented as a safety device in the four inlet lines of the Reactor Pressure Vessel (RPV) of the High Performance Light Water Reactor (HPLWR). As a passive component, the back-flow limiter has no moving parts and belongs to the group of fluid diodes. It has low flow resistance for regular operation condition and a high flow resistance when the flow direction is reversed which is the case if a break of the feedwater line occurs. The increased flow resistance is due to a substantially increased swirl for reverse flow condition. The design is optimized employing 1D flow analyses in combination with 3D CFD analyses with respect to geometrical modifications, like the nozzle shape and swirler angles. (authors)

  8. Thermal Properties of Structural Materials Found in Light Water Reactor Vessels

    SciTech Connect

    J. E. Daw; J. L. Rempe; D. L. Knudson

    2009-11-01

    High temperature material property data for structural materials used in existing Light Water Reactors (LWRs) are limited. Often, extrapolated values recommended in the literature differ significantly. To reduce such uncertainties, new data for SA533 Grade B, Class 1 (SA533B1) low alloy steel, Stainless Steel 304 (SS304), and Inconel 600, found in Light Water Reactor (LWR) vessels and penetrations, were acquired and tested using material property systems available at the High Temperature Test Laboratory (HTTL) at the Idaho National Laboratory (INL). Properties measured include thermal expansion, specific heat capacity, and thermal diffusivity for temperatures up to 1200 oC. From these results, thermal conductivity and density were calculated. Results show that, in some cases, previously recommended values for these material differ significantly from measured values at high temperatures. This is especially true for SA533B1, as previous data do not account for the phase transformation of this material between 740 oC and 840 oC.

  9. Nondestructive examination (NDE) reliability for inservice inspection of light waters reactors

    SciTech Connect

    Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T. )

    1989-11-01

    Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from April 1988 through September 1988. 33 refs., 70 figs., 12 tabs.

  10. Nondestructive examination (NDE) reliability for inservice inspection of light water reactors

    SciTech Connect

    Doctor, S.R.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V. )

    1992-05-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section 11 of the ASME Code. This is a progress report covering the programmatic work from October 1989 through September 1990.

  11. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    SciTech Connect

    Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R.; Carson, S.D.; Peterson, P.K.

    1997-11-30

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term.

  12. Expert assessments of the cost of light water small modular reactors.

    PubMed

    Abdulla, Ahmed; Azevedo, Inês Lima; Morgan, M Granger

    2013-06-11

    Analysts and decision makers frequently want estimates of the cost of technologies that have yet to be developed or deployed. Small modular reactors (SMRs), which could become part of a portfolio of carbon-free energy sources, are one such technology. Existing estimates of likely SMR costs rely on problematic top-down approaches or bottom-up assessments that are proprietary. When done properly, expert elicitations can complement these approaches. We developed detailed technical descriptions of two SMR designs and then conduced elicitation interviews in which we obtained probabilistic judgments from 16 experts who are involved in, or have access to, engineering-economic assessments of SMR projects. Here, we report estimates of the overnight cost and construction duration for five reactor-deployment scenarios that involve a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants vary by more than a factor of 2.5. Expert judgments about likely SMR costs display an even wider range. Median estimates for a 45 megawatts-electric (MWe) SMR range from $4,000 to $16,300/kWe and from $3,200 to $7,100/kWe for a 225-MWe SMR. Sources of disagreement are highlighted, exposing the thought processes of experts involved with SMR design. There was consensus that SMRs could be built and brought online about 2 y faster than large reactors. Experts identify more affordable unit cost, factory fabrication, and shorter construction schedules as factors that may make light water SMRs economically viable.

  13. Expert assessments of the cost of light water small modular reactors

    PubMed Central

    Abdulla, Ahmed; Azevedo, Inês Lima; Morgan, M. Granger

    2013-01-01

    Analysts and decision makers frequently want estimates of the cost of technologies that have yet to be developed or deployed. Small modular reactors (SMRs), which could become part of a portfolio of carbon-free energy sources, are one such technology. Existing estimates of likely SMR costs rely on problematic top-down approaches or bottom-up assessments that are proprietary. When done properly, expert elicitations can complement these approaches. We developed detailed technical descriptions of two SMR designs and then conduced elicitation interviews in which we obtained probabilistic judgments from 16 experts who are involved in, or have access to, engineering-economic assessments of SMR projects. Here, we report estimates of the overnight cost and construction duration for five reactor-deployment scenarios that involve a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants vary by more than a factor of 2.5. Expert judgments about likely SMR costs display an even wider range. Median estimates for a 45 megawatts-electric (MWe) SMR range from $4,000 to $16,300/kWe and from $3,200 to $7,100/kWe for a 225-MWe SMR. Sources of disagreement are highlighted, exposing the thought processes of experts involved with SMR design. There was consensus that SMRs could be built and brought online about 2 y faster than large reactors. Experts identify more affordable unit cost, factory fabrication, and shorter construction schedules as factors that may make light water SMRs economically viable. PMID:23716682

  14. A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building

    SciTech Connect

    Travis, J.R.; Wilson, T.L.; Spore, J.W.; Lam, K.L.; Rao, D.V.

    1994-09-01

    Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement over pressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region.

  15. 3D Simulation of Missing Pellet Surface Defects in Light Water Reactor Fuel Rods

    SciTech Connect

    B.W. Spencer; J.D. Hales; S.R. Novascone; R.L. Williamson

    2012-09-01

    The cladding on light water reactor (LWR) fuel rods provides a stable enclosure for fuel pellets and serves as a first barrier against fission product release. Consequently, it is important to design fuel to prevent cladding failure due to mechanical interactions with fuel pellets. Cladding stresses can be effectively limited by controlling power increase rates. However, it has been shown that local geometric irregularities caused by manufacturing defects known as missing pellet surfaces (MPS) in fuel pellets can lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. Nuclear fuel performance codes commonly use a 1.5D (axisymmetric, axially-stacked, one-dimensional radial) or 2D axisymmetric representation of the fuel rod. To study the effects of MPS defects, results from 1.5D or 2D fuel performance analyses are typically mapped to thermo-mechanical models that consist of a 2D plane-strain slice or a full 3D representation of the geometry of the pellet and clad in the region of the defect. The BISON fuel performance code developed at Idaho National Laboratory employs either a 2D axisymmetric or 3D representation of the full fuel rod. This allows for a computational model of the full fuel rod to include local defects. A 3D thermo-mechanical model is used to simulate the global fuel rod behavior, and includes effects on the thermal and mechanical behavior of the fuel due to accumulation of fission products, fission gas production and release, and the effects of fission gas accumulation on thermal conductivity across the fuel-clad gap. Local defects can be modeled simply by including them in the 3D fuel rod model, without the need for mapping between two separate models. This allows for the complete set of physics used in a fuel performance analysis to be included naturally in the computational representation of the local defect, and for the effects of the

  16. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

    1994-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289{degree}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  17. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J.

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  18. An innovative fuel design concept for improved light water reactor performance and safety. Final technical report

    SciTech Connect

    Tulenko, J.S.; Connell, R.G.

    1995-07-01

    Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. The purpose of this research was to explore a technique for extending fuel performance by thermally bonding LWR fuel with a non-alkaline liquid metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide (UO{sub 2}) fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Due to the thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high conductivity liquid metal thermally bonds the fuel to the cladding, and eliminates the large temperature change across the gap, while preserving the expansion and pellet loading capabilities. The resultant lower fuel temperature directly impacts fuel performance limit margins and also core transient performance. The application of liquid bonding techniques to LWR fuel was explored for the purposes of increasing LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) has been developed under the program to analyze the in-reactor performance of the liquid metal bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of Liquid Metal Bonded LWR fuel.

  19. Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1997-05-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288{degrees}C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs.

  20. Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J.

    1997-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.

  1. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    NASA Astrophysics Data System (ADS)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  2. Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

    SciTech Connect

    Thomas Baldwin; Magdy Tawfik; Leonard Bond

    2010-06-01

    In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R&D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10–12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I&C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy’s Light Water Reactor Sustainability Program. DOE has

  3. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    SciTech Connect

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  4. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  5. Storage of LWR (light-water-reactor) spent fuel in air

    SciTech Connect

    Thomas, L.E.; Charlot, L.A.; Coleman, J.E. ); Knoll, R.W. )

    1989-12-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to determine the oxidation response of light-water-reactor (LWR) spent fuels under conditions appropriate to fuel storage in air. The program is designed to investigate several independent variables that might affect the oxidation behavior of spent fuel. Included are temperature (135 to 230{degree}C), fuel burnup (to about 34 MWd/kgM), reactor type (pressurized and boiling water reactors), moisture level in the air, and the presence of a high gamma field. In continuing tests with declad spent fuel and nonirradiated UO{sub 2} specimens, oxidation rates were monitored by weight-gain measurements and the microstructures of subsamples taken during the weighing intervals were characterized by several analytical methods. The oxidation behavior indicated by weight gain and time to form powder will be reported in Volume III of this series. The characterization results obtained from x-ray diffractometry, transmission electron microscopy, scanning electron microscopy, and Auger electron spectrometry of oxidized fuel samples are presented in this report. 28 refs., 21 figs., 3 tabs.

  6. Primary disassembly of Light Water Breeder Reactor modules for core evaluation (LWBR Development Program)

    SciTech Connect

    Greenberger, R.J.; Miller, E.L.

    1987-10-01

    After successfully operating for 29,047 effective full power hours, the Light Water Breeder Reactor (LWBR) core was defueled prior to total decommissioning of the Shippingport Atomic Power Station. All nuclear fuel and much of the reactor internal hardware was removed from the reactor vessel. Non-fuel components were prepared for shipment to disposal sites, and the fuel assemblies were partially disassembled and shipped to the Expended Core Facility (ECF) in Idaho. At ECF, the fuel modules underwent further disassembly to provide fuel rods for nondestructive testing to establish the core's breeding efficiency and to provide core components for examinations to assess their performance characteristics. This report presents a basic description of the processes and equipment used to disassemble LWBR fuel modules for subsequent proof-of-breeding (POB) and core examination operations. Included are discussions of module handling fixtures and equipment, the underwater milling machine and bandsaw assemblies, and the associated design and operation of this equipment for LWBR fuel module disassembly.

  7. Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K.

    1996-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289{degrees}C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  8. Swelling in light water reactor internal components: Insights from computational modeling

    SciTech Connect

    Stoller, Roger E.; Barashev, Alexander V.; Golubov, Stanislav I.

    2015-08-01

    A modern cluster dynamics model has been used to investigate the materials and irradiation parameters that control microstructural evolution under the relatively low-temperature exposure conditions that are representative of the operating environment for in-core light water reactor components. The focus is on components fabricated from austenitic stainless steel. The model accounts for the synergistic interaction between radiation-produced vacancies and the helium that is produced by nuclear transmutation reactions. Cavity nucleation rates are shown to be relatively high in this temperature regime (275 to 325°C), but are sensitive to assumptions about the fine scale microstructure produced under low-temperature irradiation. The cavity nucleation rates observed run counter to the expectation that void swelling would not occur under these conditions. This expectation was based on previous research on void swelling in austenitic steels in fast reactors. This misleading impression arose primarily from an absence of relevant data. The results of the computational modeling are generally consistent with recent data obtained by examining ex-service components. However, it has been shown that the sensitivity of the model s predictions of low-temperature swelling behavior to assumptions about the primary damage source term and specification of the mean-field sink strengths is somewhat greater that that observed at higher temperatures. Further assessment of the mathematical model is underway to meet the long-term objective of this research, which is to provide a predictive model of void swelling at relevant lifetime exposures to support extended reactor operations.

  9. Special issue on the "Consortium for Advanced Simulation of Light Water Reactors Research and Development Progress"

    NASA Astrophysics Data System (ADS)

    Turinsky, Paul J.; Martin, William R.

    2017-04-01

    In this special issue of the Journal of Computational Physics, the research and development completed at the time of manuscript submission by the Consortium for Advanced Simulation of Light Water Reactors (CASL) is presented. CASL is the first of several Energy Innovation Hubs that have been created by the Department of Energy. The Hubs are modeled after the strong scientific management characteristics of the Manhattan Project and AT&T Bell Laboratories, and function as integrated research centers that combine basic and applied research with engineering to accelerate scientific discovery that addresses critical energy issues. Lifetime of a Hub is expected to be five or ten years depending upon performance, with CASL being granted a ten year lifetime.

  10. Categorization of failed and damaged spent LWR (light-water reactor) fuel currently in storage

    SciTech Connect

    Bailey, W.J.

    1987-11-01

    The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs.

  11. A flammability and combustion model for integrated accident analysis. [Advanced light water reactors

    SciTech Connect

    Plys, M.G.; Astleford, R.D.; Epstein, M. )

    1988-01-01

    A model for flammability characteristics and combustion of hydrogen and carbon monoxide mixtures is presented for application to severe accident analysis of Advanced Light Water Reactors (ALWR's). Flammability of general mixtures for thermodynamic conditions anticipated during a severe accident is quantified with a new correlation technique applied to data for several fuel and inertant mixtures and using accepted methods for combining these data. Combustion behavior is quantified by a mechanistic model consisting of a continuity and momentum balance for the burned gases, and considering an uncertainty parameter to match the idealized process to experiment. Benchmarks against experiment demonstrate the validity of this approach for a single recommended value of the flame flux multiplier parameter. The models presented here are equally applicable to analysis of current LWR's. 21 refs., 16 figs., 6 tabs.

  12. Overview of the U.S. DOE Light Water Reactor Sustainability Program

    SciTech Connect

    Shannong M. Bragg-Sitton; Jeremy T. Busby; Bruce P. Hallbert; Kathryn A. McCarthy; Richard Reister; Curtis L. Smith; Donald L. Williams

    2013-05-01

    The U.S. Department of Energy's Light Water Reactor Sustainability (LWRS) Program focuses on re­search and development to support the long-term operation of the nation's com­ mercial nuclear power plants. Extending the operation of current plants is essential to re­ alizing the administration's goals of reduc­ inggreenhouse gas emissions to 80 percent below 1990 levels by the year 2050. The science-based technical results from the LWRS Program provide data to help own­ ers make informed decisions on long-term operation and subsequent license renewal (the Nuclear Regulatory Commission's term for a second license renewal), thereby re­ ducing the uncertainty, and therefore the risk, associated with those decisions.

  13. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

    1995-03-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.

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

    SciTech Connect

    Bragg-Sitton, Shannon M.; Todosow, Michael; Montgomery, Robert; Stanek, Christopher R.; Carmack, W. Jon; Montgomery, Rose

    2016-07-11

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

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

    DOE PAGES

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

    2016-07-11

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

  16. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    SciTech Connect

    Rebak, Raul B.

    2014-09-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  17. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    SciTech Connect

    Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A; Matlack, Katie; Ramuhalli, Pradeep; Light, Glenn

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  18. Environmentally assisted cracking in Light Water Reactors. Volume 16: Semiannual report, October 1992--March 1993

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  19. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    SciTech Connect

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates

  20. Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Kassner, T. F.; Ruther, W. E.; Shack, W. J.; Smith, J. L.; Soppet, W. K.; Strain; R. V.

    1999-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments.

  1. Light-water-reactor safety research program. Quarterly progress report, January-March 1980

    SciTech Connect

    Massey, W.E.; Kyger, J.A.

    1980-08-01

    This progress report summarizes the Argonne National Laboratory work performed during January, February, and March 1980 on water-reactor-safety problems. The research and development area covered is Transient Fuel Response and Fission-Product Release.

  2. Analysis of assembly serial number usage in domestic light-water reactors

    SciTech Connect

    Reich, W.J. ); Moore, R.S. )

    1991-05-01

    Domestic light-water reactor (LWR) fuel assemblies are identified by a serial number that is placed on each assembly. These serial numbers are used as identifiers throughout the life of the fuel. The uniqueness of assembly serial numbers is important in determining their effectiveness as unambiguous identifiers. The purpose of this study is to determine what serial numbering schemes are used, the effectiveness of these schemes, and to quantify how many duplicate serial numbers occur on domestic LWR fuel assemblies. The serial numbering scheme adopted by the American National Standards Institute (ANSI) ensures uniqueness of assembly serial numbers. The latest numbering scheme adopted by General Electric (GE), was also found to be unique. Analysis of 70,971 fuel assembly serial numbers from permanently discharged fuel identified 11,948 serial number duplicates. Three duplicate serial numbers were found when analysis focused on duplication within the individual fuel inventory at each reactor site, but these were traced back to data entry errors and will be corrected by the Energy Information Administration (EIA). There were also three instances where the serial numbers used to identify assemblies used for hot cell studies differed from the serial numbers reported to the EIA. It is recommended that fuel fabricators and utilities adhere to the ANSI serial numbering scheme to ensure serial number uniqueness. In addition, organizations collecting serial number information, should request that all known serial numbers physically attached or associated with each assembly be reported and identified by the corresponding number scheme. 10 refs., 5 tabs.

  3. Computational Neutronics Methods and Transmutation Performance Analyses for Light Water Reactors

    SciTech Connect

    M. Asgari; B. Forget; S. Piet; R. Ferrer; S. Bays

    2007-03-01

    The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One obvious path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCm, PuNpAm, PuNp, and Pu.

  4. Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design

    SciTech Connect

    Galvin, Mark R.; Todreas, Neil E.; Conway, Larry E.

    2003-09-15

    New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented.

  5. Radionuclide characterization at US commercial light-water reactors for decommissioning assessment: Distributions, inventories, and waste disposal considerations

    SciTech Connect

    Abel, K.H.; Robertson, D.E.; Thomas, C.W.

    1992-09-01

    A continuing research program, conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission, characterizing radionuclide concentrations associated with US light-water reactors has been conducted for more than a decade. The research initially focused upon sampling and analytical measurements for the purpose of establishing radionuclide distributions and inventories for decommissioning assessment, since very little empirical data existed. The initial phase of the research program examined radionuclide concentrations and distributions external to the reactor vessel at seven US light water reactors. Later stages of the research program have examined the radionuclide distributions in the highly radioactive reactor internals and fuel assembly. Most recently, the research program is determining radionuclide concentrations in these highly radioactive components and comparing empirical results with those derived from the several nonempirical methodologies employed to estimate radionuclide inventories for disposal classification. The results of the research program to date are summarized, and their implications and significance for the decommissioning process are noted.

  6. Environmentally assisted cracking in light water reactors annual report January - December 2005.

    SciTech Connect

    Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

    2007-08-31

    This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data

  7. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-08-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  8. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-01-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  9. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    SciTech Connect

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  10. Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

    SciTech Connect

    Thomas, Ken; Lawrie, Sean; Hart, Adam; Vlahoplus, Chris

    2014-09-01

    The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on

  11. Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

    SciTech Connect

    Turinsky, Paul J.; Kothe, Douglas B.

    2016-05-15

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics “core simulator” based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M

  12. Environmentally assisted cracking in light water reactors : semiannual report, July 2000 - December 2000.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

    2002-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from July 2000 to December 2000. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. The fatigue strain-vs.-life data are summarized for the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Effects of the reactor coolant environment on the mechanism of fatigue crack initiation are discussed. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. A fracture toughness J-R curve test was conducted on a commercial heat of Type 304 SS that was irradiated to {approx}2.0 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. The results were compared with the data obtained earlier on steels irradiated to 0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) (0.45 and 1.35 dpa). Neutron irradiation at 288 C was found to decrease the fracture toughness of austenitic SSs. Tests were conducted on compact-tension specimens of Alloy 600 under cyclic loading to evaluate the enhancement of crack growth rates in LWR environments. Then, the existing fatigue crack growth data on Alloys 600 and 690 were analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range

  13. THERMIT-2: a two-fluid model for light water reactor subchannel transient analysis. Final report

    SciTech Connect

    Kelly, J.E.; Kao, S.P.; Kazimi, M.S.

    1981-04-01

    The broad effort of developing and assessing the two fluid model computer code THERMIT for light water reactor (LWR) subchannel analysis is described. The developmental effort required a reformulation of the coolant-to-fuel rod coupling so that THERMIT is now capable of traditional coolant-centered subchannel analysis. A model that accounts for mass, momentum and energy transport between mesh cells due to turbulent mixing for two-phase conditions has also been introduced. This model is the first such attempt in a two-fluid context. The liquid-vapor interfacial exchange terms in the two-fluid model have been modified for improved accuracy. A systematic evaluation of the exchange models has been performed. The mass and momentum exchange rates between the vapor and the liquid for pre-CHF conditions were evaluated by comparison to void fraction data in over 30 one-dimensional steady-state experiments reported in the open literature. The liquid-vapor energy exchange rate for post-CHF conditions was assessed using 15 steady-state, one-dimensional wall temperature measurements. The mixing model was tested against G.E. and Ispra BWR and PWR rod-bundle measurements. Comparisons with these measurements have shown the appropriateness of this model.

  14. Qualification Requirements of Guided Ultrasonic Waves for Inspection of Piping in Light Water Reactors

    SciTech Connect

    Meyer, Ryan M.; Ramuhalli, Pradeep; Doctor, Steven R.; Bond, Leonard J.

    2013-08-01

    Guided ultrasonic waves (GUW) are being increasingly used for both NDT and monitoring of piping. GUW offers advantages over many conventional NDE technologies due to the ability to inspect large volumes of piping components without significant removal of thermal insulation or protective layers. In addition, regions rendered inaccessible to more conventional NDE technologies may be more accessible using GUW techniques. For these reasons, utilities are increasingly considering the use of GUWs for performing the inspection of piping components in nuclear power plants. GUW is a rapidly evolving technology and its usage for inspection of nuclear power plant components requires refinement and qualification to ensure it is able to achieve consistent and acceptable levels of performance. This paper will discuss potential requirements for qualification of GUW techniques for the inspection of piping components in light water reactors (LWRs). The Nuclear Regulatory Commission has adopted ASME Boiler and Pressure Vessel Code requirements in Sections V, III, and XI for nondestructive examination methods, fabrication inspections, and pre-service and in-service inspections. A Section V working group has been formed to place the methodology of GUW into the ASME Boiler and Pressure Vessel Code but no requirements for technique, equipment, or personnel exist in the Code at this time.

  15. Passive ALWR requirements to prevent containment failure. [Advanced light water reactors

    SciTech Connect

    Additon, S.L.; Blanchard, D.P.; Leaver, D.E.; Persinko, D. TENERA, L.P., Bethesda, MD )

    1991-12-01

    The purpose of this report is to document a systematic evaluation of the Passive Advanced Light Water Reactor (ALWR) design requirements which address severe accident mitigation. This evaluation was performed concurrent with completion of the ALWR Requirements Document to assure the adequacy of these mitigation requirements. The passive plant approach to containment integrity assurance reflects an expansion of the approach established earlier for evolutionary ALWRs. The report identifies containment challenges that might occur coincident with or result from a core damage event, compiles the set of passive ALWR design requirements which addresses each challenge, and evaluates each set of requirements on an integrated basis to confirm that the requirements provide substantial assurance that coincident core damage and containment failure are precluded. Based on past PRAs, a review of pertinent safety functions, severe accident analyses, current regulatory requirements, and reviews by ALWR design personnel, twenty-three (23) potential containment challenges were identified. The report concludes that the relevant ALWR requirements severe to limit the likelihood and magnitude of the challenges, and to assure the capability of the containment to accommodate all challenges which remain potentially risk-significant.

  16. Microstructural examination of fatigue accumulation in critical LWR (light water reactor) components: Final report

    SciTech Connect

    Allen, A.J.; Buttle, D.J.; Coleman, C.F.; Smith, F.A.; Smith, R.L.

    1988-01-01

    This report describes a morphological study of the feasibility of measuring the fatigue damage accumulation state of critical light water reactor (LWR) components by microstructural examination. The changes in microstructure associated with fatigue processes are first discussed so that relevant NDE measurement parameters can be identified. (The creep regime is not considered in this report). The candidate NDE techniques are then reviewed in detail under the following headings: positron annihilation, x-ray diffraction, magnetic techniques, the magnetic Barkhausen effect, the magneto acoustic technique, acoustic emission, ultrasonic techniques and finally other miscellaneous techniques applicable to fatigue damage assessment. All the feasible techniques are summarised and rated in a set of comparison tables. A possible programme for the immediate development of the positron annihilation lineshape technique is proposed. It is concluded that the most successful method of measuring the fatigue accumulation in LWR critical components in a way which relates to the intent of the ASME pressure vessel codes, is likely to be the use of several techniques together and the cross-relation of the results obtained by each. Five techniques are highlighted for immediate possible development: 'etching and surface replication', 'positron annihilation lineshapes', 'x-ray diffraction residual stress', 'acoustic emission' and 'ultrasonic surface acoustic waves'.

  17. Insights for aging management of light water reactor components: Metal containments. Volume 5

    SciTech Connect

    Shah, V.N.; Sinha, U.P.; Smith, S.K.

    1994-03-01

    This report evaluates the available technical information and field experience related to management of aging damage to light water reactor metal containments. A generic aging management approach is suggested for the effective and comprehensive aging management of metal containments to ensure their safe operation. The major concern is corrosion of the embedded portion of the containment vessel and detection of this damage. The electromagnetic acoustic transducer and half-cell potential measurement are potential techniques to detect corrosion damage in the embedded portion of the containment vessel. Other corrosion-related concerns include inspection of corrosion damage on the inaccessible side of BWR Mark I and Mark II containment vessels and corrosion of the BWR Mark I torus and emergency core cooling system piping that penetrates the torus, and transgranular stress corrosion cracking of the penetration bellows. Fatigue-related concerns include reduction in the fatigue life (a) of a vessel caused by roughness of the corroded vessel surface and (b) of bellows because of any physical damage. Maintenance of surface coatings and sealant at the metal-concrete interface is the best protection against corrosion of the vessel.

  18. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    SciTech Connect

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  19. Development plan for the External Hazards Experimental Group. Light Water Reactor Sustainability Program

    SciTech Connect

    Coleman, Justin Leigh; Smith, Curtis Lee; Burns, Douglas Edward; Kammerer, Annie

    2016-03-01

    This report describes the development plan for a new multi-partner External Hazards Experimental Group (EHEG) coordinated by Idaho National Laboratory (INL) within the Risk-Informed Safety Margin Characterization (RISMC) technical pathway of the Light Water Reactor Sustainability Program. Currently, there is limited data available for development and validation of the tools and methods being developed in the RISMC Toolkit. The EHEG is being developed to obtain high-quality, small- and large-scale experimental data validation of RISMC tools and methods in a timely and cost-effective way. The group of universities and national laboratories that will eventually form the EHEG (which is ultimately expected to include both the initial participants and other universities and national laboratories that have been identified) have the expertise and experimental capabilities needed to both obtain and compile existing data archives and perform additional seismic and flooding experiments. The data developed by EHEG will be stored in databases for use within RISMC. These databases will be used to validate the advanced external hazard tools and methods.

  20. Light Water Reactor Sustainability Program A Reference Plan for Control Room Modernization: Planning and Analysis Phase

    SciTech Connect

    Jacques Hugo; Ronald Boring; Lew Hanes; Kenneth Thomas

    2013-09-01

    The U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) program is collaborating with a U.S. nuclear utility to bring about a systematic fleet-wide control room modernization. To facilitate this upgrade, a new distributed control system (DCS) is being introduced into the control rooms of these plants. The DCS will upgrade the legacy plant process computer and emergency response facility information system. In addition, the DCS will replace an existing analog turbine control system with a display-based system. With technology upgrades comes the opportunity to improve the overall human-system interaction between the operators and the control room. To optimize operator performance, the LWRS Control Room Modernization research team followed a human-centered approach published by the U.S. Nuclear Regulatory Commission. NUREG-0711, Rev. 3, Human Factors Engineering Program Review Model (O’Hara et al., 2012), prescribes four phases for human factors engineering. This report provides examples of the first phase, Planning and Analysis. The three elements of Planning and Analysis in NUREG-0711 that are most crucial to initiating control room upgrades are: • Operating Experience Review: Identifies opportunities for improvement in the existing system and provides lessons learned from implemented systems. • Function Analysis and Allocation: Identifies which functions at the plant may be optimally handled by the DCS vs. the operators. • Task Analysis: Identifies how tasks might be optimized for the operators. Each of these elements is covered in a separate chapter. Examples are drawn from workshops with reactor operators that were conducted at the LWRS Human System Simulation Laboratory HSSL and at the respective plants. The findings in this report represent generalized accounts of more detailed proprietary reports produced for the utility for each plant. The goal of this LWRS report is to disseminate the technique and provide examples sufficient to

  1. Weapons-grade plutonium dispositioning. Volume 4. Plutonium dispositioning in light water reactors

    SciTech Connect

    Sterbentz, J.W.; Olsen, C.S.; Sinha, U.P.

    1993-06-01

    This study is in response to a request by the Reactor Panel Subcommittee of the National Academy of Sciences (NAS) Committee on International Security and Arms Control (CISAC) to evaluate the feasibility of using plutonium fuels (without uranium) for disposal in existing conventional or advanced light water reactor (LWR) designs and in low temperature/pressure LWR designs that might be developed for plutonium disposal. Three plutonium-based fuel forms (oxides, aluminum metallics, and carbides) are evaluated for neutronic performance, fabrication technology, and material and compatibility issues. For the carbides, only the fabrication technologies are addressed. Viable plutonium oxide fuels for conventional or advanced LWRs include plutonium-zirconium-calcium oxide (PuO{sub 2}-ZrO{sub 2}-CaO) with the addition of thorium oxide (ThO{sub 2}) or a burnable poison such as erbium oxide (Er{sub 2}O{sub 3}) or europium oxide (Eu{sub 2}O{sub 3}) to achieve acceptable neutronic performance. Thorium will breed fissile uranium that may be unacceptable from a proliferation standpoint. Fabrication of uranium and mixed uranium-plutonium oxide fuels is well established; however, fabrication of plutonium-based oxide fuels will require further development. Viable aluminum-plutonium metallic fuels for a low temperature/pressure LWR include plutonium aluminide in an aluminum matrix (PuAl{sub 4}-Al) with the addition of a burnable poison such as erbium (Er) or europium (Eu). Fabrication of low-enriched plutonium in aluminum-plutonium metallic fuel rods was initially established 30 years ago and will require development to recapture and adapt the technology to meet current environmental and safety regulations. Fabrication of high-enriched uranium plate fuel by the picture-frame process is a well established process, but the use of plutonium would require the process to be upgraded in the United States to conform with current regulations and minimize the waste streams.

  2. Environmentally assisted cracking in light water reactors - annual report, January-December 2001.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E; Hiller, R. W.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

    2003-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level in water, and material heat treatment, on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments. Slow-strain-rate tensile tests and post-test fractographic analyses were conducted on several model SS alloys irradiated to {approx}2 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) ({approx}3 dpa) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at {approx}325 C to 5, 10, 20, and 40 dpa. Crack growth tests were completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600

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

    SciTech Connect

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

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

  4. Safe new reactor for radionuclide production

    SciTech Connect

    Gray, P.L.

    1995-02-15

    In late 1995, DOE is schedule to announce a new tritium production unit. Near the end of the last NPR (New Production Reactors) program, work was directed towards eliminating risks in current designs and reducing effects of accidents. In the Heavy Water Reactor Program at Savannah River, the coolant was changed from heavy to light water. An alternative, passively safe concept uses a heavy-water-filled, zircaloy reactor calandria near the bottom of a swimming pool; the calandria is supported on a light-water-coolant inlet plenum and has upflow through assemblies in the calandria tubes. The reactor concept eliminates or reduces significantly most design basis and severe accidents that plague other deigns. The proven, current SRS tritium cycle remains intact; production within the US of medical isotopes such as Mo-99 would also be possible.

  5. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Semiannual report, October 1990--March 1991: Volume 13

    SciTech Connect

    Doctor, S.R.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

    1992-07-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties.

  6. Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

    NASA Astrophysics Data System (ADS)

    Turinsky, Paul J.; Kothe, Douglas B.

    2016-05-01

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics ;core simulator; based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M

  7. Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013

    SciTech Connect

    Idaho National Laboratory

    2009-12-01

    Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement

  8. Aging of the containment pressure boundary in light-water reactor plants

    SciTech Connect

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

    1997-01-01

    Research is being conducted by the Oak Ridge National Laboratory to address aging of the containment pressure boundary in light-water reactor plants. The objectives of this work are to (1) identify the significant factors related to occurrence of corrosion, efficacy of inspection, and structural capacity reduction of steel containments and liners of concrete containments, and to make recommendations on use of risk models in regulatory decisions; (2) provide NRC reviewers a means of establishing current structural capacity margins for steel containments, and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by NRC reviewers in assessing the seriousness of reported incidences of containment degradation. In meeting these objectives research is being conducted in two primary task areas - pressure boundary condition assessment and root-cause resolution practices, and reliability-based condition assessments. Under the first task area a degradation assessment methodology was developed for use in characterizing the in-service condition of metal and concrete containment pressure boundary components and quantifying the amount of damage that is present. An assessment of available destructive and nondestructive techniques for examining steel containments and liners is ongoing. Under the second task area quantitative structural reliability analysis methods are being developed for application to degraded metallic pressure boundaries to provide assurances that they will be able to withstand future extreme loads during the desired service period with a level of reliability that is sufficient for public safety. To date, mathematical models that describe time-dependent changes in steel due to aggressive environmental factors have been identified, and statistical data supporting their use in time-dependent reliability analysis have been summarized.

  9. Conducting thermomechanical fatigue test in air at light water reactor relevant temperature intervals

    NASA Astrophysics Data System (ADS)

    Ramesh, Mageshwaran; Leber, Hans J.; Diener, Markus; Spolenak, Ralph

    2011-08-01

    In Light Water Reactors (LWR), many structural components are made of austenitic stainless steels (SS). These components are subject to extreme conditions, such as large temperature gradients and pressure loads during service. Hence, the fatigue and fracture behavior of austenitic SS under these conditions has evoked consistent interest over the years. Most studies dealing with this problem in the past, investigated the isothermal fatigue (IF) condition, which is not the case in the service, and less attention has been paid to thermomechanical fatigue (TMF). Moreover, the existing codes of practice and standards for TMF testing are mainly derived from the high temperature TMF tests ( T mean > 400 °C). This work presents the development of a facility to perform TMF tests under LWR relevant temperature interval in air. The realized testing parameters and tolerances are compared with the recommendations of existing codes of practice and standards from high temperature tests. The effectiveness of the testing facility was verified with series of TMF and IF tests performed on specimens made out of a commercial austenitic SS TP347 pipe material. The results revealed that the existing tolerances in standards are quite strict for the application of lower temperature ranges TMF tests. It was found that the synchronous, in-phase (IP) TMF tested specimens possess a higher lifetime than those subjected to the asynchronous, out-of-phase (OP) TMF and IF at T max in the investigated strain range for austenitic SS. Nevertheless, the fatigue lifetime of all the test conditions was similar in the engineering scale.

  10. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2013

    SciTech Connect

    Hallbert, Bruce; Thomas, Ken

    2014-09-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  11. Light Water Reactor Sustainability Program Grizzly Year-End Progress Report

    SciTech Connect

    Benjamin Spencer; Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner; Marie Backman; Brian Wirth; Stephen Novascone; Jason Hales

    2013-09-01

    The Grizzly software application is being developed under the Light Water Reactor Sustainability (LWRS) program to address aging and material degradation issues that could potentially become an obstacle to life extension of nuclear power plants beyond 60 years of operation. Grizzly is based on INL’s MOOSE multiphysics simulation environment, and can simultaneously solve a variety of tightly coupled physics equations, and is thus a very powerful and flexible tool with a wide range of potential applications. Grizzly, the development of which was begun during fiscal year (FY) 2012, is intended to address degradation in a variety of critical structures. The reactor pressure vessel (RPV) was chosen for an initial application of this software. Because it fulfills the critical roles of housing the reactor core and providing a barrier to the release of coolant, the RPV is clearly one of the most safety-critical components of a nuclear power plant. In addition, because of its cost, size and location in the plant, replacement of this component would be prohibitively expensive, so failure of the RPV to meet acceptance criteria would likely result in the shutting down of a nuclear power plant. The current practice used to perform engineering evaluations of the susceptibility of RPVs to fracture is to use the ASME Master Fracture Toughness Curve (ASME Code Case N-631 Section III). This is used in conjunction with empirically based models that describe the evolution of this curve due to embrittlement in terms of a transition temperature shift. These models are based on an extensive database of surveillance coupons that have been irradiated in operating nuclear power plants, but this data is limited to the lifetime of the current reactor fleet. This is an important limitation when considering life extension beyond 60 years. The currently available data cannot be extrapolated with confidence further out in time because there is a potential for additional damage mechanisms (i

  12. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    DOE PAGES

    Williams, M. L.; Wiarda, D.; Ilas, G.; ...

    2014-06-15

    Recently, we processed a new covariance data library based on ENDF/B-VII.1 for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. Moreover, the cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  13. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    NASA Astrophysics Data System (ADS)

    Williams, M. L.; Wiarda, D.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2015-01-01

    A new covariance data library based on ENDF/B-VII.1 was recently processed for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. The cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  14. How well does ORIGEN predict spent LWR (Light Water Reactor) fuel characteristics

    SciTech Connect

    Mailen, J.C.; Roddy, J.W.

    1987-01-01

    The ORIGEN computer code is widely used to estimate the radionuclide content (actinides, activation and fission products) of irradiated reactor fuel and the resultant heat generation and radiation levels associated with such fuel. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of the fuel. Calculated values determined using several variations of ORIGEN have been compared with experimentally determined values for actual fuel for many, but not all, of the parameters desired. In most cases, the comparisons did not use the most recent ORIGEN2 program, the most recent data libraries, or currently required quality assurance (QA) procedures. Comparisons of fuel composition data with ORIGEN2 are very limited, and the only data with proper QA are currently being acquired by Battelle Pacific Northwest Laboratory. This survey summarizes the fuel data available in the open literature and, where given, the calculated values by ORIGEN. Plans for additional analyses of well-characterized reactor fuel samples to improve the validation of ORIGEN2 are discussed.

  15. Development of high-fidelity multiphysics system for light water reactor analysis

    NASA Astrophysics Data System (ADS)

    Magedanz, Jeffrey W.

    There has been a tendency in recent years toward greater heterogeneity in reactor cores, due to the use of mixed-oxide (MOX) fuel, burnable absorbers, and longer cycles with consequently higher fuel burnup. The resulting asymmetry of the neutron flux and energy spectrum between regions with different compositions causes a need to account for the directional dependence of the neutron flux, instead of the traditional diffusion approximation. Furthermore, the presence of both MOX and high-burnup fuel in the core increases the complexity of the heat conduction. The heat transfer properties of the fuel pellet change with irradiation, and the thermal and mechanical expansion of the pellet and cladding strongly affect the size of the gap between them, and its consequent thermal resistance. These operational tendencies require higher fidelity multi-physics modeling capabilities, and this need is addressed by the developments performed within this PhD research. The dissertation describes the development of a High-Fidelity Multi-Physics System for Light Water Reactor Analysis. It consists of three coupled codes -- CTF for Thermal Hydraulics, TORT-TD for Neutron Kinetics, and FRAPTRAN for Fuel Performance. It is meant to address these modeling challenges in three ways: (1) by resolving the state of the system at the level of each fuel pin, rather than homogenizing entire fuel assemblies, (2) by using the multi-group Discrete Ordinates method to account for the directional dependence of the neutron flux, and (3) by using a fuel-performance code, rather than a Thermal Hydraulics code's simplified fuel model, to account for the material behavior of the fuel and its feedback to the hydraulic and neutronic behavior of the system. While the first two are improvements, the third, the use of a fuel-performance code for feedback, constitutes an innovation in this PhD project. Also important to this work is the manner in which such coupling is written. While coupling involves combining

  16. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    SciTech Connect

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' (Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety) is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document.

  17. Gigawatt-year nuclear-geothermal energy storage for light-water and high-temperature reactors

    SciTech Connect

    Forsberg, C. W.; Lee, Y.; Kulhanek, M.; Driscoll, M. J.

    2012-07-01

    Capital-intensive, low-operating cost nuclear plants are most economical when operated under base-load conditions. However, electricity demand varies on a daily, weekly, and seasonal basis. In deregulated utility markets this implies high prices for electricity at times of high electricity demand and low prices for electricity at times of low electricity demand. We examined coupling nuclear heat sources to geothermal heat storage systems to enable these power sources to meet hourly to seasonal variable electricity demand. At times of low electricity demand the reactor heats a fluid that is then injected a kilometer or more underground to heat rock to high temperatures. The fluid travels through the permeable-rock heat-storage zone, transfers heat to the rock, is returned to the surface to be reheated, and re-injected underground. At times of high electricity demand the cycle is reversed, heat is extracted, and the heat is used to power a geothermal power plant to produce intermediate or peak power. When coupling geothermal heat storage with light-water reactors (LWRs), pressurized water (<300 deg. C) is the preferred heat transfer fluid. When coupling geothermal heat storage with high temperature reactors at higher temperatures, supercritical carbon dioxide is the preferred heat transfer fluid. The non-ideal characteristics of supercritical carbon dioxide create the potential for efficient coupling with supercritical carbon dioxide power cycles. Underground rock cannot be insulated, thus small heat storage systems with high surface to volume ratios are not feasible because of excessive heat losses. The minimum heat storage capacity to enable seasonal storage is {approx}0.1 Gigawatt-year. Three technologies can create the required permeable rock: (1) hydro-fracture, (2) cave-block mining, and (3) selective rock dissolution. The economic assessments indicated a potentially competitive system for production of intermediate load electricity. The basis for a nuclear

  18. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect

    Not Available

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  19. Light water reactor fuel reprocessing: dissolution studies of voloxidized and nonvoloxidized fuel

    SciTech Connect

    Johnson, D.R.; Stone, J.A.

    1980-04-01

    Small-scale tests with irradiated Zircaloy-clad fuels from Robinson, Oconee, Saxton, and Point Beach reactors with burnups from about 200 to 28,000 MWD/MTHM have been made to determine the dissolution behavior of both voloxidized (U{sub 3}O{sub 8}) and nonvoloxidized (UO{sub 2}) fuel. No significant technical problems were encountered in batch-dissolving of either form. Dissolution rates were well-controlled in all tests. Significant characteristics of U{sub 3}O{sub 8} dissolution that differed from UO{sub 2} dissolution included: (1) reduced tritium and ruthenium ({sup 106}Ru) concentrations in product solutions, (2) increased insoluble noble metal fission product residue (about 2.2X greater), and (3) increased insoluble plutonium in the fission product residue. The insoluble plutonium is easily leached from the residue by 10M HNO{sub 3}. The weight of the fission product residue collected from both U{sub 3}O{sub 8} and UO{sub 2} fuels increased aproximately linearly with fuel burnup. A major fraction (>83%) of the {sup 85}Kr was evolved from U{sub 3}O{sub 8} fuel during dissolution rather than voloxidation. The {sup 85}Kr evolution rate was an appropriate monitor of fuel dissolution rate. Virtually all of the {sup 129}I was evolved by air sparging of the dissolver solution during dissolution. 30 tables, 18 figures.

  20. Environmentally assisted cracking in light-water reactors: Semi-annual report, January--June 1997. Volume 24

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1998-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in low-DO, simulated pressurized water reactor environments.

  1. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Semiannual report, April 1992--September 1992: Volume 16

    SciTech Connect

    Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Greenwood, M.S.; Heasler, P.G.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Vo, T.V.

    1993-11-01

    The Evaluation and Improvement of NDE Reliability for Inservice inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs);using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel and other components inspected in accordance with Section XI of the ASME Code. This is a programs report covering the programmatic work from April 1992 through September 1992.

  2. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Volume 14, Semiannual report, April 1991--September 1991

    SciTech Connect

    Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Good, M.S.; Greenwood, M.S.; Heasler, P.G.; Hockey, R.L.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

    1992-07-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWR`s); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from April 1991 through September 1991.

  3. Nondestructive examination (NDE) reliability for inservice inspection of light water reactors. Annual report, October 1989--September 1990: Volume 12

    SciTech Connect

    Doctor, S.R.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

    1992-05-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section 11 of the ASME Code. This is a progress report covering the programmatic work from October 1989 through September 1990.

  4. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Volume 15, Semiannual report: October 1991--March 1992

    SciTech Connect

    Doctor, S.R.; Diaz, A.A.; Friley, J.R.

    1993-09-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from October 1991 through March 1992.

  5. Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

    SciTech Connect

    Hursin, M.; Perret, G.

    2012-07-01

    The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

  6. Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

    SciTech Connect

    Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

    1991-10-01

    This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs.

  7. Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1995--September 30, 1996

    SciTech Connect

    1996-12-31

    The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1996 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2 and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems.

  8. Acoustic Emission and Guided Ultrasonic Waves for Detection and Continuous Monitoring of Cracks in Light Water Reactor Components

    SciTech Connect

    Meyer, Ryan M.; Coble, Jamie B.; Ramuhalli, Pradeep; Watson, Bruce E.; Cumblidge, Stephen E.; Doctor, Steven R.; Bond, Leonard J.

    2012-06-28

    Acoustic emission (AE) and guided ultrasonic waves (GUW) are considered for continuous monitoring and detection of cracks in Light Water Reactor (LWR) components. In this effort, both techniques are applied to the detection and monitoring of fatigue crack growth in a full scale pipe component. AE results indicated crack initiation and rapid growth in the pipe, and significant GUW responses were observed in response to the growth of the fatigue crack. After initiation, the crack growth was detectable with AE for approximately 20,000 cycles. Signals associated with initiation and rapid growth where distinguished based on total rate of activity and differences observed in the centroid frequency of hits. An intermediate stage between initiation and rapid growth was associated with significant energy emissions, though few hits. GUW exhibit a nearly monotonic trend with crack length with an exception of measurements obtained at 41 mm and 46 mm.

  9. Large-Scale Water-Vapor Two-Phase Flow Simulations in Advanced Light Water Reactor Cores

    SciTech Connect

    Hiroyuki, Yoshida; Kazuyuki, Takase; Hidesada, Tamai; Hajime, Akimoto; Yasuo, Ose

    2004-07-01

    Fluid flow characteristics in a fuel bundle of a reduced-moderation light water reactor (RMWR) with a tight-lattice core were analyzed numerically using a newly developed two-phase flow analysis code under the full bundle size condition. Conventional analysis methods such as subchannel codes need composition equations based on the experimental data. In case that there are no experimental data regarding to the thermal-hydraulics in the tight-lattice core, therefore, it is difficult to obtain high prediction accuracy on the thermal design of the RMWR. Then the large-scale direct numerical simulations with a super computer were chosen. The axial velocity distribution in a fuel bundle changed sharply around a spacer. Momentum transfer of vapor in a tight-lattice core is linear along the flow direction. The interface characteristics between water and vapor were clarified quantitatively. (authors)

  10. Protective structures on the surface of zirconium components of light water reactor cores: Formation, testing, and prototype equipment

    NASA Astrophysics Data System (ADS)

    Begrambekov, L. B.; Gordeev, A. A.; Evsin, A. E.; Ivanova, S. V.; Kaplevsky, A. S.; Sadovskiy, Ya. A.

    2015-12-01

    The results of tests of plasma treatment of zirconium and deposition of protective yttrium coatings used as the methods of protection of zirconium components of light water reactor cores against hydrogenation are detailed. The amount of hydrogen in the treated sample exposed to superheated steam for 2500 h at temperature T = 400°C and pressure p = 1 atm was five times lower than the corresponding value for the untreated one. The amount of hydrogen in the sample coated with yttrium remained almost unchanged in 4000 h of exposure. A plasma method for rapid testing for hydrogen resistance is proposed. The hydrogenation rate provided by this method is 700 times higher than that in tests with superheated steam. The results of preliminary experiments confirm the possibility of constructing a unit for batch processing of the surfaces of fuel rod claddings.

  11. An Estimate of the Cost of Electricity from Light Water Reactors and Fossil Plants with Carbon Capture and Sequestration

    SciTech Connect

    Simon, A J

    2009-08-21

    As envisioned in this report, LIFE technology lends itself to large, centralized, baseload (or 'always on') electrical generation. Should LIFE plants be built, they will have to compete in the electricity market with other generation technologies. We consider the economics of technologies with similar operating characteristics: significant economies of scale, limited capacity for turndown, zero dependence on intermittent resources and ability to meet environmental constraints. The five generation technologies examined here are: (1) Light Water Reactors (LWR); (2) Coal; (3) Coal with Carbon Capture and Sequestration (CCS); (4) Natural Gas; and (5) Natural Gas with Carbon Capture and Sequestration. We use MIT's cost estimation methodology (Du and Parsons, 2009) to determine the cost of electricity at which each of these technologies is viable.

  12. Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods

    SciTech Connect

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

  13. Protective structures on the surface of zirconium components of light water reactor cores: Formation, testing, and prototype equipment

    SciTech Connect

    Begrambekov, L. B.; Gordeev, A. A.; Evsin, A. E. Ivanova, S. V.; Kaplevsky, A. S.; Sadovskiy, Ya. A.

    2015-12-15

    The results of tests of plasma treatment of zirconium and deposition of protective yttrium coatings used as the methods of protection of zirconium components of light water reactor cores against hydrogenation are detailed. The amount of hydrogen in the treated sample exposed to superheated steam for 2500 h at temperature T = 400°C and pressure p = 1 atm was five times lower than the corresponding value for the untreated one. The amount of hydrogen in the sample coated with yttrium remained almost unchanged in 4000 h of exposure. A plasma method for rapid testing for hydrogen resistance is proposed. The hydrogenation rate provided by this method is 700 times higher than that in tests with superheated steam. The results of preliminary experiments confirm the possibility of constructing a unit for batch processing of the surfaces of fuel rod claddings.

  14. Light Water Reactor Sustainability Program Risk-Informed Safety Margins Characterization (RISMC) PathwayTechnical Program Plan

    SciTech Connect

    Curtis Smith; Cristian Rabiti; Richard Martineau

    2012-11-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”

  15. Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

    NASA Technical Reports Server (NTRS)

    Whitmarsh, C. L., Jr.

    1973-01-01

    An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

  16. Progress in understanding of direct containment heating phenomena in pressurized light water reactors

    SciTech Connect

    Ginsberg, T.; Tutu, N.K.

    1988-01-01

    Progress is described in development of a mechanistic understanding of direct containment heating phemonena arising during high-pressure melt ejection accidents in pressurized water reactor systems. The experimental data base is discussed which forms the basis for current assessments of containment pressure response using current lumped-parameter containment analysis methods. The deficiencies in available methods and supporting data base required to describe major phenomena occurring in the reactor cavity, intermediate subcompartments and containment dome are highlighted. Code calculation results presented in the literature are cited which demonstrate that the progress in understanding of DCH phenomena has also resulted in current predictions of containment pressure loadings which are significantly lower than are predicted by idealized, thermodynamic equilibrium calculations. Current methods are, nonetheless, still predicting containment-threatening loadings for large participating melt masses under high-pressure ejection conditions. Recommendations for future research are discussed. 36 refs., 5 figs., 1 tab.

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

    NASA Astrophysics Data System (ADS)

    Metzger, Kathryn E.

    Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it

  18. Environmentally assisted cracking in light water reactors. Semiannual report, April--September 1991: Volume 13

    SciTech Connect

    Kassner, T F; Ruther, W E; Chung, H M; Hicks, P D; Hins, A G; Park, J Y; Soppet, W K; Shack, W J

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with {approx} 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289{degrees}C.

  19. Nuclear Thermal Propulsion engine based on Particle Bed Reactor using light water steam as a propellant

    SciTech Connect

    Powell, J.R.; Ludewig, H.; Maise, G.

    1993-06-01

    In this paper the possibility of configuring a water cooled Nuclear Thermal Propulsion (NTP) rocket, based on a Particle Bed Reactor (PBR) is investigated. This rocket will be used to operate on water obtained from near earth objects. The conclusions reached in this paper indicate that it is possible to configure a PBR based NTP rocket to operate on water and meet the mission requirements envisioned for it. No insurmountable technology issues have been identified.

  20. Light-water reactors: preliminary safety and environmental information document. Volume I

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle (Pu/ThO/sub 2/ Burner).

  1. A new thermal hydraulics code coupled to agent for light water reactor analysis

    NASA Astrophysics Data System (ADS)

    Eklund, Matthew Deric

    A new numerical model for coupling a thermal hydraulics method based on the Drift Flux and Homogeneous Equilibrium Mixture (HEM) models, with a deterministic neutronics code system AGENT (Arbitrary Geometry Neutron Transport), is developed. Named the TH thermal hydraulics code, it is based on the mass continuity, momentum, and energy equations integrated with appropriate relations for liquid and vapor phasic velocities. The modified conservation equations are then evaluated in one-dimensional (1D) steady-state conditions for LWR coolant subchannel in the axial direction. This permits faster computation times without sacrificing significant accuracy, as compared to other three-dimensional (3D) codes such as RELAP5/TRACE. AGENT is a deterministic neutronics code system based on the Method of Characteristics to solve the 2D/3D neutron transport equation in current and future reactor systems. The coupling scheme between the TH and AGENT codes is accomplished by computing the normalized fission rate profile in the LWR fuel elements by AGENT. The normalized fission rate profile is then transferred to the TH thermal hydraulics code for computing the reactor coolant properties. In conjunction with the 1D axial TH code, a separate 1D radial heat transfer model within the TH code is used to determine the average fuel temperature at each node where coolant properties are calculated. These properties then are entered into Scale 6.1, a criticality analysis code, to recalculate fuel pin neutron interaction cross sections based on thermal feedback. With updated fuel neutron interaction cross sections, the fission rate profile is recalculated in AGENT, and the cycle continues until convergence is reached. The TH code and coupled AGENT-TH code are benchmarked against the TRACE reactor analysis software, showing required agreement in evaluating the basic reactor parameters.

  2. A MELCOR Application to Two Light Water Reactor Nuclear Power Plant Core Melt Scenarios with Assumed Cavity Flooding Action

    SciTech Connect

    Martin-Fuertes, Francisco; Martin-Valdepenas, Juan Manuel; Mira, Jose; Sanchez, Maria Jesus

    2003-10-15

    The MELCOR 1.8.4 code Bottom Head package has been applied to simulate two reactor cavity flooding scenarios for when the corium material relocates to the lower-plenum region in postulated severe accidents. The applications were preceded by a review of two main physical models, which highly impacted the results. A model comparison to available bibliography models was done, which allowed some code modifications on selected default assumptions to be undertaken. First, the corium convective heat transfer to the wall when it becomes liquid was modified, and second, the default nucleate boiling regime curve in a submerged hemisphere was replaced by a new curve (and, to a much lesser extent, the critical heat flux curve was slightly varied).The applications were devoted to two prototypical light water reactor nuclear power plants, a 2700-MW(thermal) pressurized water reactor (PWR) and a 1381-MW(thermal) boiling water reactor (BWR). The main conclusions of the cavity flooding simulations were that the PWR lower-head survivability is extended although it is clearly not guaranteed, while in the BWR sequence the corium seems to be successfully arrested in the lower plenum.Three applications of the CFX 4.4 computational fluid dynamics code were carried out in the context of the BWR scenario to support the first modification of the aforementioned two scenarios for MELCOR.Finally, in the same BWR context, a statistic predictor of selected output parameters as a function of input parameters is presented, which provides reasonable results when compared to MELCOR full calculations in much shorter CPU processing times.

  3. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    SciTech Connect

    Hagrman, D.T.; Allison, C.M.; Berna, G.A.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

  4. Spent nuclear fuel shipping cask handling capabilities of commercial light water reactors

    SciTech Connect

    Daling, P.M.; Konzek, G.J.; Lezberg, A.J.; Votaw, E.F.; Collingham, M.I.

    1985-04-01

    This report describes an evaluation of the cask handling capabilities of those reactors which are operating or under construction. A computerized data base that includes cask handling information was developed with information from the literature and utility-supplied data. The capability of each plant to receive and handle existing spent fuel shipping casks was then evaluated. Modal fractions were then calculated based on the results of these evaluations and the quantities of spent fuel projected to be generated by commercial nuclear power plants through 1998. The results indicated that all plants are capable of receiving and handling truck casks. Up to 118 out of 130 reactors (91%) could potentially handle the larger and heavier rail casks if the maximum capability of each facility is utilized. Design and analysis efforts and physical modifications to some plants would be needed to achieve this high rail percentage. These modifications would be needed to satisfy regulatory requirements, increase lifting capabilities, develop rail access, or improve other deficiencies. The remaining 12 reactors were determined to be capable of handling only the smaller truck casks. The percentage of plants that could receive and handle rail casks in the near-term would be reduced to 64%. The primary reason for a plant to be judged incapable of handling rail casks in the near-term was a lack of rail access. The remaining 36% of the plants would be limited to truck shipments. The modal fraction calculations indicated that up to 93% of the spent fuel accumulated by 1998 could be received at federal storage or disposal facilities via rail (based on each plant's maximum capabilities). If the near-term cask handling capabilities are considered, the rail percentage is reduced to 62%.

  5. A concept of JAERI passive safety light water reactor system (JPSR)

    SciTech Connect

    Murao, Y.; Araya, F.; Iwamura, T.

    1995-09-01

    The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

  6. Fatigue crack initiation in carbon and low-alloy steels in light water reactor environments : mechanism and prediction.

    SciTech Connect

    Chopra, O. K.; Shack, W. J.

    1998-01-27

    Section 111 of the ASME Boiler and Pressure Vessel Code specifies fatigue design curves for structural materials. The effects of reactor coolant environments are not explicitly addressed by the Code design curves. Recent test data illustrate potentially significant effects of light water reactor (LWR) coolant environments on the fatigue resistance of carbon and low-alloy steels. Under certain loading and environmental conditions, fatigue lives of test specimens may be shorter than those in air by a factor of {approx}70. The crack initiation and crack growth characteristics of carbon and low-alloy steels in LWR environments are presented. Decreases in fatigue life of these steels in high-dissolved-oxygen water are caused primarily by the effect of environment on growth of short cracks < 100 {micro}m in depth. The material and loading parameters that influence fatigue life in LWR environments are defined. Fatigue life is decreased significantly when five conditions are satisfied simultaneously, viz., applied strain range, service temperature, dissolved oxygen in water, and S content in steel are above a threshold level, and loading strain rate is below a threshold value. Statistical models have been developed for estimating the fatigue life of these steels in LWR environments. The significance of the effect of environment on the current Code design curve is evaluated.

  7. Metrics for the Evaluation of Light Water Reactor Accident Tolerant Fuel

    SciTech Connect

    Shannon M. Bragg-Sitton

    2001-09-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of LWRs became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of accident tolerant fuel (ATF) development is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. The U.S. Department of Energy is sponsoring multiple teams to develop ATF concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This paper summarizes technical evaluation methodology proposed in the U.S. to aid in the optimization and down-selection of candidate ATF designs. This methodology will continue to be refined via input from the research community and industry, such that it is available to support the planned down-selection of ATF concepts in 2016.

  8. Plan for Demonstration of Online Monitoring for the Light Water Reactor Sustainability Online Monitoring Project

    SciTech Connect

    Magdy S. Tawfik; Vivek Agarwal; Nancy J. Lybeck

    2011-09-01

    Condition based online monitoring technologies and development of diagnostic and prognostic methodologies have drawn tremendous interest in the nuclear industry. It has become important to identify and resolve problems with structures, systems, and components (SSCs) to ensure plant safety, efficiency, and immunity to accidents in the aging fleet of reactors. The Machine Condition Monitoring (MCM) test bed at INL will be used to demonstrate the effectiveness to advancement in online monitoring, sensors, diagnostic and prognostic technologies on a pilot-scale plant that mimics the hydraulics of a nuclear plant. As part of this research project, INL will research available prognostics architectures and their suitability for deployment in a nuclear power plant. In addition, INL will provide recommendation to improve the existing diagnostic and prognostic architectures based on the experimental analysis performed on the MCM test bed.

  9. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    SciTech Connect

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  10. A procedure for evaluating residual life of major components in light water reactors

    SciTech Connect

    Uchida, S.; Fujimori, H.; Ibe, E.; Kuniya, J.; Hayashi, M.; Fuse, M.; Yamauchi, K.

    1995-12-31

    A computer program for evaluating residual life of major components in boiling water reactors is proposed. It divides the stress corrosion cracking process into two stages; a probabilistic crack generation stage and a deterministic crack propagation one. The minimum period of the crack generation stage is evaluated assuming an exponential distribution of the stage. The crack propagation rate is calculated by the slip-dissolution/film-rupture model. The neutron flux and fluence dependence of the neutron radiation effects on material properties was evaluated by using theoretical models of radiation damage. The computer program works on an engineering work station. Evaluated results are displayed as a map of the residual life, or as graphs of crack length evolution.

  11. Spent fuel shipping cask handling capability assessment of 27 selected light water reactors

    SciTech Connect

    Konzek, G.J.; Daling, P.M.

    1984-11-01

    This report presents an assessment of the spent fuel shipping cask handling capabilities of those nuclear plants currently projected to lose full core reserve capability in their spent fuel storage basins in the near future. The purpose of this assessment is to determine which cask types, in the current fleet, each of the selected reactors can handle. The cask handling capability of a nuclear plant depends upon both external and internal conditions at the plant. The availability of a rail spur, the lifting capacity of the crane, the adequacy of clearances in the cask receiving, loading, and decontamination areas and similar factors can limit the types of casks that can be utilized at a particular plant. This report addresses the major facility capabilities used in assessing the types of spent fuel shipping casks that can be handled at each of the 27 selected nuclear plants approaching a critical storage situation. The results of this study cannot be considered to be final and are not intended to be used to force utilities to ship by a particular mode. In addition, many utilities have never shipped spent fuel. Readers are cautioned that the results of this study reflect the current situation at the selected plants and are based on operator perceptions and guidance from NRC related to the control of heavy loads at nuclear power plants. Thus, the cask handling capabilities essentially represent snap-shots in time and could be subject to change as plants further analyze their capabilities, even in the near-term. The results of this assessment indicate that 48% of the selected plants have rail access and 59% are judged to be candidates for overweight truck shipments (with 8 unknowns due to unavailability of verifiable data). Essentially all of the reactors can accommodate existing legal-weight truck casks. 12 references, 1 figure, 4 tables.

  12. Advanced Light Water Reactor Plants System 80+{trademark} Design Certification Program. Annual progress report, October 1, 1992--September 30, 1993

    SciTech Connect

    Not Available

    1993-12-31

    The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW{sub t} (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment.

  13. Validation efforts for the neutronics of a plutonium erbium zirconium oxide inert matrix light water reactor fuel

    NASA Astrophysics Data System (ADS)

    Paratte, J. M.; Chawla, R.; Früh, R.; Joneja, O. P.; Pelloni, S.; Pralong, C.

    1999-08-01

    Light water reactor (LWR) neutronics codes and cross-section libraries need further qualification when used for the calculation of inert matrix fuel (IMF) cells. Three types of validation efforts have been undertaken for the PuO 2-Er 2O 3-ZrO 2 IMF concept under development at the Paul Scherrer Institute (PSI). Firstly, the PSI calculational scheme, based on the BOXER code and its data library, has been applied to the analysis of a range of LWR experiments with PuO 2-UO 2 fuel, conducted earlier at PSI's PROTEUS facility. The generally good agreement obtained between calculated and measured parameters gives confidence in the ability of the employed calculational scheme to correctly modelize Pu-containing fuel cells. Secondly, reactivity effects of various burnable poisons in a ZrO 2 matrix were measured in the CROCUS reactor of the Swiss Federal Institute of Technology at Lausanne. Modelling these experiments with BOXER resulted in satisfactory prediction of measured reactivity ratios (relative to a soluble-boron standard) for most of the experimental rods employed. This was particularly the case for experiments with erbium, as well as with mixtures of erbium and europium (the latter being used to simulate the effects of overlapping resonances, as would be expected in the case of a Pu-Er IMF). Finally, as there are no experimental results available from power reactors employing IMFs, the validation of burnup calculations (at the cell level) has been based on results obtained in the framework of an international benchmark exercise on the physics of LWRs employing IMFs. Certain discrepancies in calculated parameters have been observed in this context, several of which can be attributed to specific differences in cross-section libraries.

  14. Validation Work to Support the Idaho National Engineering and Environmental Laboratory Calculational Burnup Methodology Using Shippingport Light Water Breeder Reactor (LWBR) Spent Fuel Assay Data

    SciTech Connect

    J. W. Sterbentz

    1999-08-01

    Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a depletion methodology previously employed to evaluate many of the radionuclide inventories for spent nuclear fuels at the Idaho National Engineering and Environmental Laboratory. The primary goal of the calculational task was to further support the validation of this particular calculational methodology and its application to diverse reactor types and fuels. Result comparisons between the calculated and measured mass concentrations in the three rods indicate good agreement for the three major uranium isotopes (U-233, U-234, U-235) with differences of less than 20%. For the seed and standard blanket rod, the U-233 and U-234 differences were within 5% of the measured values (these two isotopes alone represent greater than 97% of the EOL total uranium mass). For the major krypton and xenon fission product isotopes, differences of less than 20% and less than 30% were observed, respectively. In general, good agreement was obtained for nearly all the measured isotopes. For these isotopes exhibiting significant differences, possible explanations are discussed in terms of measurement uncertainty, complex transmutations, etc.

  15. Investigation and Application of Neutron Damage to Bipolar Transistors in Light Water Reactor Dosimetry.

    NASA Astrophysics Data System (ADS)

    Roknizadeh, Mansour

    environment can be determined. The developed techniques have been used in five reactor dosimetry experiments. Four of these experiments were conducted at reactor buildings of Unit-I and Unit-II of the Arkansas Nuclear One power plant. Another experiment was performed at the nuclear reactor laboratory of University of Illinois. In these experiments, the 1-MeV equivalent fluence rates for several irradiation positions have been calculated.

  16. The potential impact of enhanced accident tolerant cladding materials on reactivity initiated accidents in light water reactors

    SciTech Connect

    Brown, Nicholas R.; Wysocki, Aaron J.; Terrani, Kurt A.; Xu, Kevin G.; Wachs, Daniel M.

    2017-01-01

    Here, advanced cladding materials with potentially enhanced accident tolerance will yield different light-water-reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to cladding material properties, reactor physics, thermal, and hydraulic characteristics. Differences in reactors physics characteristics are driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and also by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermal hydraulic limits after transition from the current zirconium alloy cladding to the advanced materials will also affect the transient response of the integral fuel. This paper describes three-dimensional nodal kinetics simulations of a reactivity-initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon-carbide (SiC-SiC)-based cladding materials. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus that of reference Zr cladding is predominantly due to differences in (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores resulting from hardened (or softened) spectrum. This study shows similar behavior for SiC-SiC-based cladding configurations on the transient response versus reference Zircaloy cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. This is due to the shorter neutron generation time of the models with FeCrAl cladding. Therefore, the FeCrAl-based cases have a more

  17. The potential impact of enhanced accident tolerant cladding materials on reactivity initiated accidents in light water reactors

    DOE PAGES

    Brown, Nicholas R.; Wysocki, Aaron J.; Terrani, Kurt A.; ...

    2017-01-01

    Here, advanced cladding materials with potentially enhanced accident tolerance will yield different light-water-reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to cladding material properties, reactor physics, thermal, and hydraulic characteristics. Differences in reactors physics characteristics are driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and also by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermal hydraulic limits after transition from the current zirconium alloy cladding to the advanced materials will also affect the transientmore » response of the integral fuel. This paper describes three-dimensional nodal kinetics simulations of a reactivity-initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon-carbide (SiC-SiC)-based cladding materials. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus that of reference Zr cladding is predominantly due to differences in (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores resulting from hardened (or softened) spectrum. This study shows similar behavior for SiC-SiC-based cladding configurations on the transient response versus reference Zircaloy cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. This is due to the shorter neutron generation time of the models with FeCrAl cladding. Therefore, the FeCrAl-based cases have

  18. Assessment of spent nuclear fuel shipping cask handling capabilities of commercial light water reactors

    SciTech Connect

    Daling, P.M.

    1985-08-01

    Realistic truck/rail modal fractions are specifically needed to support the Monitored Retrievable Storage (MRS) and repository facility designs and envirionmental assessment activities. The objective of this study was to evaluate the spent fuel shipping cask handling capabilities at operating and planned commercial LWRs and use this information to estimate realistic truck/rail modal fractions. The cask handling parameter data collected in this study includes cask handling crane capabilities, dimensions of loading pools, structural limits, availability of rail service, past experience with spent fuel shipments (i.e., which cask was used.), and any other conditions which could impede or preclude use of a particular shipping cask. The results of this evaluation are presented for each reactor. A summary of the results which indicates the number of plants that are capable of handling each transport mode is presented. Note that two types of highway shipments are considered; legal-weight truck (LWT) and overweight truck (OWT). The primary differences between these two types of highway shipments are the size and cargo capacity of the spent fuel shipping casks. The OWT cask is roughly 50% heavier, 50% larger in diameter, and has a 300% larger cargo capacity. As a result of this size differential, some plants are capable of handling LWT casks but not OWT casks.

  19. Fracture mechanics models developed for piping reliability assessment in light water reactors: piping reliability project

    SciTech Connect

    Harris, D.O.; Lim, E.Y.; Dedhia, D.D.; Woo, H.H.; Chou, C.K.

    1982-06-01

    The efforts concentrated on modifications of the stratified Monte Carlo code called PRAISE (Piping Reliability Analysis Including Seismic Events) to make it more widely applicable to probabilistic fracture mechanics analysis of nuclear reactor piping. Pipe failures are considered to occur as the result of crack-like defects introduced during fabrication, that escape detection during inspections. The code modifications allow the following factors in addition to those considered in earlier work to be treated: other materials, failure criteria and subcritical crack growth characteristic; welding residual and vibratory stresses; and longitudinal welds (the original version considered only circumferential welds). The fracture mechanics background for the code modifications is included, and details of the modifications themselves provided. Additionally, an updated version of the PRAISE user's manual is included. The revised code, known as PRAISE-B was then applied to a variety of piping problems, including various size lines subject to stress corrosion cracking and vibratory stresses. Analyses including residual stresses and longitudinal welds were also performed.

  20. Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors

    SciTech Connect

    Iouri I. Balachov; Takao Kobayashi; Francis Tanzella; Indira Jayaweera; Palitha Jayaweera; Petri Kinnunen; Martin Bojinov; Timo Saario

    2004-11-17

    This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.

  1. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    SciTech Connect

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177{degrees}C (350{degrees}F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program.

  2. Boric acid corrosion of light water reactor pressure vessel head materials.

    SciTech Connect

    Park, J.-H.; Chopra, O. K.; Natesan, K.; Shack, W. J.; Cullen, Jr.; W. H.; Energy Technology; USNRC

    2005-01-01

    This work presents experimental data on electrochemical potential and corrosion rates for the materials found in the reactor pressure vessel head and control rod drive mechanism (CRDM) nozzles in boric acid solutions of varying concentrations at temperatures of 95-316 C. Tests were conducted in (a) high-temperature, high-pressure aqueous solutions with a range of boric acid concentrations, (b) high-temperature (150-316 C)H-B-Osolutions at ambient pressure, in wet and dry conditions, and (c) low-temperature (95 C) saturated, aqueous, boric acid solutions. These correspond to the following situations: (a) low leakage through the nozzle and nozzle/head annulus plugged, (b) low leakage through the nozzle and nozzle/head annulus open, and (c) significant cooling due to high leakage and nozzle/head annulus open. The results showed significant corrosion only for the low-alloy steel and no corrosion for Alloy 600 or 308 stainless steel cladding. Also, corrosion rates were significant in saturated boric acid solutions, and no material loss was observed in H-B-O solution in the absence of moisture. The results are compared with the existing corrosion/wastage data in the literature.

  3. Identification and evaluation of facilitation techniques for decommissioning light water power reactors

    SciTech Connect

    LaGuardia, T.S.; Risley, J.F.

    1986-06-01

    This report describes a study sponsored by the US Nuclear Regulatory Commission to identify practical techniques to facilitate the decommissioning of nuclear power generating facilities. The objective of these ''facilitation techniques'' is to reduce the radioactive exposures and/or volumes of waste generated during the decommissioning process. The report presents the possible facilitation techniques identified during the study and discusses the corresponding facilitation of the decommissioning process. Techniques are categorized by their applicability of being implemented during the three stages of power reactor life: design/construction, operation, or decommissioning. Detailed cost-benefit analyses were performed for each technique to determine the anticipated exposure and/or radioactive waste reduction; the estimated costs for implementing each technique were then calculated. Finally, these techniques were ranked by their effectiveness in facilitating the decommissioning process. This study is a part of the Nuclear Regulatory Commission's evaluation of decommissioning policy and its modification of regulations pertaining to the decommissioning process. The findings can be used by the utilities in the planning and establishment of activities to ensure that all objectives of decommissioning will be achieved.

  4. Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

    SciTech Connect

    Korsah, K.; Clark, R.L.; Wood, R.T.

    1994-04-01

    Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I&C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems` environmental qualification and functional reliability. To bound the problem of new I&C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I&C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I&C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software.

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

    SciTech Connect

    Hallbert, Bruce P

    2015-01-01

    Reliable instrumentation, information, and control systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration. The NPP owners and operators realize that this analog technology represents a significant challenge to sustaining the operation of the current fleet of NPPs. Beyond control systems, new technologies are needed to monitor and characterize the effects of aging and degradation in critical areas of key structures, systems, and components. The objective of the efforts sponsored by the U.S. Department of Energy is to develop, demonstrate, and deploy new digital technologies for II&C architectures and provide monitoring capabilities to ensure the continued safe, reliable, and economic operation of the nation’s NPPs.

  6. Acoustic emission and guided ultrasonic waves for detection and continuous monitoring of cracks in light water reactor components

    SciTech Connect

    Meyer, R. M.; Coble, J.; Ramuhalli, P.; Watson, B.; Cumblidge, S. E.; Doctor, S. R.; Bond, L. J.

    2012-07-01

    Acoustic emission (AE) and guided ultrasonic waves (GUW) are considered for continuous monitoring and detection of cracks in Light Water Reactor (LWR) components. In this effort, both techniques are applied to the detection and monitoring of fatigue crack growth in a full scale pipe component. AE results indicated crack initiation and rapid growth in the pipe, and significant GUW responses were observed in response to the growth of the fatigue crack. After initiation, the crack growth was detectable with AE for approximately 20,000 cycles. Signals associated with initiation and rapid growth were distinguished based on total rate of activity and differences observed in the centroid frequency of hits. An intermediate stage between initiation and rapid growth was associated with significant energy emissions, though few hits. GUW exhibit a nearly monotonic trend with crack length with an exception of measurements obtained at crack lengths of 41 mm and 46 mm. Coupling variability and shadowing by the electro-discharge machining (EDM) starter notch set the lower limit of detectability. (authors)

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

    NASA Astrophysics Data System (ADS)

    Yamamoto, Y.; Pint, B. A.; Terrani, K. A.; Field, K. G.; Yang, Y.; Snead, L. L.

    2015-12-01

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

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

    DOE PAGES

    Hallbert, Bruce P

    2015-01-01

    Reliable instrumentation, information, and control systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration. The NPP owners and operators realize that this analog technology represents a significant challenge to sustaining the operation of the current fleet of NPPs. Beyond control systems, new technologies are neededmore » to monitor and characterize the effects of aging and degradation in critical areas of key structures, systems, and components. The objective of the efforts sponsored by the U.S. Department of Energy is to develop, demonstrate, and deploy new digital technologies for II&C architectures and provide monitoring capabilities to ensure the continued safe, reliable, and economic operation of the nation’s NPPs.« less

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

    DOE PAGES

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

    2015-10-19

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

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

    SciTech Connect

    Yamamoto, Yukinori; Pint, Bruce A.; Terrani, Kurt A.; Field, Kevin G.; Yang, Ying; Snead, Lance Lewis

    2015-10-19

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

  11. Transmutation Performance Analysis for Inert Matrix Fuels in Light Water Reactors and Computational Neutronics Methods Capabilities at INL

    SciTech Connect

    Michael A. Pope; Samuel E. Bays; S. Piet; R. Ferrer; Mehdi Asgari; Benoit Forget

    2009-05-01

    The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF, thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCmCf, PuNpAm, PuNp, and Pu. The results of this infinite lattice assembly-wise study using SCALE 5.1 indicate that it may be feasible to recycle TRU in this manner using an otherwise typical PWR assembly without violating peaking factor limits.

  12. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    SciTech Connect

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  13. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

    SciTech Connect

    Smith, Curtis; Mandelli, Diego; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

  14. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Arzu Alpan, F.

    2016-02-01

    This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  15. DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program. Joint Research and Development Plan

    SciTech Connect

    Williams, Don

    2014-04-01

    Nuclear power has contributed almost 20% of the total amount of electricity generated in the United States over the past two decades. High capacity factors and low operating costs make nuclear power plants (NPPs) some of the most economical power generators available. Further, nuclear power remains the single largest contributor (nearly 70%) of non-greenhouse gas-emitting electric power generation in the United States. Even when major refurbishments are performed to extend operating life, these plants continue to represent cost-effective, low-carbon assets to the nation’s electrical generation capability. By the end of 2014, about one-third of the existing domestic fleet will have passed their 40th anniversary of power operations, and about one-half of the fleet will reach the same 40-year mark within this decade. Recognizing the challenges associated with pursuing extended service life of commercial nuclear power plants, the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) and the Electric Power Research Institute (EPRI) have established separate but complementary research and development programs (DOE-NE’s Light Water Reactor Sustainability [LWRS] Program and EPRI’s Long-Term Operations [LTO] Program) to address these challenges. To ensure that a proper linkage is maintained between the programs, DOE-NE and EPRI executed a memorandum of understanding in late 2010 to “establish guiding principles under which research activities (between LWRS and LTO) could be coordinated to the benefit of both parties.” This document represents the third annual revision to the initial version (March 2011) of the plan as called for in the memorandum of understanding.

  16. Light Water Reactor Sustainability Program: Computer-Based Procedures for Field Activities: Results from Three Evaluations at Nuclear Power Plants

    SciTech Connect

    Oxstrand, Johanna; Le Blanc, Katya; Bly, Aaron

    2014-09-01

    The Computer-Based Procedure (CBP) research effort is a part of the Light-Water Reactor Sustainability (LWRS) Program, which is a research and development (R&D) program sponsored by Department of Energy (DOE) and performed in close collaboration with industry R&D programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants. One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. One area that could yield tremendous savings in increased efficiency and safety is in improving procedure use. Nearly all activities in the nuclear power industry are guided by procedures, which today are printed and executed on paper. This paper-based procedure process has proven to ensure safety; however, there are improvements to be gained. Due to its inherent dynamic nature, a CBP provides the opportunity to incorporate context driven job aids, such as drawings, photos, and just-in-time training. Compared to the static state of paper-based procedures (PBPs), the presentation of information in CBPs can be much more flexible and tailored to the task, actual plant condition, and operation mode. The dynamic presentation of the procedure will guide the user down the path of relevant steps, thus minimizing time spent by the field worker to evaluate plant conditions and decisions related to the applicability of each step. This dynamic presentation of the procedure also minimizes the risk of conducting steps out of order and/or incorrectly assessed applicability of steps.

  17. Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

    SciTech Connect

    Bruce P. Hallbert; J. J. Persensky; Carol Smidts; Tunc Aldemir; Joseph Naser

    2009-08-01

    The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I&C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I&C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R&D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R&D to develop scientific knowledge in the areas of: • Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs) • Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information • New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available electricity generation

  18. Verification and Validation of Neutronic/Thermalhydraulic 3D-Time Dependent Model for Treatment of Super-critical States of Light water Research Reactors Accidents

    SciTech Connect

    Khaled, S.M.

    2015-07-01

    This work presents the Verification and testing both the neutronic and thermal-hydraulics response of the positive reactivity-initiated power excursion accidents in small light water research reactors. Some research reactors have to build its own severe accidents code system. In this sense, a 3D space-time-dependent neutron diffusion models with thermal hydraulic feedback have been introduced, compared and tested both experimentally at criticality 14-cent and theoretically up to 1.5 $ with a number of similar codes. The results shows that no expected core failure or moderator boiling. (author)

  19. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Pressurized Water Reactor Standard Core Loading Benchmark Problem

    NASA Astrophysics Data System (ADS)

    Arzu Alpan, F.; Kulesza, Joel A.

    2016-02-01

    This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237Np(n,f) in the cavity capsule using BUGLE-B7.

  20. Research at the CEA in the field of safety in 2nd and 3rd generation light water reactors

    NASA Astrophysics Data System (ADS)

    Billot, Philippe

    2012-05-01

    The research programs at the CEA in the field of safety in nuclear reactors are carried out in a framework of international partnerships. Their purpose is to develop studies on: The methods allowing for the determination of earthquake hazards and their consequences; The behaviour of fuel in an accident situation; The comprehension of deflagration and detonation phenomena of hydrogen and the search for effective prevention methods involving an explosion risk; The cooling of corium in order to stop its progression in and outside the vessel thereby reducing the risk of perforating the basemat; The behaviour of the different fission product families according to their volatility for the UO2 and MOX fuels.

  1. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    SciTech Connect

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

  2. Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

    SciTech Connect

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

    1996-09-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations.

  3. Understanding the Atomic-Level Chemistry and Structure of Oxide Deposits on Fuel Rods in Light Water Nuclear Reactors Using First Principles Methods

    NASA Astrophysics Data System (ADS)

    Rak, Zs.; O'Brien, C. J.; Brenner, D. W.; Andersson, D. A.; Stanek, C. R.

    2016-11-01

    The results of recent studies are discussed in which first principles calculations at the atomic level have been used to expand the thermodynamic database for science-based predictive modeling of the chemistry, composition and structure of unwanted oxides that deposit on the fuel rods in pressurized light water nuclear reactors. Issues discussed include the origin of the particles that make up deposits, the structure and properties of the deposits, and the forms by which boron uptake into the deposits can occur. These first principles approaches have implications for other research areas, such as hydrothermal synthesis and the stability and corrosion resistance of other materials under other extreme conditions.

  4. Preliminary analysis of the postulated changes needed to achieve rail cask handling capabilities at selected light water reactors

    SciTech Connect

    Konzek, G.J.

    1986-02-01

    Reactor-specific railroad and crane information for all LWRs in the US was extracted from current sources of information. Based on this information, reactors were separated into two basic groups consisting of reactors with existing, usable rail cask capabilities and those without these capabilities. The latter group is the main focus of this study. The group of reactors without present rail cask handling capabilities was further separated into two subgroups consisting of reactors considered essentially incapable of handling a large rail cask of about 100 tons and reactors where postulated facility changes could result in rail cask handling capabilities. Based on a selected population of 127 reactors, the results of this assessment indicate that usable rail cask capabilities exist at 83 (65%) of the reactors. Twelve (27%) of the remaining 44 reactors are deemed incapable of handling a large rail cask without major changes, and 32 reactors are considered likely candidates for potentially achieving rail cask handling capabilities. In the latter group, facility changes were postulated that would conceptually enable these reactors to handle large rail casks. The estimated cost per plant of required facility changes varied widely from a high of about $35 million to a low of <$0.3 million. Only 11 of the 32 plants would require crane upgrades. Spur track and right-of-way costs would apparently vary widely among sites. These results are based on preliminary analyses using available generic cost data. They represent lower bound values that are useful for developing an initial assessment of the viability of the postulated changes on a system-wide basis, but are not intended to be absolute values for specific reactors or sites.

  5. Economic Analyiss of "Symbiotic" Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

    SciTech Connect

    Williams, Kent Alan; Shropshire, David E.

    2009-01-01

    A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

  6. Reactor production of Thorium-229

    DOE PAGES

    Boll, Rose Ann; Murphy, Karen E.; Denton, David L.; ...

    2016-05-03

    Limited availability of 229Th for clinical applications of 213Bi necessitates investigation of alternative production routes. In reactor production, 229Th is produced from neutron transmutation of 226Ra, 228Ra, 227Ac and 228Th. Here, we evaluate irradiations of 226Ra, 228Ra, and 227Ac targets at the ORNL High Flux Isotope Reactor.

  7. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    SciTech Connect

    Not Available

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  8. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    SciTech Connect

    Not Available

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  9. Validation of standardized computer analyses for licensing evaluation/TRITON two-dimensional and three-dimensional models for light water reactor fuel

    SciTech Connect

    Bowman, S. M.; Gill, D. F.

    2006-07-01

    The isotopic depletion capabilities of the new Standardized Computer Analyses for Licensing Evaluation control module TRITON, coupled with ORIGEN-S, were evaluated using spent fuel assays from several commercial light water reactors with both standard and mixed-oxide fuel assemblies. Calculations were performed using the functional modules NEWT and KENO-VI. NEWT is a two-dimensional, arbitrary-geometry, discrete-ordinates transport code, and KENO-VI is a three-dimensional Monte Carlo transport code capable of handling complex three-dimensional geometries. To validate the codes and data used in depletion calculations, numerical predictions were compared with experimental measurements for a total of 29 samples taken from the Calvert Cliffs, Obrigheim, and San Onofre pressurized water reactors and the Gundremmingen boiling water reactor. Similar comparisons have previously been performed at the Oak Ridge National Laboratory for the one-dimensional SAS2H control module. The SAS2H, TRITON/KENO-VI, and TRITON/NEWT results were compared for corresponding samples. All analyses showed that TRITON/KENO-VI and TRITON/NEWT produced typically similar or better results than SAS2H. The calculations performed in this validation study demonstrate that the depletion capabilities of TRITON accurately model spent fuel depletion and decay. (authors)

  10. Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

    NASA Astrophysics Data System (ADS)

    Carroll, Spencer

    As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel

  11. An innovative fuel design concept for improved Light Water Reactor performance and safety. Final technical report, April 24, 1992--April 23, 1993

    SciTech Connect

    Tulenko, J.S.; Connell, R.G.

    1993-12-31

    The primary goal of this research is to develop a new fuel design which will have improved thermal/mechanical performance characteristics greatly superior to current thermal and mechanical design performance. The mechanical/thermal constraints define the lifetime of the fuel, the maximum power at which the fuel can be operated, the probability of fuel failure over core lifetime, and the integrity of a core during a transient excursion. The thermal/mechanical limits act to degrade fuel integrity when they are violated. The purpose of this project is to investigate a novel design for light water reactor fuel which will extend fuel performance limits and improve reactor safety even further than is currently achieved. This project is investigating liquid metal bonding of LWR fuel in order to radically decrease fuel centerline temperatures which has major performance and safety benefits. The project will verify the compatibility of the liquid metal bond with both the fuel pellets and cladding material, verify the performance enhancement features of the new design over the fuel lifetime, and verify the economic fabricability of the concept and will show how this concept will benefit the LWR nuclear industry.

  12. Analysis of Radiation Damage in Light Water Reactors: Comparison of Cluster Analysis Methods for the Analysis of Atom Probe Data.

    PubMed

    Hyde, Jonathan M; DaCosta, Gérald; Hatzoglou, Constantinos; Weekes, Hannah; Radiguet, Bertrand; Styman, Paul D; Vurpillot, Francois; Pareige, Cristelle; Etienne, Auriane; Bonny, Giovanni; Castin, Nicolas; Malerba, Lorenzo; Pareige, Philippe

    2017-01-30

    Irradiation of reactor pressure vessel (RPV) steels causes the formation of nanoscale microstructural features (termed radiation damage), which affect the mechanical properties of the vessel. A key tool for characterizing these nanoscale features is atom probe tomography (APT), due to its high spatial resolution and the ability to identify different chemical species in three dimensions. Microstructural observations using APT can underpin development of a mechanistic understanding of defect formation. However, with atom probe analyses there are currently multiple methods for analyzing the data. This can result in inconsistencies between results obtained from different researchers and unnecessary scatter when combining data from multiple sources. This makes interpretation of results more complex and calibration of radiation damage models challenging. In this work simulations of a range of different microstructures are used to directly compare different cluster analysis algorithms and identify their strengths and weaknesses.

  13. A SCOPING STUDY: Development of Probabilistic Risk Assessment Models for Reactivity Insertion Accidents During Shutdown In U.S. Commercial Light Water Reactors

    SciTech Connect

    S. Khericha

    2011-06-01

    This report documents the scoping study of developing generic simplified fuel damage risk models for quantitative analysis from inadvertent reactivity insertion events during shutdown (SD) in light water pressurized and boiling water reactors. In the past, nuclear fuel reactivity accidents have been analyzed both mainly deterministically and probabilistically for at-power and SD operations of nuclear power plants (NPPs). Since then, many NPPs had power up-rates and longer refueling intervals, which resulted in fuel configurations that may potentially respond differently (in an undesirable way) to reactivity accidents. Also, as shown in a recent event, several inadvertent operator actions caused potential nuclear fuel reactivity insertion accident during SD operations. The set inadvertent operator actions are likely to be plant- and operation-state specific and could lead to accident sequences. This study is an outcome of the concern which arose after the inadvertent withdrawal of control rods at Dresden Unit 3 in 2008 due to operator actions in the plant inadvertently three control rods were withdrawn from the reactor without knowledge of the main control room operator. The purpose of this Standardized Plant Analysis Risk (SPAR) Model development project is to develop simplified SPAR Models that can be used by staff analysts to perform risk analyses of operating events and/or conditions occurring during SD operation. These types of accident scenarios are dominated by the operator actions, (e.g., misalignment of valves, failure to follow procedures and errors of commissions). Human error probabilities specific to this model were assessed using the methodology developed for SPAR model human error evaluations. The event trees, fault trees, basic event data and data sources for the model are provided in the report. The end state is defined as the reactor becomes critical. The scoping study includes a brief literature search/review of historical events, developments of

  14. Reduction of the Radiotoxicity of Spent Nuclear Fuel Using a Two-Tiered System Comprising Light Water Reactors and Accelerator-Driven Systems

    SciTech Connect

    Trellue, Holly R.

    2003-06-01

    Two main issues regarding the disposal of spent nuclear fuel from nuclear reactors in the United States in the geological repository Yucca Mountain are: (1) Yucca Mountain is not designed to hold the amount of fuel that has been and is proposed to be generated in the next few decades, and (2) the radiotoxicity (i.e., biological hazard) of the waste (particularly the actinides) does not decrease below that of natural uranium ore for hundreds of thousands of years. One solution to these problems may be to use transmutation to convert the nuclides in spent nuclear fuel to ones with shorter half-lives. Both reactor and accelerator-based systems have been examined in the past for transmutation; there are advantages and disadvantages associated with each. By using existing Light Water Reactors (LWRs) to burn a majority of the plutonium in spent nuclear fuel and Accelerator-Driven Systems (ADSs) to transmute the remainder of the actinides, the benefits of each type of system can be realized. The transmutation process then becomes more efficient and less expensive. This research searched for the best combination of LWRs with multiple recycling of plutonium and ADSs to transmute spent nuclear fuel from past and projected nuclear activities (assuming little growth of nuclear energy). The neutronic design of each system is examined in detail although thermal hydraulic performance would have to be considered before a final system is designed. The results are obtained using the Monte Carlo burnup code Monteburns, which has been successfully benchmarked for MOX fuel irradiation and compared to other codes for ADS calculations. The best combination of systems found in this research includes 41 LWRs burning mixed oxide fuel with two recycles of plutonium (~40 years operation each) and 53 ADSs to transmute the remainder of the actinides from spent nuclear fuel over the course of 60 years of operation.

  15. Reactor production of Thorium-229.

    PubMed

    Hogle, Susan; Boll, Rose Ann; Murphy, Karen; Denton, David; Owens, Allison; Haverlock, Tamara J; Garland, Marc; Mirzadeh, Saed

    2016-08-01

    Limited availability of (229)Th for clinical applications of (213)Bi necessitates investigation of alternative production routes. In reactor production, (229)Th is produced from neutron transmutation of (226)Ra, (228)Ra, (227)Ac and (228)Th. Irradiations of (226)Ra, (228)Ra, and (227)Ac targets at the Oak Ridge National Laboratory High Flux Isotope Reactor result in yields of (229)Th at 26 days of 74.0±7.4MBq/g, 260±10MBq/g, and 1200±50MBq/g, respectively. Intermediate radionuclide yields and cross sections are also studied.

  16. Reactor production of Thorium-229

    SciTech Connect

    Boll, Rose Ann; Murphy, Karen E.; Denton, David L.; Tamara J. Haverlock; Garland, Marc A.; Mirzadeh, Saed; Hogle, Susan; Owens, Allison

    2016-05-03

    Limited availability of 229Th for clinical applications of 213Bi necessitates investigation of alternative production routes. In reactor production, 229Th is produced from neutron transmutation of 226Ra, 228Ra, 227Ac and 228Th. Here, we evaluate irradiations of 226Ra, 228Ra, and 227Ac targets at the ORNL High Flux Isotope Reactor.

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

    PubMed

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

    2015-01-01

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

  18. Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors (Progress report for work through June 2002, 12th quarterly report)

    SciTech Connect

    Mac Donald, Philip Elsworth

    2002-09-01

    The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: · Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs, · Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically, · Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards, · Task 4 will determine the long-term stability of ThO2/UO2 high-level waste, and · Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements.

  19. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    SciTech Connect

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  20. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR)

    NASA Astrophysics Data System (ADS)

    Flaspoehler, Timothy; Petrovic, Bojan

    2016-02-01

    One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV). While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR) Power Plant, including DPA (displacements per atom) and fast neutron fluence (>1 MeV and >0.1MeV). I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling) methodology to bias a fixed-source MC (Monte Carlo) simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  1. DESCRIPTION OF THE TRITIUM-PRODUCING BURNABLE ABSORBER ROD FOR THE COMMERCIAL LIGHT WATER REACTOR TTQP-1-015 Rev 19

    SciTech Connect

    Burns, Kimberly A.; Love, Edward F.; Thornhill, Cheryl K.

    2012-02-01

    Tritium-producing burnable absorber rods (TPBARs) used in the U.S. Department of Energy’s Tritium Readiness Program are designed to produce tritium when placed in a Westinghouse or Framatome 17x17 fuel assembly and irradiated in a pressurized water reactor (PWR). This document provides an unclassified description of the current design baseline for the TPBARs. This design baseline is currently valid only for Watts Bar reactor production cores. A description of the Lead Use TPBARs will not be covered in the text of the document, but the applicable drawings, specifications and test plan will be included in the appropriate appendices.

  2. A study of a zone approach to IAEA (International Atomic Energy Agency) safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle

    SciTech Connect

    Fishbone, L.G.; Higinbotham, W.A.

    1986-06-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches.

  3. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    SciTech Connect

    Not Available

    1994-01-15

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

  4. Savannah River Site production reactor technical specifications. K Production Reactor

    SciTech Connect

    1996-02-01

    These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

  5. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    SciTech Connect

    Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  6. New Production Reactors Program Plan

    SciTech Connect

    Not Available

    1990-12-01

    Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs.

  7. DOE plutonium disposition study: Analysis of existing ABB-CE Light Water Reactors for the disposition of weapons-grade plutonium. Final report

    SciTech Connect

    Not Available

    1994-06-01

    Core reactivity and basic fuel management calculations were conducted on the selected reactors (with emphasis on the System 80 units as being the most desirable choice). Methods used were identical to those reported in the Evolutionary Reactor Report. From these calculations, the basic mission capability was assessed. The selected reactors were studied for modification, such as the addition of control rod nozzles to increase rod worth, and internals and control system modifications that might also be needed. Other system modifications studied included the use of enriched boric acid as soluble poison, and examination of the fuel pool capacities. The basic geometry and mechanical characteristics, materials and fabrication techniques of the fuel assemblies for the selected existing reactors are the same as for System 80+. There will be some differences in plutonium loading, according to the ability of the reactors to load MOX fuel. These differences are not expected to affect licensability or EPA requirements. Therefore, the fuel technology and fuel qualification sections provided in the Evolutionary Reactor Report apply to the existing reactors. An additional factor, in that the existing reactor availability presupposes the use of that reactor for the irradiation of Lead Test Assemblies, is discussed. The reactor operating and facility licenses for the operating plants were reviewed. Licensing strategies for each selected reactor were identified. The spent fuel pool for the selected reactors (Palo Verde) was reviewed for capacity and upgrade requirements. Reactor waste streams were identified and assessed in comparison to uranium fuel operations. Cost assessments and schedules for converting to plutonium disposition were estimated for some of the major modification items. Economic factors (incremental costs associated with using weapons plutonium) were listed and where possible under the scope of work, estimates were made.

  8. Greater-than-Class C low-level radioactive waste characterization. Appendix A-3: Basis for greater-than-Class C low-level radioactive waste light water reactor projections

    SciTech Connect

    Mancini, A.; Tuite, P.; Tuite, K.; Woodberry, S.

    1994-09-01

    This study characterizes low-level radioactive waste types that may exceed Class C limits at light water reactors, estimates the amounts of waste generated, and estimates radionuclide content and distribution within the waste. Waste types that may exceed Class C limits include metal components that become activated during operations, process wastes such as cartridge filters and decontamination resins, and activated metals from decommissioning activities. Operating parameters and current management practices at operating plants are reviewed and used to estimate the amounts of low-level waste exceeding Class C limits that is generated per fuel cycle, including amounts of routinely generated activated metal components and process waste. Radionuclide content is calculated for specific activated metals components. Empirical data from actual low-level radioactive waste are used to estimate radionuclide content for process wastes. Volumes and activities are also estimated for decommissioning activated metals that exceed Class C limits. To estimate activation levels of decommissioning waste, six typical light water reactors are modeled and analyzed. This study does not consider concentration averaging.

  9. Reactors are indispensable for radioisotope production.

    PubMed

    Mushtaq, Ahmad

    2010-12-01

    Radioisotopes can be produced by reactors and accelerators. For certain isotopes there could be an advantage to a certain production method. However, nowadays many reports suggest, that useful isotopes needed in medicine, industry and research could be produced efficiently and dependence on reactors using enriched U-235 may be eliminated. In my view reactors and accelerators will continue to play their role side by side in the supply of suitable and economical sources of isotopes.

  10. Crystal chemistry of sodium zirconium phosphate based simulated ceramic waste forms of effluent cations (Ba(2+), Sn(4+), Fe(3+), Cr(3+), Ni(2+) and Si(4+)) from light water reactor fuel reprocessing plants.

    PubMed

    Shrivastava, O P; Chourasia, Rashmi

    2008-05-01

    A novel concept of immobilization of light water reactor (LWR) fuel reprocessing waste effluent through interaction with sodium zirconium phosphate (NZP) has been established. Such conversion utilizes waste materials like zirconium and nickel alloys, stainless steel, spent solvent tri-butyl phosphate and concentrated solution of NaNO(3). The resultant multi component NZP material is a physically and chemically stable single phase crystalline product having good mechanical strength. The NZP matrix can also incorporate all types of fission product cations in a stable crystalline lattice structure; therefore, the resultant solid solutions deserve quantification of crystallographic data. In this communication, crystal chemistry of the two types of simulated waste forms (type I-Na(1.49)Zr(1.56)Sn(0.02)Fe(0).(28)Cr(0.07)Ni(0.07)P(3)O(12) and type II-Na(1.35)Ba(0.14)Zr(1.56)Sn(0.02)Fe(0).(28)Cr(0.07)Ni(0.07)P(2.86)Si(0.14)O(12)) has been investigated using General Structure Analysis System (GSAS) programming of the X-ray powder diffraction data. About 4001 data points of each have been subjected to Rietveld analysis to arrive at a satisfactory structural convergence of Rietveld parameters; R-pattern (R(p))=0.0821, R-weighted pattern (R(wp))=0.1266 for type I and R(p)=0.0686, R(wp)=0.0910 for type II. The structure of type I and type II waste forms consist of ZrO(6) octahedra and PO(4) tetrahedra linked by the corners to form a three-dimensional network. Each phosphate group is on a two-fold rotation axis and is linked to four ZrO(6) octahedra while zirconium octahedra lies on a three-fold rotation axis and is connected to six PO(4) tetrahedra. Though the expansion along c-axis and shrinkage along a-axis with slight distortion of bond angles in the synthesized crystal indicate the flexibility of the structure, the waste forms are basically of NZP structure. Morphological examination by SEM reveals that the size of almost rectangular parallelepiped crystallites varies

  11. Final report on the study of the effect of primary and secondary plant parameters on light-water-reactor transient behavior. Progress report, July 1-December 31, 1982

    SciTech Connect

    Baratta, A.J.; Robinson, G.E.; Kim, D.; Brownson, D.A.; Morales, J.B.; Huegel, D.; Gul, M.; Macario, F.

    1983-04-01

    Since the Three Mile Island Accident of March 1979, various approaches have been undertaken to reduce the probability of such occurrences. In general, these involve the addition of more complex and sophisticated engineered safety systems and instrumentation and control systems. Such improvements will undoubtedly improve overall plant safety. This project explores an alternative approach - the optimization of the transient behavior of the reactor and associated systems. Specifically, the ability of B and W and Westinghouse reactor to withstand a turbine trip was investigated. The project examined whether the plant's transient behavior with respect to peak reactor coolant system (RCS) behavior could be optimized. Several key parameters including pressurizer size, surge line impedence, time to SCRAM, and fuel and moderator temperature coefficients were varied. It was found that increasing pressurizer size decreased peak RCS pressure.

  12. Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

    SciTech Connect

    Seniuk, P.J.

    1996-12-01

    This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, {open_quotes}Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plants{close_quotes}. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O&M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O&M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion.

  13. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    SciTech Connect

    2014-08-31

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  14. Comparison of actinide production in traveling wave and pressurized water reactors

    SciTech Connect

    Osborne, A.G.; Smith, T.A.; Deinert, M.R.

    2013-07-01

    The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

  15. Scalable Methods for Uncertainty Quantification, Data Assimilation and Target Accuracy Assessment for Multi-Physics Advanced Simulation of Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Khuwaileh, Bassam

    High fidelity simulation of nuclear reactors entails large scale applications characterized with high dimensionality and tremendous complexity where various physics models are integrated in the form of coupled models (e.g. neutronic with thermal-hydraulic feedback). Each of the coupled modules represents a high fidelity formulation of the first principles governing the physics of interest. Therefore, new developments in high fidelity multi-physics simulation and the corresponding sensitivity/uncertainty quantification analysis are paramount to the development and competitiveness of reactors achieved through enhanced understanding of the design and safety margins. Accordingly, this dissertation introduces efficient and scalable algorithms for performing efficient Uncertainty Quantification (UQ), Data Assimilation (DA) and Target Accuracy Assessment (TAA) for large scale, multi-physics reactor design and safety problems. This dissertation builds upon previous efforts for adaptive core simulation and reduced order modeling algorithms and extends these efforts towards coupled multi-physics models with feedback. The core idea is to recast the reactor physics analysis in terms of reduced order models. This can be achieved via identifying the important/influential degrees of freedom (DoF) via the subspace analysis, such that the required analysis can be recast by considering the important DoF only. In this dissertation, efficient algorithms for lower dimensional subspace construction have been developed for single physics and multi-physics applications with feedback. Then the reduced subspace is used to solve realistic, large scale forward (UQ) and inverse problems (DA and TAA). Once the elite set of DoF is determined, the uncertainty/sensitivity/target accuracy assessment and data assimilation analysis can be performed accurately and efficiently for large scale, high dimensional multi-physics nuclear engineering applications. Hence, in this work a Karhunen-Loeve (KL

  16. Synfuel production in nuclear reactors

    DOEpatents

    Henning, C.D.

    Apparatus and method for producing synthetic fuels and synthetic fuel components by using a neutron source as the energy source, such as a fusion reactor. Neutron absorbers are disposed inside a reaction pipe and are heated by capturing neutrons from the neutron source. Synthetic fuel feedstock is then placed into contact with the heated neutron absorbers. The feedstock is heated and dissociates into its constituent synfuel components, or alternatively is at least preheated sufficiently to use in a subsequent electrolysis process to produce synthetic fuels and synthetic fuel components.

  17. Online stress corrosion crack and fatigue usages factor monitoring and prognostics in light water reactor components: Probabilistic modeling, system identification and data fusion based big data analytics approach

    SciTech Connect

    Mohanty, Subhasish M.; Jagielo, Bryan J.; Iverson, William I.; Bhan, Chi Bum; Soppet, William S.; Majumdar, Saurin M.; Natesan, Ken N.

    2014-12-10

    Nuclear reactors in the United States account for roughly 20% of the nation's total electric energy generation, and maintaining their safety in regards to key component structural integrity is critical not only for long term use of such plants but also for the safety of personnel and the public living around the plant. Early detection of damage signature such as of stress corrosion cracking, thermal-mechanical loading related material degradation in safety-critical components is a necessary requirement for long-term and safe operation of nuclear power plant systems.

  18. Simulation of (16)O (n, p) (16)N reaction rate and nitrogen-16 inventory in a high performance light water reactor with one pass core.

    PubMed

    Kebwaro, Jeremiah Monari; Zhao, Yaolin; He, Chaohui

    2014-12-01

    The rate of activation of the isotope (16)O to (16)N in a typical HPLWR one pass concept was calculated using MCNP code. A mathematical model was used to track the inventory of the radioisotope (16)N in a unit mass of coolant traversing the system. The water leaving the moderator channels has the highest activity in the circuit, but due to interaction with fresh coolant at the lower plenum, the activity is downscaled. The calculated core exit activity is higher than values reported in literature for commercial boiling water reactors.

  19. Description of the Canadian particulate-fill waste-package (WP) system for spent-nuclear fuel (SNF) and its applicability to light-water reactor SNF WPs with depleted uranium-dioxide fill

    SciTech Connect

    Forsberg, C.W.

    1997-10-20

    The US is beginning work on an advanced, light-water reactor (LWR), spent nuclear fuel (SNF), waste package (WP) that uses depleted uranium dioxide (UO{sub 2}) fill. The Canadian nuclear fuel waste management program has completed a 15-year development program of its repository concept for CANadian Deuterium Uranium (CANDU) reactor SNF. As one option, Canada has developed a WP that uses a glass-bead or silica-sand fill. The Canadian development work on fill materials inside WPs can provide a guide for the development of LWR SNF WPs using depleted uranium (DU) fill materials. This report summarizes the Canadian work, identifies similarities and differences between the Canadian design and the design being investigated in the US to use DU fill, and identifies what information is applicable to the development of a DU fill for LWR SNF WPs. In both concepts, empty WPs are loaded with SNF, the void space between the fuel pins and the outer void space between SNF assemblies and the inner WP wall would be filled with small particles, the WPs are then sealed, and the WPs are placed into the repository.

  20. Isotope production in fast reactor blankets

    SciTech Connect

    Zvonarev, A.V.; Koloskov, B.V.; Kochetkov, L.A.

    1993-12-31

    At the BR-10 research reactor radioactive isotopes are produced that are required for the production of radiopharmaceuticals and phosphor 32 used for the synthesis of biochemical labelled compounds. A procedure has been developed of uranium target irradiation followed by radiochemical processing aimed at isotope isolation of molybdenum 99, xenon 133 and iodine 131,132,and 133 isotopes mixture. Irradiation is carried out in a special channel of the radial blanket. The production of cobalt 60 at the BN-600 reactor and facilities are also described.

  1. Effects of alloying elements on radiation hardening based on loop formation of electron-irradiated light water reactor pressure vessel model steels

    NASA Astrophysics Data System (ADS)

    Nishi, Takakuni; Hashimoto, N.; Ohnuki, S.; Yamamoto, T.; Odette, G. R.

    2011-10-01

    Electron irradiations using a high voltage electron microscope were conducted on several reactor pressure vessel model alloys in order to investigate the effects of alloying elements on the formation and development of defect clusters. In addition, the effects of alloying elements on yield stress change after irradiation were considered, comparing the mean size and number density of dislocation loops with the irradiation-induced hardening. High Cu alloys formed Cu and Mn-Ni-Si rich clusters, and these are important in determining the yield stress increase. High Ni alloys formed a high density of small dislocation loops and probably Mn-Ni-Si rich cluster, which have the effect of increasing the yield stress. High P enhanced radiation-induced segregation on grain boundary, helping prevent dislocation movement.

  2. Silicon production in an aerosol reactor

    NASA Technical Reports Server (NTRS)

    Wu, J. J.; Flagan, R. C.

    1986-01-01

    An aerosol reactor system was developed in which large particles of silicon can be grown by silane pyrolysis. To grow particles to sizes larger than one micron, vapor deposition must be used to grow a relatively small number of seed particles. Suppression of nucleation is achieved by limiting the rate of gas phase chemical reactions such that the condensible products of the gas phase chemical reactions diffuse to the surface of the seed particles as rapidly as they are produced. This prevents high degrees of supersaturation and runaway nucleation during the growth process. Particles on the order of 10 microns were grown repeatedly with the present aersol reactor. The nucleation controlled aerosol reactor is, therefore, a suitable system for the production of powders that can readily be separated from the gas by aerodynamic means.

  3. Hybrid reactors. [Fuel cycle

    SciTech Connect

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  4. Pebble Bed Reactor Dust Production Model

    SciTech Connect

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-09-01

    The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors’ PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production.

  5. Evaluating Environmental, Health and Safety Impacts from Two Nuclear Fuel Cycles: A Comparative Analysis of Once-Through Uranium Use and Plutonium Recycle in Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Smith, Bethan L.

    The work presented in this dissertation represents a systems-level approach to investigate potential net impacts with respect to human health and the environment associated with transitioning to the MOC for the U.S. In Chapter 2, an updated systems-level conceptual model of the OTC is presented to more accurately portray the OTC as currently implemented in the U.S. The conceptual model is the basis for estimating the worker collective doses at each operational stage, and the first demonstration of a quantitative comparative radiological impact assessment from expected normal operations is presented. In the course of evaluating worker collective dose associated with modern OTC practices, it was found that the relative contributions from the two grouped operations (front-end operations for preparing reactor fuel and reactor operations) were substantially different from historical data and conventional wisdom. As a bookend to Chapter 2, a summary is provided that describes the nature of the differences and factors that led to these differences. Detailed information of the work as part of the published journal article based off of this corollary work is included as an Appendix (C). In Chapter 3, the study of worker collective doses from the phased introduction of reprocessing in the MOC scenario, and is presented similarly to the results in Chapter 2. MOC performance was also estimated by evaluating the radioactive waste generated that can be disposed and managed through known disposal practices in shallow-land burial. Relative to the OTC, MOC performance with respect to worker collective dose was not discernibly different; while the volume of radioactive waste generated decreased. It was found that although the sheer volume of radioactive waste avoided is large, the waste disposition pathway is known for the majority of this waste. The radioactive waste that requires disposal at a licensed off-site facility is examined in closer detail. The verification process for

  6. Reactors Save Energy, Costs for Hydrogen Production

    NASA Technical Reports Server (NTRS)

    2014-01-01

    While examining fuel-reforming technology for fuel cells onboard aircraft, Glenn Research Center partnered with Garrettsville, Ohio-based Catacel Corporation through the Glenn Alliance Technology Exchange program and a Space Act Agreement. Catacel developed a stackable structural reactor that is now employed for commercial hydrogen production and results in energy savings of about 20 percent.

  7. Silicon production in a fluidized bed reactor

    NASA Technical Reports Server (NTRS)

    Rohatgi, N. K.

    1986-01-01

    Part of the development effort of the JPL in-house technology involved in the Flat-Plate Solar Array (FSA) Project was the investigation of a low-cost process to produce semiconductor-grade silicon for terrestrial photovoltaic cell applications. The process selected was based on pyrolysis of silane in a fluidized-bed reactor (FBR). Following initial investigations involving 1- and 2-in. diameter reactors, a 6-in. diameter, engineering-scale FBR was constructed to establish reactor performance, mechanism of silicon deposition, product morphology, and product purity. The overall mass balance for all experiments indicates that more than 90% of the total silicon fed into the reactor is deposited on silicon seed particles and the remaining 10% becomes elutriated fines. Silicon production rates were demonstrated of 1.5 kg/h at 30% silane concentration and 3.5 kg/h at 80% silane concentration. The mechanism of silicon deposition is described by a six-path process: heterogeneous deposition, homogeneous decomposition, coalescence, coagulation, scavenging, and heterogeneous growth on fines. The bulk of the growth silicon layer appears to be made up of small diameter particles. This product morphology lends support to the concept of the scavenging of homogeneously nucleated silicon.

  8. Steam turbine: Alternative emergency drive for the secure removal of residual heat from the core of light water reactors in ultimate emergency situation

    SciTech Connect

    Souza Dos Santos, R.

    2012-07-01

    In 2011 the nuclear power generation has suffered an extreme probation. That could be the meaning of what happened in Fukushima Nuclear Power Plants. In those plants, an earthquake of 8.9 on the Richter scale was recorded. The quake intensity was above the trip point of shutting down the plants. Since heat still continued to be generated, the procedure to cooling the reactor was started. One hour after the earthquake, a tsunami rocked the Fukushima shore, degrading all cooling system of plants. Since the earthquake time, the plant had lost external electricity, impacting the pumping working, drive by electric engine. When operable, the BWR plants responded the management of steam. However, the lack of electricity had degraded the plant maneuvers. In this paper we have presented a scheme to use the steam as an alternative drive to maintain operable the cooling system of nuclear power plant. This scheme adds more reliability and robustness to the cooling systems. Additionally, we purposed a solution to the cooling in case of lacking water for the condenser system. In our approach, steam driven turbines substitute electric engines in the ultimate emergency cooling system. (authors)

  9. Nuclear reactors for research and radioisotope production in Argentina

    SciTech Connect

    Duran, H.H.

    1981-01-01

    In Argentina, the construction, operation, and use of research and radioisotope production reactors is and has been an important method of personnel preparation for the nuclear power program. Moreover, it is a very suitable means for technology transfer to countries developing their own nuclear programs. At present, the following research reactors are in operation in Argentina: Argentine Reactor 0 (RA-0); Argentine Reactor 1 (RA-1); Argentine Reactor 2 (RA-2); Argentine Reactor 3 (RA-3); Argentine Reactor 4 (RA-4). The Argentine Reactor 6 (RA-6), under construction, should reach criticality in 1981.

  10. Measuring the productivity of university research reactors

    SciTech Connect

    Voth, M.H.

    1989-11-01

    University Research Reactors (URRs) on 33 campuses in the United States provide valuable contributions to academic instruction and research programs. In most cases, there are no alternative diagnostic techniques to supplant the need for a reactor and associated facilities. Since URRs constitute a major financial commitment, it is important that they be operated in a productive manner. Productivity may be defined as the sum of new knowledge generated, existing knowledge transferred to others, and analytical services provided to assist in the generation of new knowledge; another definition of productivity is this sum expressed as a function of the cost incurred. In either case, a consistent measurement is difficult and more qualitative than quantitative. A uniform reporting system has been proposed that defines simplified categories through which meaningful comparisons can be performed.

  11. [The Chinese nuclear test and 'atoms for peace' as a measure for preventing nuclear armament of Japan: the nuclear non-proliferation policy of the United States and the introduction of light water reactors into Japan, 1964-1968].

    PubMed

    Yamazaki, Masakatsu

    2014-07-01

    Japan and the United States signed in 1968 a new atomic energy agreement through which US light-water nuclear reactors, including those of the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company, were to be introduced into Japan. This paper studies the history of negotiations for the 1968 agreement using documents declassified in the 1990s in the US and Japan. After the success of the Chinese nuclear test in October 1964, the United States became seriously concerned about nuclear armament of other countries in Asia including Japan. Expecting that Japan would not have its own nuclear weapons, the US offered to help the country to demonstrate its superiority in some fields of science including peaceful nuclear energy to counter the psychological effect of the Chinese nuclear armament. Driven by his own political agenda, the newly appointed Prime Minister Eisaku Sato responded to the US expectation favorably. When he met in January 1965 with President Johnson, Sato made it clear that Japan would not pursue nuclear weapons. Although the US continued its support after this visit, it nevertheless gave priority to the control of nuclear technology in Japan through the bilateral peaceful nuclear agreement. This paper argues that the 1968 agreement implicitly meant a strategic measure to prevent Japan from going nuclear and also a tactic to persuade Japan to join the Nuclear Non -Proliferation Treaty.

  12. Fusion reactors for hydrogen production via electrolysis

    NASA Astrophysics Data System (ADS)

    Fillo, J. A.; Powell, J. R.; Steinberg, M.

    The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and supplement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Depending on design, electric generation efficiencies of 40 to 60% and hydrogen production efficiencies by high temperature electrolysis of 50 to 70% are projected for fusion reactors using high temperature blankets.

  13. Continuous-flow biodiesel production using slit-channel reactors.

    PubMed

    Kalu, Egwu Eric; Chen, Ken S; Gedris, Tom

    2011-03-01

    Slit-channel reactors are reactors whose active surface areas are orders of magnitude higher than those of micro-reactors but have low fabrication costs relative to micro-reactors. We successfully produced biodiesel with different degrees of conversion using homogeneous catalyst in the slit-channel reactor. The reactor performance shows that percent conversion of soybean oil to biodiesel increases with channel depth, as expected, due to more efficient mixing. Shallow slit-channels require short average residence times for complete product conversion. Present results show that the slit-channel reactor provides an improved performance over traditional batch reactors using homogeneous sodium alkoxide catalyst. It is aimed to couple the reactors with solid catalysts in converting soybean oil to biodiesel and implementation method is suggested. The cost advantages resulting from the ease of fabrication of slit-channel reactors over micro-reactors and how these factors relate to the oil conversion efficiency to biodiesel are briefly noted and discussed.

  14. Effect of welding conditions on transformation and properties of heat-affected zones in LWR (light-water reactor) vessel steels

    SciTech Connect

    Lundin, C.D.; Mohammed, S. . Welding Research and Engineering)

    1990-11-01

    The continuous cooling transformation behavior (CCT) and isothermal transformation (IT) behavior were determined for SA-508 and SA-533 materials for conditions pertaining to standard heat treatment and for the coarse-grained region of the heat-affected zone (HAZ). The resulting diagrams help to select welding conditions that produce the most favorable microconstituent for the development of optimum postweld heat treatment (PWHT) toughness levels. In the case of SA-508 and SA-533, martensite responds more favorably to PWHT than does bainite. Bainite is to be avoided for the optimum toughness characteristics of the HAZ. The reheat cracking tendency for both steels was evaluated by metallographic studies of simulated HAZ structures subjected to PWHT cycles and simultaneous restraint. Both SA-533, Grade B, Class 1, and SA-508, Class 2, cracked intergranularly. The stress rupture parameter (the product of the stress for a rupture life of 10 min and the corresponding reduction of area) calculated for both steels showed that SA-508, Class 2, was more susceptible to reheat cracking than SA-533, Grade B, Class 1. Cold cracking tests (Battelle Test and University of Tennessee modified hydrogen susceptibility test) indicated that a higher preheat temperature is required for SA-508, Class 2, to avoid cracking than is required for SA-533, Grade B, Class 1. Further, the Hydrogen Susceptibility Test showed that SA-508, Class 2, is more susceptible to hydrogen embrittlement than is SA-533, Grade B, Class 1.

  15. Radioisotope research, production, and processing at the University of Missouri Research Reactor

    SciTech Connect

    Ehrhardt, G.J.; Ketring, A.R.; Ja, Wei; Ma, D.; Zinn, K.; Lanigan, J.

    1995-12-31

    The University of Missouri Research Reactor (MURR) is a 10 MW, light-water-cooled and moderated research reactor which first achieved criticality in 1996 and is currently the highest powered university-owned research reactor in the U.S. For many years a major supplier of reactor-produced isotopes for research and commercial purposes, in the last 15 years MURR has concentrated on development of reactor-produced beta-particle emitters for experimental use in nuclear medicine therapy of cancer and rheumatoid arthritis. MURR has played a major role in the development of bone cancer pain palliation with the agents {sup 153}Sm EDTMP and {sup 186}Re/{sup 188}Re HEDP, as well as in the use of {sup 186}Re, {sup 177}Lu, {sup 166}Ho, and {sup 105}Rh for radioimmunotherapy and receptor-agent-guided radiotherapy. MURR is also responsible for the development of therapeutic, {sup 90}Y-labeled glass microspheres for the treatment of liver tumors, a product ({sup 90}Y Therasphere{trademark}) which is currently an approved drug in Canada. MURR has also pioneered the development of {sup 188}W/{sup 188}Re and {sup 99}Mo/{sup 99m}Tc gel generators, which make the use of low specific activity {sup 188}W and {sup 99}Mo practical for such isotope generators.

  16. Light water detritiation

    SciTech Connect

    Fedorchenko, O.A.; Aleksee, I.A.; Bondarenko, S.D.; Vasyanina, T.V.

    2015-03-15

    Hundreds of thousands of tons of tritiated light water have been accumulating from the enterprises of nuclear fuel cycles around the world. The Dual-Temperature Water-Hydrogen (DTWH) process looks like the only practical alternative to Combined Electrolysis and Catalytic Exchange (CECE). In DTWH power-consuming lower reflux device (electrolytic cell) is replaced by a so-called 'hot tower' (LPCE column operating at conditions which ensure relatively small value of elementary separation factor α(hot)). In the upper, cold tower, the tritium transfers from hydrogen to water while in the lower, hot tower - in the opposite direction - from water to hydrogen. The DTWH process is much more complicated compared to CECE; it must be thoroughly computed and strictly controlled by an automatic control system. The use of a simulation code for DTWH is absolutely important. The simulation code EVIO-5 deals with 3 flows inside a column (hydrogen gas, water vapour and liquid water) and 2 simultaneous isotope exchange sub-processes (counter-current phase exchange and co-current catalytic exchange). EVIO-5 takes into account the strong dependence of process performance on given conditions (temperature and pressure). It calculates steady-state isotope concentration profiles considering a full set of reversible exchange reactions between different isotope modifications of water and hydrogen (12 molecular species). So the code can be used for simulation of LPCE column operation for detritiation of hydrogen and water feed, which contains H and D not only at low concentrations but above 10 at.% also. EVIO-5 code is used to model a Tritium Removal Facility with a throughput capacity of about 400 m{sup 3}/day. Simulation results show that a huge amount of wet-proofed catalyst is required (about 6000 m{sup 3}), mainly (90%) in the first stage. One reason for these large expenses (apart from a big scale of the problem itself) is the relatively high tritium separation factor in the hot tower

  17. Robustness of RISMC Insights under Alternative Aleatory/Epistemic Uncertainty Classifications: Draft Report under the Risk-Informed Safety Margin Characterization (RISMC) Pathway of the DOE Light Water Reactor Sustainability Program

    SciTech Connect

    Unwin, Stephen D.; Eslinger, Paul W.; Johnson, Kenneth I.

    2012-09-20

    The Risk-Informed Safety Margin Characterization (RISMC) pathway is a set of activities defined under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability Program. The overarching objective of RISMC is to support plant life-extension decision-making by providing a state-of-knowledge characterization of safety margins in key systems, structures, and components (SSCs). A technical challenge at the core of this effort is to establish the conceptual and technical feasibility of analyzing safety margin in a risk-informed way, which, unlike conventionally defined deterministic margin analysis, would be founded on probabilistic characterizations of uncertainty in SSC performance. In the context of probabilistic risk assessment (PRA) technology, there has arisen a general consensus about the distinctive roles of two types of uncertainty: aleatory and epistemic, where the former represents irreducible, random variability inherent in a system, whereas the latter represents a state of knowledge uncertainty on the part of the analyst about the system which is, in principle, reducible through further research. While there is often some ambiguity about how any one contributing uncertainty in an analysis should be classified, there has nevertheless emerged a broad consensus on the meanings of these uncertainty types in the PRA setting. However, while RISMC methodology shares some features with conventional PRA, it will nevertheless be a distinctive methodology set. Therefore, the paradigms for classification of uncertainty in the PRA setting may not fully port to the RISMC environment. Yet the notion of risk-informed margin is based on the characterization of uncertainty, and it is therefore critical to establish a common understanding of uncertainty in the RISMC setting.

  18. Criticality Studies of Graphite Moderated Production Reactors

    DTIC Science & Technology

    1980-01-01

    eu~ i~ a i5Report) 21 o ages Release unlimited 2.fciri ~i ae 2 rc (See ANSI-Z39. 1S) See Instructions on Reverse OPTMOAL PORN 272 (4-47) (Formerly...E 26 tl ,,-4 26 0 24.5 . 1.65 .7 5.75 1.80 Graphit Denltl 1in/cm Fig 6. 11 Vaito in opiu fue cell pic wit grpht denity for wae-coe reactor.IIl...Powdered Pitch BinderI l " Green Product FeedFI0MOLDING OR EXTRUSION - Green hapes 󈧄 l" lCALCINING-/ Heat 1400OF CARBON RODUCT Pic SokiT

  19. (Fission product transport processes in reactor accidents)

    SciTech Connect

    Hodge, S.A.; Beahm, E.C.; Kress, T.S.; Malinauskas, A.P.

    1989-06-14

    The purpose of this trip was to participate in and to hold informal discussions with other participants in the International Centre for Heat and Mass Transfer (ICHMT) International Seminar on Fission Product Transport Processes held at Dubrovnik, Yugoslavia, during the week of May 22--26, 1989. There were 129 participants from 20 countries at the Seminar. The travelers delivered two invited lectures and presented four invited papers based upon NRC-sponsored work at Oak Ridge National Laboratory. One of the travelers also served as Chairman of the Session entitled Transport Phenomena in the Reactor Coolant System'' and appeared as a Panelist in the Closing Session of the Seminar.

  20. Effect of anaerobic reactor process configuration on useful energy production.

    PubMed

    DiStefano, Thomas D; Palomar, Albert

    2010-04-01

    The effect of reactor process configuration on anaerobic production of useful energy (hydrogen and methane) from a complex substrate was investigated for the following reactor systems: suspended growth, two-phase mixed, two-stage mixed, upflow anaerobic sludge blanket (UASB) reactor, and two-phase UASB. The mixed two-phase and two-stage configurations yielded the highest specific energy productions of 13.3 and 13.4 kJ/g COD fed, respectively. Reactor process configuration influenced microbial pathways in acidogenic reactors in that butyrate was the predominant volatile acid in phased configurations, whereas acetate was predominant in the staged configuration. The UASB reactor achieved the highest average daily energy production per reactor volume of 101 kJ/L reactor-d. All reactor configurations achieved high COD removals on the order of 99%. However, hydrogen represented only 3% of the total energy produced by the two-phase mixed and two-phase UASB configurations. Theoretical analysis revealed that the maximum specific energy production by the two-phase suspended-growth configuration is only 9% higher than that for a single-stage mixed reactor. Consequently, the production of hydrogen from complex substrates in these process configurations does not seem to be justifiable solely from an energy point of view. Instead, it is suggested that phased anaerobic systems should be considered primarily for improved process stability whereas resultant hydrogen production is of secondary benefit.

  1. Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 1, Summary

    SciTech Connect

    Not Available

    1991-04-01

    This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public.

  2. Draft environmental impact statement siting, construction, and operation of New Production Reactor capacity. Volume 4, Appendices D-R

    SciTech Connect

    1991-04-01

    This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains 15 appendices.

  3. NPR (New Production Reactor) capacity cost evaluation

    SciTech Connect

    1988-07-01

    The ORNL Cost Evaluation Technical Support Group (CETSG) has been assigned by DOE-HQ Defense Programs (DP) the task defining, obtaining, and evaluating the capital and life-cycle costs for each of the technology/proponent/site/revenue possibilities envisioned for the New Production Reactor (NPR). The first part of this exercise is largely one of accounting, since all NPR proponents use different accounting methodologies in preparing their costs. In order to address this problem of comparing ''apples and oranges,'' the proponent-provided costs must be partitioned into a framework suitable for all proponents and concepts. If this is done, major cost categories can then be compared between concepts and major cost differences identified. Since the technologies proposed for the NPR and its needed fuel and target support facilities vary considerably in level of technical and operational maturity, considerable care must be taken to evaluate the proponent-derived costs in an equitable manner. The use of cost-risk analysis along with derivation of single point or deterministic estimates allows one to take into account these very real differences in technical and operational maturity. Chapter 2 summarizes the results of this study in tabular and bar graph form. The remaining chapters discuss each generic reactor type as follows: Chapter 3, LWR concepts (SWR and WNP-1); Chapter 4, HWR concepts; Chapter 5, HTGR concept; and Chapter 6, LMR concept. Each of these chapters could be a stand-alone report. 39 refs., 36 figs., 115 tabs.

  4. POTENTIAL BENCHMARKS FOR ACTINIDE PRODUCTION IN HANFORD REACTORS

    SciTech Connect

    PUIGH RJ; TOFFER H

    2011-10-19

    A significant experimental program was conducted in the early Hanford reactors to understand the reactor production of actinides. These experiments were conducted with sufficient rigor, in some cases, to provide useful information that can be utilized today in development of benchmark experiments that may be used for the validation of present computer codes for the production of these actinides in low enriched uranium fuel.

  5. IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

    SciTech Connect

    Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci

    2015-07-01

    IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

  6. The effective management of medical isotope production in research reactors

    SciTech Connect

    Drummond, D.T. )

    1993-01-01

    During the 50-yr history of the use of radioisotopes for medical applications, research reactors have played a pivotal role in the production of many if not most of the key products. The marriage between research reactors and production operations is subject to significant challenges on two fronts. The medical applications of the radioisotope products impose some unique constraints and requirements on the production process. In addition, the mandates and priorities of a research reactor are not always congruent with the demands of a production environment. This paper briefly reviews the historical development of medical isotope production, identifies the unique challenges facing this endeavor, and discusses the management of the relationship between the isotope producer and the research reactor operator. Finally, the key elements of a successful relationship are identified.

  7. Aerosol reactor production of uniform submicron powders

    DOEpatents

    Flagan, Richard C.; Wu, Jin J.

    1991-02-19

    A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

  8. Aerosol reactor production of uniform submicron powders

    NASA Technical Reports Server (NTRS)

    Flagan, Richard C. (Inventor); Wu, Jin J. (Inventor)

    1991-01-01

    A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

  9. Supplying the nuclear arsenal: American production reactors, 1942--1992

    SciTech Connect

    Carlisle, R.P.; Zenzen, J.M.

    1996-01-01

    Although the history of commercial-power nuclear reactors is well known, the story of the government reactors that produce weapons-grade plutonium and tritium has been shrouded in secrecy. In the first detailed look at the origin and development of these production reactors, the authors describe a fifty-year government effort no less complex, expensive, and technologically demanding than the Polaris or Apollo programs--yet one about which most Americans know virtually nothing. The book describes the evolution of the early reactors, the atomic weapons establishment that surrounded them, and the sometimes bitter struggles between business and political constituencies for their share of 'nuclear pork.' They show how, since the 1980s, aging production reactors have increased the risk of radioactive contamination of the atmosphere and water table. And they describe how the Department of Energy mounted a massive effort to find the right design for a new generation of reactors, only to abandon that effort with the end of the Cold War. Today, all American production reactors remain closed. Due to short half-life, the nation's supply of tritium, crucial to modern weapons, is rapidly dwindling. As countries like Iraq and North Korea threaten to join the nuclear club, the authors contend, the United States needs to revitalize tritium production capacity in order to maintain a viable nuclear deterrent. Meanwhile, as slowly decaying artifacts of the Cold War, the closed production reactors at Hanford, Washington, and Savannah River, South Carolina, loom ominously over the landscape.

  10. Continuous production of tritium in an isotope-production reactor with a separate circulation system

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

  11. Production capabilities in US nuclear reactors for medical radioisotopes

    SciTech Connect

    Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. ); Schenter, R.E. )

    1992-11-01

    The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

  12. Survey of Dust Production in Pebble Bed Reactors Cores

    SciTech Connect

    Joshua J. Cogliati; Abderafi M. Ougouag; Javier Ortensi

    2011-06-01

    Graphite dust produced via mechanical wear from the pebbles in a pebble bed reactor is an area of concern for licensing. Both the German pebble bed reactors produced graphite dust that contained activated elements. These activation products constitute an additional source term of radiation and must be taken under consideration during the conduct of accident analysis of the design. This paper discusses the available literature on graphite dust production and measurements in pebble bed reactors. Limited data is available on the graphite dust produced from the AVR and THTR-300 pebble bed reactors. Experiments that have been performed on wear of graphite in pebble-bed-like conditions are reviewed. The calculation of contact forces, which are a key driving mechanism for dust in the reactor, are also included. In addition, prior graphite dust predictions are examined, and future areas of research are identified.

  13. Homogeneous fast-flux isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

  14. Method of producing gaseous products using a downflow reactor

    SciTech Connect

    Cortright, Randy D; Rozmiarek, Robert T; Hornemann, Charles C

    2014-09-16

    Reactor systems and methods are provided for the catalytic conversion of liquid feedstocks to synthesis gases and other noncondensable gaseous products. The reactor systems include a heat exchange reactor configured to allow the liquid feedstock and gas product to flow concurrently in a downflow direction. The reactor systems and methods are particularly useful for producing hydrogen and light hydrocarbons from biomass-derived oxygenated hydrocarbons using aqueous phase reforming. The generated gases may find used as a fuel source for energy generation via PEM fuel cells, solid-oxide fuel cells, internal combustion engines, or gas turbine gensets, or used in other chemical processes to produce additional products. The gaseous products may also be collected for later use or distribution.

  15. An immobilized cell reactor with simultaneous product separation. II. Experimental reactor performance.

    PubMed

    Dale, M C; Okos, M R; Wankat, P C

    1985-07-01

    The simultaneous separation of volatile fermentation products from product-inhibited fermentations can greatly increase the productivity of a bioreactor by reducing the product concentration in the bioreactor, as well as concentrating the product in an output stream free of cells, substrate, or other feed impurities. The Immobilized Cell Reactor-Separator (ICRS) consists of two column reactors: a cocurrent gas-liquid "enricher" followed by a countercurrent "stripper" The columns are four-phase tubular reactors consisting of (1) an inert gas phase, (2) the liquid fermentation broth, (3) the solid column internal packing, and (4) the immobilized biological catalyst or cells. The application of the ICRS to the ethanol-from-whey-lactose fermentation system has been investigated. Operation in the liquid continuous or bubble flow regime allows a high liquid holdup in the reactor and consequent long and controllable liquid residence time but results in a high gas phase pressure drop over the length of the reactor and low gas flow rates. Operation in the gas continuous regime gives high gas flow rates and low pressure drop but also results in short liquid residence time and incomplete column wetting at low liquid loading rates using conventional gas-liquid column packings. Using cells absorbed to conventional ceramic column packing (0.25-in. Intalox saddles), it was found that a good reaction could be obtained in the liquid continuous mode, but little separation, while in the gas continuous mode there was little reaction but good separation. Using cells sorbed to an absorbant matrix allowed operation in the gas continuous regime with a liquid holdup of up to 30% of the total reactor volume. Good reaction rates and product separation were obtained using this matrix. High reaction rates were obtained due to high density cell loading in the reactor. A dry cell density of up to 92 g/L reactor was obtained in the enricher. The enricher ethanol productivity ranged from 50 to 160

  16. Chemistry of fission product iodine under nuclear reactor accident conditions

    SciTech Connect

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs.

  17. Silicon production in an aerosol reactor

    NASA Technical Reports Server (NTRS)

    Wu, J. J.; Alam, M. K.; Johnson, B. E.; Flagan, R. C.

    1984-01-01

    An aerosol reactor for the growth of large silicon particles by silane pyrolysis was shown to demonstrate the following properties: (1) generate seed particles by pyrolysis of a small amount of silane; (2) mix seed aerosol with primary silane flow, limiting number concentration such that the amount of silane is sufficient to grow the desired size of particles from the seed; and (3) react the silane at a rate which is controlled such that the seed particles scavenge the condensible vapors rapidly enough to inhibit further nucleation.

  18. Improving Jet Reactor Configuration for Production of Carbon Nanotubes

    NASA Technical Reports Server (NTRS)

    Povitsky, Alex

    2000-01-01

    The jet mixing reactor has been proposed for the industrial production of fullerene carbon nanotubes. Here we study the flowfield of this reactor using the SIMPLER algorithm. Hot peripheral jets are used to enhance heating of the central jet by mixing with the ambiance of reactor. Numerous configurations of peripheral jets with various number of jets, distance between nozzles, angles between the central jet and a peripheral jets, and twisted configuration of nozzles are considered. Unlike the previous studies of jet mixing, the optimal configuration of peripheral jets produces strong non-uniformity of the central jet in a cross-section. The geometrical shape of reactor is designed to obtain a uniform temperature of a catalyst.

  19. Moving bed reactor for solar thermochemical fuel production

    DOEpatents

    Ermanoski, Ivan

    2013-04-16

    Reactors and methods for solar thermochemical reactions are disclosed. Embodiments of reactors include at least two distinct reactor chambers between which there is at least a pressure differential. In embodiments, reactive particles are exchanged between chambers during a reaction cycle to thermally reduce the particles at first conditions and oxidize the particles at second conditions to produce chemical work from heat. In embodiments, chambers of a reactor are coupled to a heat exchanger to pre-heat the reactive particles prior to direct exposure to thermal energy with heat transferred from reduced reactive particles as the particles are oppositely conveyed between the thermal reduction chamber and the fuel production chamber. In an embodiment, particle conveyance is in part provided by an elevator which may further function as a heat exchanger.

  20. Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 3, Sections 7-12, Appendices A-C

    SciTech Connect

    Not Available

    1991-04-01

    This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains references; a list of preparers and recipients; acronyms, abbreviations, and units of measure; a glossary; an index and three appendices.

  1. Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 2, Sections 1-6

    SciTech Connect

    Not Available

    1991-04-01

    This (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains the analysis of programmatic alternatives, project alternatives, affected environment of alternative sites, environmental consequences, and environmental regulations and permit requirements.

  2. Hanford production reactor heat releases 1951--1971

    SciTech Connect

    Kannberg, L.D.

    1992-04-01

    The purpose of this report is to document and detail the thermal releases from the Hanford nuclear production reactors during the period 1951 through 1971, and to put these releases in historical perspective with respect to changing Columbia River flows and temperatures. This information can also be used as a foundation for further ecological evaluations. When examining Hanford production reactor thermal releases to the Columbia River all related factors affecting the releases and the characteristics of the river should be considered. The major considerations in the present study were the characteristics of the releases themselves (primarily coolant flow rate, temperatures, discharge facilities, period of operation, and level of operation) and the characteristics of the river in that reach (primarily flow rate, temperature and mixing characteristics; the effects of dam construction were also taken into account). In addition, this study addressed ecological effects of thermal releases on aquatic species. Accordingly, this report includes discussion of the reactor cooling system, historical heat releases, thermal mixing and transport studies, hydroelectric power development, and ecologic effects of Hanford production reactor heat releases on salmon and trout. Appendix A contains reactor operating statistics, and Appendix B provide computations of heat added to the Columbia River between Priest Rapids Dam and Richland, Washington.

  3. Transformation products of clindamycin in moving bed biofilm reactor (MBBR).

    PubMed

    Ooi, Gordon T H; Escola Casas, Monica; Andersen, Henrik R; Bester, Kai

    2017-04-15

    Clindamycin is widely prescribed for its ability to treat a number of common bacterial infections. Thus, clindamycin enters wastewater via human excretion or disposal of unused medication and widespread detection of pharmaceuticals in rivers proves the insufficiency of conventional wastewater treatment plants in removing clindamycin. Recently, it has been discovered that attached biofilm reactors, e.g., moving bed biofilm reactors (MBBRs) obtain a higher removal of pharmaceuticals than conventional sludge wastewater treatment plants. Therefore, this study investigated the capability of MBBRs applied in the effluent of conventional wastewater treatment plants to remove clindamycin. First, a batch experiment was executed with a high initial concentration of clindamycin to identify the transformation products. It was shown that clindamycin can be removed from wastewater by MBBR and the treatment process converts clindamycin into the, possibly persistent, products clindamycin sulfoxide and N-desmethyl clindamycin as well as 3 other mono-oxygenated products. Subsequently, the removal kinetics of clindamycin and the formation of the two identified products were investigated in batch experiments using MBBR carriers from polishing and nitrifying reactors. Additionally, the presence of these two metabolites in biofilm-free wastewater effluent was studied. The nitrifying biofilm reactor had a higher biological activity with k-value of 0.1813 h(-1) than the reactor with polishing biofilm (k = 0.0161 h(-1)) which again has a much higher biological activity for removal of clindamycin than of the suspended bacteria (biofilm-free control). Clindamycin sulfoxide was the main transformation product which was found in concentrations exceeding 10% of the initial clindamycin concentration after 1 day of MBBR treatment. Thus, MBBRs should not necessarily be considered as reactors mineralizing clindamycin as they perform transformation reactions at least to some extent.

  4. NOVEL REACTOR FOR THE PRODUCTION OF SYNTHESIS GAS

    SciTech Connect

    Vasilis Papavassiliou; Leo Bonnell; Dion Vlachos

    2004-12-01

    Praxair investigated an advanced technology for producing synthesis gas from natural gas and oxygen This production process combined the use of a short-reaction time catalyst with Praxair's gas mixing technology to provide a novel reactor system. The program achieved all of the milestones contained in the development plan for Phase I. We were able to develop a reactor configuration that was able to operate at high pressures (up to 19atm). This new reactor technology was used as the basis for a new process for the conversion of natural gas to liquid products (Gas to Liquids or GTL). Economic analysis indicated that the new process could provide a 8-10% cost advantage over conventional technology. The economic prediction although favorable was not encouraging enough for a high risk program like this. Praxair decided to terminate development.

  5. Thermal reactor. [liquid silicon production from silane gas

    NASA Technical Reports Server (NTRS)

    Levin, H.; Ford, L. B. (Inventor)

    1982-01-01

    A thermal reactor apparatus and method of pyrolyticaly decomposing silane gas into liquid silicon product and hydrogen by-product gas is disclosed. The thermal reactor has a reaction chamber which is heated well above the decomposition temperature of silane. An injector probe introduces the silane gas tangentially into the reaction chamber to form a first, outer, forwardly moving vortex containing the liquid silicon product and a second, inner, rewardly moving vortex containing the by-product hydrogen gas. The liquid silicon in the first outer vortex deposits onto the interior walls of the reaction chamber to form an equilibrium skull layer which flows to the forward or bottom end of the reaction chamber where it is removed. The by-product hydrogen gas in the second inner vortex is removed from the top or rear of the reaction chamber by a vortex finder. The injector probe which introduces the silane gas into the reaction chamber is continually cooled by a cooling jacket.

  6. Light Water Reactor Sustainability Program BWR High-Fluence Material Project: Assessment of the Role of High-Fluence on the Efficiency of HWC Mitigation on SCC Crack Growth Rates

    SciTech Connect

    Sebastien Teysseyre

    2014-04-01

    As nuclear power plants age, the increasing neutron fluence experienced by stainless steels components affects the materials resistance to stress corrosion cracking and fracture toughness. The purpose of this report is to identify any new issues that are expected to rise as boiling water reactor power plants reach the end of their initial life and to propose a path forward to study such issues. It has been identified that the efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report summarizes the data available to support this hypothesis and describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. This program plan includes acquisition of irradiated materials, generation of material via irradiation in a test reactor, and description of the test plan. This plan offers three approaches, each with an estimated timetable and budget.

  7. INEL advanced test reactor plutonium-238 production feasibility assessment

    SciTech Connect

    Schnitzler, B.G. )

    1993-01-10

    Results of a preliminary neutronics assessment indicate the feasibility of [sup 238]Pu production in the Idaho National Engineering Laboratory Advanced Test Reactor (ATR). Based on the results of this assessment, an annual production of 11.3 kg [sup 238]Pu can be achieved in the ATR. An annual loading of 102 kg [sup 237]Np is required for the particular target configuration and irradiation scenario examined. The [sup 236]Pu contaminant level is approximately 6 parts per million at zero cooling time. The product quality is about 90% [sup 238]Pu. Neptunium feedstock requirements, [sup 238]Pu production rates, or product purity can be optimized depending on their relative importances.

  8. INEL advanced test reactor plutonium-238 production feasibility assessment

    NASA Astrophysics Data System (ADS)

    Schnitzler, Bruce G.

    1993-01-01

    Results of a preliminary neutronics assessment indicate the feasibility of 238Pu production in the Idaho National Engineering Laboratory Advanced Test Reactor (ATR). Based on the results of this assessment, an annual production of 11.3 kg 238Pu can be achieved in the ATR. An annual loading of 102 kg 237Np is required for the particular target configuration and irradiation scenario examined. The 236Pu contaminant level is approximately 6 parts per million at zero cooling time. The product quality is about 90% 238Pu. Neptunium feedstock requirements, 238Pu production rates, or product purity can be optimized depending on their relative importances.

  9. A microBio reactor for hydrogen production.

    SciTech Connect

    Volponi, Joanne V.; Walker, Andrew William

    2003-12-01

    The purpose of this work was to explore the potential of developing a microfluidic reactor capable of enzymatically converting glucose and other carbohydrates to hydrogen. This aggressive project was motivated by work in enzymatic hydrogen production done by Woodward et al. at OWL. The work reported here demonstrated that hydrogen could be produced from the enzymatic oxidation of glucose. Attempts at immobilizing the enzymes resulted in reduced hydrogen production rates, probably due to buffer compatibility issues. A novel in-line sensor was also developed to monitor hydrogen production in real time at levels below 1 ppm. Finally, a theoretical design for the microfluidic reactor was developed but never produced due to the low production rates of hydrogen from the immobilized enzymes. However, this work demonstrated the potential of mimicking biological systems to create energy on the microscale.

  10. A proposed standard on medical isotope production in fission reactors

    SciTech Connect

    Schenter, R. E.; Brown, G. J.; Holden, C. S.

    2006-07-01

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  11. ``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

    SciTech Connect

    Jerde, E.A.; Glasgow, D.C.

    1998-09-01

    At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.

  12. Hydrogen Production via a Commercially Ready Inorganic membrane Reactor

    SciTech Connect

    Paul K.T. Liu

    2005-08-23

    Single stage low-temperature-shift water-gas-shift (WGS-LTS) via a membrane reactor (MR) process was studied through both mathematical simulation and experimental verification in this quarter. Our proposed MR yields a reactor size that is 10 to >55% smaller than the comparable conventional reactor for a CO conversion of 80 to 90%. In addition, the CO contaminant level in the hydrogen produced via MR ranges from 1,000 to 4,000 ppm vs 40,000 to >70,000 ppm via the conventional reactor. The advantages of the reduced WGS reactor size and the reduced CO contaminant level provide an excellent opportunity for intensification of the hydrogen production process by the proposed MR. To prepare for the field test planned in Yr III, a significant number (i.e., 98) of full-scale membrane tubes have been produced with an on-spec ratio of >76% during this first production trial. In addition, an innovative full-scale membrane module has been designed, which can potentially deliver >20 to 30 m{sup 2}/module making it suitable for large-scale applications, such as power generation. Finally, we have verified our membrane performance and stability in a refinery pilot testing facility on a hydrocracker purge gas. No change in membrane performance was noted over the >100 hrs of testing conducted in the presence of >30% H{sub 2}S, >5,000 ppm NH{sub 3} (estimated), and heavy hydrocarbons on the order of 25%. The high stability of these membranes opens the door for the use of our membrane in the WGS environment with significantly reduced pretreatment burden.

  13. Mass production of magnetic nickel nanoparticle in thermal plasma reactor

    SciTech Connect

    Kanhe, Nilesh S.; Nawale, Ashok B.; Bhoraskar, S. V.; Mathe, V. L.; Das, A. K.

    2014-04-24

    We report the mass production of Ni metal nanoparticles using dc transferred arc thermal plasma reactor by homogeneous gas phase condensation process. To increase the evaporation rate and purity of Ni nanoparticles small amount of hydrogen added along with argon in the plasma. Crystal structure analysis was done by using X-ray diffraction technique. The morphology of as synthesized nanoparticles was carried out using FESEM images. The magnetic properties were measured by using vibrating sample magnetometer at room temperature.

  14. USE OF THE MODULAR HELIUM REACTOR FOR HYDROGEN PRODUCTION

    SciTech Connect

    SCHULTZ,KR

    2003-09-01

    OAK-B135 A significant ''Hydrogen Economy'' is predicted that will reduce our dependence on petroleum imports and reduce pollution and greenhouse gas emissions. Hydrogen is an environmentally attractive fuel that has the potential to displace fossil fuels, but contemporary hydrogen production is primarily based on fossil fuels. The author has recently completed a three-year project for the US Department of Energy (DOE) whose objective was to ''define an economically feasible concept for production of hydrogen, using an advanced high-temperature nuclear reactor as the energy source''. Thermochemical water-slitting, a chemical process that accomplishes the decomposition of water into hydrogen and oxygen, met this objective. The goal of the first phase of this study was to evaluate thermochemical processes which offer the potential for efficient, cost-effective, large-scale production of hydrogen, and to select one for further detailed consideration. They selected the Sulfur-Iodine cycle. In the second phase, they reviewed all the basic reactor types for suitability to provide the high temperature heat needed by the selected thermochemical water splitting cycle and chose the helium gas-cooled reactor. In the third phase they designed the chemical flowsheet for the thermochemical process and estimated the efficiency and cost of the process and the projected cost of producing hydrogen. These results are summarized in this report.

  15. Reactor design for minimizing product inhibition during enzymatic lignocellulose hydrolysis: II. Quantification of inhibition and suitability of membrane reactors.

    PubMed

    Andrić, Pavle; Meyer, Anne S; Jensen, Peter A; Dam-Johansen, Kim

    2010-01-01

    Product inhibition of cellulolytic enzymes affects the efficiency of the biocatalytic conversion of lignocellulosic biomass to ethanol and other valuable products. New strategies that focus on reactor designs encompassing product removal, notably glucose removal, during enzymatic cellulose conversion are required for alleviation of glucose product inhibition. Supported by numerous calculations this review assesses the quantitative aspects of glucose product inhibition on enzyme-catalyzed cellulose degradation rates. The significance of glucose product inhibition on dimensioning of different ideal reactor types, i.e. batch, continuous stirred, and plug-flow, is illustrated quantitatively by modeling different extents of cellulose conversion at different reaction conditions. The main operational challenges of membrane reactors for lignocellulose conversion are highlighted. Key membrane reactor features, including system set-up, dilution rate, glucose output profile, and the problem of cellobiose are examined to illustrate the quantitative significance of the glucose product inhibition and the total glucose concentration on the cellulolytic conversion rate. Comprehensive overviews of the available literature data for glucose removal by membranes and for cellulose enzyme stability in membrane reactors are given. The treatise clearly shows that membrane reactors allowing continuous, complete, glucose removal during enzymatic cellulose hydrolysis, can provide for both higher cellulose hydrolysis rates and higher enzyme usage efficiency (kg(product)/kg(enzyme)). Current membrane reactor designs are however not feasible for large scale operations. The report emphasizes that the industrial realization of cellulosic ethanol requires more focus on the operational feasibility within the different hydrolysis reactor designs, notably for membrane reactors, to achieve efficient enzyme-catalyzed cellulose degradation.

  16. Process aspects of hydrogen production using the Tandem Mirror Reactor

    SciTech Connect

    Galloway, T.R.

    1980-10-08

    An SO/sub 3/ decomposer design is presented which provides the necessary production of SO/sub 2/ + O/sub 2/ for the thermochemical production of H/sub 2/ using the thermal heat from the blanket of a Tandem Mirror Fusion Reactor. The decomposer is a catalytic, fluidized bed reactor operating at 1050/sup 0/K that contains internal heat exchanger tubes to provide the heat for the highly endothermic SO/sub 3/ decomposition reactor. The conversion per pass is around 55% using either CuO or platinum-based catalysts. We examine the impact on the design of varying the temperature 100/sup 0/C higher or lower than 1050/sup 0/K and we also look at both liquid sodium and helium as the heat transfer medium to carry the heat from the TMR to the decomposer. To drive a H/sub 2/ plant with 15 x 10/sup 6/ sm/sup 3//d there would be around 7 decomposers, 10 m in height and 6 m in diameter, made of Incoloy-800H at $24/kg at a total capital cost of around $50 million - which translates to about 12 cents/GJ of H/sub 2/ produced with conventional economics. Helium-driven units would approximately double the cost in trade for the additional safety isolation.

  17. Preconceptual design of the new production reactor circulator test facility

    SciTech Connect

    Thurston, G.

    1990-06-01

    This report presents the results of a study of a new circulator test facility for the New Production Reactor Modular High-Temperature Gas-Cooled Reactor. The report addresses the preconceptual design of a stand-alone test facility with all the required equipment to test the Main Circulator/shutoff valve and Shutdown Cooling Circulator/shutoff valve. Each type of circulator will be tested in its own full flow, full power helium test loop. Testing will cover the entire operating range of each unit. The loop will include a test vessel, in which the circulator/valve will be mounted, and external piping. The external flow piping will include a throttle valve, flowmeter, and heat exchanger. Subsystems will include helium handling, helium purification, and cooling water. A computer-based data acquisition and control system will be provided. The estimated costs for the design and construction of this facility are included. 2 refs., 15 figs.

  18. Westinghouse independent safety review of Savannah River production reactors

    SciTech Connect

    Leggett, W.D.; McShane, W.J. ); Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E. . Nuclear and Advanced Technology Div.); Toto, G. . Nuclear Services Div.); Fauske, H.K. ); Call, D.W. (Westinghouse Savannah R

    1989-04-01

    Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K,L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone.

  19. Production of transplutonium elements in the high flux isotope reactor

    SciTech Connect

    Bigelow, J.E.; Corbett, B.L.; King, L.J.; McGuire, S.C.; Sims, T.M.

    1981-01-01

    The techniques described here have been demonstrated to predict the contents of transplutonium element production targets, at least for isotopes of mass 253 or less. The HFIR irradiation model is a workhorse for planning the TRU processing campaigns, for certifying the heat evolution rate of targets prior to insertion in the reactor, for predicting future production capabilities over a multi-year period, and for making optimization studies. Practical considerations, however, may limit the range of available options so that optimum operation is not always achievable. We do intend, however, to keep fine-tuning the constants which define the cross sections as time permits. We need to do more work on optimizing the production of /sup 250/Cm, /sup 254/Es, /sup 255/Es, and ultimately /sup 257/Fm, since researchers are interested in obtaining larger quantities of these rare and difficult-to-produce nuclides. 7 figures, 2 tables.

  20. Studies of Plutonium-238 Production at the High Flux Isotope Reactor

    SciTech Connect

    Lastres, Oscar; Chandler, David; Jarrell, Joshua J; Maldonado, G. Ivan

    2011-01-01

    The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is a versatile 85 MW{sub th}, pressurized, light water-cooled and -moderated research reactor. The core consists of two fuel elements, an inner fuel element (IFE) and an outer fuel element (OFE), each constructed of involute fuel plates containing high-enriched-uranium (HEU) fuel ({approx}93 wt% {sup 235}U/U) in the form of U{sub 3}O{sub 8} in an Al matrix and encapsulated in Al-6061 clad. An over-moderated flux trap is located in the center of the core, a large beryllium reflector is located on the outside of the core, and two control elements (CE) are located between the fuel and the reflector. The flux trap and reflector house numerous experimental facilities which are used for isotope production, material irradiation, and cold/thermal neutron scattering. Over the past five decades, the US Department of Energy (DOE) and its agencies have been producing radioisotope power systems used by the National Aeronautics and Space Administration (NASA) for unmanned, long-term space exploration missions. Plutonium-238 is used to power Radioisotope Thermoelectric Generators (RTG) because it has a very long half-life (t{sub 1/2} {approx} 89 yr.) and it generates about 0.5 watts/gram when it decays via alpha emission. Due to the recent shortage and uncertainty of future production, the DOE has proposed a plan to the US Congress to produce {sup 238}Pu by irradiating {sup 237}Np as early as in fiscal year 2011. An annual production rate of 1.5 to 2.0 kg of {sup 238}Pu is expected to satisfy these needs and could be produced in existing national nuclear facilities like HFIR and the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Reactors at the Savannah River Site were used in the past for {sup 238}Pu production but were shut down after the last production in 1988. The nation's {sup 237}Np inventory is currently stored at INL. A plan for producing {sup 238}Pu at US research reactor

  1. Bio-hydrogen production from molasses by anaerobic fermentation in continuous stirred tank reactor

    NASA Astrophysics Data System (ADS)

    Han, Wei; Li, Yong-feng; Chen, Hong; Deng, Jie-xuan; Yang, Chuan-ping

    2010-11-01

    A study of bio-hydrogen production was performed in a continuous flow anaerobic fermentation reactor (with an available volume of 5.4 L). The continuous stirred tank reactor (CSTR) for bio-hydrogen production was operated under the organic loading rates (OLR) of 8-32 kg COD/m3 reactor/d (COD: chemical oxygen demand) with molasses as the substrate. The maximum hydrogen production yield of 8.19 L/d was obtained in the reactor with the OLR increased from 8 kg COD/m3 reactor/d to 24 kg COD/m3 d. However, the hydrogen production and volatile fatty acids (VFAs) drastically decreased at an OLR of 32 kg COD/m3 reactor/d. Ethanoi, acetic, butyric and propionic were the main liquid fermentation products with the percentages of 31%, 24%, 20% and 18%, which formed the mixed-type fermentation.

  2. Neutron economic reactivity control system for light water reactors

    DOEpatents

    Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.; Gregurech, Steve

    1989-01-01

    A neutron reactivity control system for a LWBR incorporating a stationary seed-blanket core arrangement. The core arrangement includes a plurality of contiguous hexagonal shaped regions. Each region has a central and a peripheral blanket area juxapositioned an annular seed area. The blanket areas contain thoria fuel rods while the annular seed area includes seed fuel rods and movable thoria shim control rods.

  3. Technology programs in support of advanced light water reactors

    SciTech Connect

    Not Available

    1990-03-01

    Design for Constructability's overall purpose was to identify and address changes in the nuclear industry to restore nuclear energy as an attractive option for new generating capacity. The program stove to meet the following goals related to the future construction of nuclear power plants: reduced costs; assurance of improved quality; and shortened construction schedules. This is Volume 1 of three volumes.

  4. Strategic Plan for Light Water Reactor Research and Development

    SciTech Connect

    2004-02-01

    The purpose of this strategic plan is to establish a framework that will allow the Department of Energy (DOE) and the nuclear power industry to jointly plan the nuclear energy research and development (R&D) agenda important to achieving the Nation's energy goals. This strategic plan has been developed to focus on only those R&D areas that will benefit from a coordinated government/industry effort. Specifically, this plan focuses on safely sustaining and expanding the electricity output from currently operating nuclear power plants and expanding nuclear capacity through the deployment of new plants. By focusing on R&D that addresses the needs of both current and future nuclear plants, DOE and industry will be able to take advantage of the synergism between these two technology areas, thus improving coordination, enhancing efficiency, and further leveraging public and private sector resources. By working together under the framework of this strategic plan, DOE and the nuclear industry reinforce their joint commitment to the future use of nuclear power and the National Energy Policy's goal of expanding its use in the United States. The undersigned believe that a public-private partnership approach is the most efficient and effective way to develop and transfer new technologies to the marketplace to achieve this goal. This Strategic Plan is intended to be a living document that will be updated annually.

  5. Light Water Reactor Sustainability Program Advanced Seismic Soil Structure Modeling

    SciTech Connect

    Bolisetti, Chandrakanth; Coleman, Justin Leigh

    2015-06-01

    Risk calculations should focus on providing best estimate results, and associated insights, for evaluation and decision-making. Specifically, seismic probabilistic risk assessments (SPRAs) are intended to provide best estimates of the various combinations of structural and equipment failures that can lead to a seismic induced core damage event. However, in some instances the current SPRA approach has large uncertainties, and potentially masks other important events (for instance, it was not the seismic motions that caused the Fukushima core melt events, but the tsunami ingress into the facility). SPRA’s are performed by convolving the seismic hazard (this is the estimate of all likely damaging earthquakes at the site of interest) with the seismic fragility (the conditional probability of failure of a structure, system, or component given the occurrence of earthquake ground motion). In this calculation, there are three main pieces to seismic risk quantification, 1) seismic hazard and nuclear power plants (NPPs) response to the hazard, 2) fragility or capacity of structures, systems and components (SSC), and 3) systems analysis. Two areas where NLSSI effects may be important in SPRA calculations are, 1) when calculating in-structure response at the area of interest, and 2) calculation of seismic fragilities (current fragility calculations assume a lognormal distribution for probability of failure of components). Some important effects when using NLSSI in the SPRA calculation process include, 1) gapping and sliding, 2) inclined seismic waves coupled with gapping and sliding of foundations atop soil, 3) inclined seismic waves coupled with gapping and sliding of deeply embedded structures, 4) soil dilatancy, 5) soil liquefaction, 6) surface waves, 7) buoyancy, 8) concrete cracking and 9) seismic isolation The focus of the research task presented here-in is on implementation of NLSSI into the SPRA calculation process when calculating in-structure response at the area of interest. The specific nonlinear soil behavior included in the NLSSI calculation presented in this report is gapping and sliding. Other NLSSI effects are not included in the calculation. The results presented in this report document initial model runs in the linear and nonlinear analysis process. Final comparisons between traditional and advanced SPRA will be presented in the September 30th deliverable.

  6. Study of Intermetallic Nanostructures for Light-Water Reactors

    SciTech Connect

    Jensen, Niels Grobech; Asta, Mark D.; Hosemann, Peter; Maloy, Stuart

    2015-09-30

    High temperature mechanical measurements were conducted to study the effect of the dynamic precipitation process of PH 13-8 Mo maraging steel. Yield stress, ultimate tensile strength, total elongation, hardness, strain rate sensitivity and activation volume were evaluated as a function of the temperature. The dynamic changes in the mechanical properties at different temperatures were evaluated and a balance between precipitation hardening and annealed softening is discussed. A comparison between hardness and yield stress and ultimate tensile strength over a temperature range from 300 to 600 °C is made. The behavior of the strain rate sensitivity was correlated with the intermetallic precipitates formed during the experiments.

  7. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    SciTech Connect

    Dautel, W.A.

    1996-10-01

    The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.

  8. A NOVEL MEMBRANE REACTOR FOR DIRECT HYDROGEN PRODUCTION FROM COAL

    SciTech Connect

    Shain Doong; Estela Ong; Mike Atroshenko; Francis Lau; Mike Roberts

    2004-01-22

    Gas Technology Institute is developing a novel concept of membrane gasifier for high efficiency, clean and low cost production of hydrogen from coal. The concept incorporates a hydrogen-selective membrane within a gasification reactor for direct extraction of hydrogen from coal synthesis gases. The objective of this project is to determine the technical and economic feasibility of this concept by screening, testing and identifying the potential candidate membranes under high temperature, high pressure, and harsh environments of the coal gasification conditions. The best performing membranes will be selected for preliminary reactor design and cost estimates. To evaluate the candidate membrane performance under the gasification conditions, a high temperature/high pressure hydrogen permeation unit will be constructed in this project. During this reporting period, the design of this unit was completed. The unit will be capable of operating at temperatures up to 1100 C and pressures to 60 atm for evaluation of ceramic membranes such as mixed ionic conducting membrane. The membranes to be tested will be in disc form with a diameter of about 3 cm. By operating at higher temperatures and higher hydrogen partial pressures, we expect to demonstrate commercially relevant hydrogen flux, 10 {approx} 50 cc/min/cm{sup 2}, from the membranes made of the perovskite type of ceramic material. The construction of the unit is planned to be completed by the end of the next reporting period.

  9. Flat plate heat exchangers for the new production reactor

    SciTech Connect

    Ondrejcin, R.S.

    1988-12-07

    The New Production Reactor (NPR) will require heat exchangers (HX) as part of the ancillary equipment. The most common type of heat exchanger in the US is the shell and tube, the type presently in SRP reactor service. This type of design is the one that is normally described in detail in heat transfer texts. Other designs are more efficient, and there was a period when these designs were actively modified. The largest driving force for higher efficiency HX was the OPEC produced energy shortage of the early 1970's. Several comments were made by B.S. Spangler about flat plate HX for the NPR after my initial comparison between shell and tube and flat plate HX (DPST-88-729). Since answers were not readily available, Philippe Marchal of Packinox (Louveciennes, France) agreed to visit SRL. This report contains a series of answers to all comments raised in DPST-88-743 and some general statements comparing flat plate HX to shell and tube HX. 1 fig., 2 tabs.

  10. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  11. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    SciTech Connect

    Djurcic, Zelimir; Detwiler, Jason A.; Piepke, Andreas; Foster Jr., Vince R.; Miller, Lester; Gratta, Giorgio

    2008-08-06

    Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in {bar {nu}}{sub e} detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties, and their relevance to reactor {bar {nu}}{sub e} experiments.

  12. Biohydrogen production from tequila vinasses using a fixed bed reactor.

    PubMed

    Buitrón, Germán; Prato-Garcia, Dorian; Zhang, Axue

    2014-01-01

    In Mexico, the industrial production of tequila leads to the discharge of more than 31.2 million of m(3) of vinasse, which causes serious environmental issues because of its acidity, high organic load and the presence of recalcitrant compounds. The aim of this research was to study the feasibility of a fixed bed reactor for the production of biohydrogen by using tequila vinasse as substrate. The experiments were carried out in a continuous mode under mesophilic and acidic conditions. The maximum hydrogen yield and hydrogen production rate were 1.3 mol H2 mol/mol glucose and 72 ± 9 mL H2/(Lreactor h), respectively. Biogas consisted of carbon dioxide (36%) and hydrogen (64%); moreover methane was not observed. The electron-equivalent mass balance fitted satisfactorily (sink of electrons from 0.8 to 7.6%). For vinasses, hydrogen production accounted for 10.9% of the total available electron-equivalents. In the liquid phase, the principal metabolites identified were acetic, butyric and iso-butyric acids, which indicated a butyrate-acetate type fermentation. Tequila vinasses did not result in potential inhibition of the fermentative process. Considering the process as a water treatment system, only 20% of the original carbon was removed (as carbon dioxide and biomass) when the tequila vinasses are used.

  13. Reactor production of sup 252 Cf and transcurium isotopes

    SciTech Connect

    Alexander, C.W.; Halperin, J.; Walker, R.L.; Bigelow, J.E.

    1990-01-01

    Berkelium, californium, einsteinium, and fermium are currently produced in the High Flux Isotope Reactor (HFIR) and recovered in the Radiochemical Engineering Development Center (REDC) at the Oak Ridge National Laboratory (ORNL). All the isotopes are used for research. In addition, {sup 252}Cf, {sup 253}Es, and {sup 255}Fm have been considered or are used for industrial or medical applications. ORNL is the sole producer of these transcurium isotopes in the western world. A wide range of actinide samples were irradiated in special test assemblies at the Fast Flux Test Facility (FFTF) at Hanford, Washington. The purpose of the experiments was to evaluate the usefulness of the two-group flux model for transmutations in the special assemblies with an eventual goal of determining the feasibility of producing macro amounts of transcurium isotopes in the FFTF. Preliminary results from the production of {sup 254g}Es from {sup 252}Cf will be discussed. 14 refs., 5 tabs.

  14. Biobutanol production in a Clostridium acetobutylicum biofilm reactor integrated with simultaneous product recovery by adsorption

    PubMed Central

    2014-01-01

    Background Clostridium acetobutylicum can propagate on fibrous matrices and form biofilms that have improved butanol tolerance and a high fermentation rate and can be repeatedly used. Previously, a novel macroporous resin, KA-I, was synthesized in our laboratory and was demonstrated to be a good adsorbent with high selectivity and capacity for butanol recovery from a model solution. Based on these results, we aimed to develop a process integrating a biofilm reactor with simultaneous product recovery using the KA-I resin to maximize the production efficiency of biobutanol. Results KA-I showed great affinity for butanol and butyrate and could selectively enhance acetoin production at the expense of acetone during the fermentation. The biofilm reactor exhibited high productivity with considerably low broth turbidity during repeated batch fermentations. By maintaining the butanol level above 6.5 g/L in the biofilm reactor, butyrate adsorption by the KA-I resin was effectively reduced. Co-adsorption of acetone by the resin improved the fermentation performance. By redox modulation with methyl viologen (MV), the butanol-acetone ratio and the total product yield increased. An equivalent solvent titer of 96.5 to 130.7 g/L was achieved with a productivity of 1.0 to 1.5 g · L-1 · h-1. The solvent concentration and productivity increased by 4 to 6-fold and 3 to 5-fold, respectively, compared to traditional batch fermentation using planktonic culture. Conclusions Compared to the conventional process, the integrated process dramatically improved the productivity and reduced the energy consumption as well as water usage in biobutanol production. While genetic engineering focuses on strain improvement to enhance butanol production, process development can fully exploit the productivity of a strain and maximize the production efficiency. PMID:24401161

  15. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    SciTech Connect

    Rosenthal, Murray Wilford

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  16. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  17. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  18. International Forum for Reactor Aging Management (IFRAM)

    SciTech Connect

    Bond, Leonard J.

    2010-11-01

    The Nuclear Regulatory Commission has undertaken a program to lay the groundwork for defining proactive actions to manage degradation of materials in light water reactors (LWRs). This article discusses the international forum for reactor aging management.

  19. Activation product safety in the ARIES-I reactor design

    SciTech Connect

    Herring, J.S. ); Sze, D.K. ); Wong, C.; Cheng, E.T. ); Grotz, S.P. )

    1990-01-01

    The ARIES design effort has sought to maximize the environmental and safety advantages of fusion through careful selection of materials and careful design. Three goals are that the reactor achieve inherent or passive safety, that no public evacuation plan be necessary and that the waste be disposable as 10CFR61 Class C waste. The ARIES-I reactor consists of a SiC composite structure for the first wall and blanket, cooled by 10 MPa He. The breeder is Li{sub 2}ZrO{sub 3}, although Li{sub 2}O and Li{sub 4}SiO{sub 4} were also considered. The divertor consists of SiC composite tubes coated with 2 mm of tungsten. Due to the minimal afterheat of this blanket design, LOCA calculations indicate maximum temperatures will not cause damage if the plasma is promptly extinguished. Two primary safety issues are the zirconium in the breeder and tungsten on the divertor. Li{sub 2}ZrO{sub 3} was chosen because of its demonstrated high-temperature stability. The other breeders have lower afterheat and activation. Use of zirconium in the breeder will necessitate isotopic tailoring to remove {sup 90}Zr and {sup 94}Zr. The 5.8 tonnes of W on the divertor would also have to be tailored to remove {sup 186}W and/or to concentrate {sup 183}W. Thus the ARIES-I design achieves the passive safety and low-level waste disposal criteria with respect to activation products. Development of low activation materials to replace zirconium and tungsten is needed to avoid requiring an evacuation plan.

  20. Safety and licensing for small and medium power reactors

    SciTech Connect

    Trauger, D.B.

    1987-01-01

    Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat.

  1. Design basis for protection of light water nuclear power plants against effects of postulated pipe rupture

    SciTech Connect

    Not Available

    1981-01-01

    This standard addresses the design bases for light water reactor, nuclear power plant structures and components essential for the protection of public health and safety from the potential adverse effects of pipe whip, jet impingement, pressurization of compartments outside containment, environmental conditions and flooding associated with a postulated pipe rupture. The design bases for missile protection and the design bases for containment pressurization are not within this standard.

  2. Hydrogen Production via a Commerically Ready Inorganic membrane Reactor

    SciTech Connect

    Paul Liu

    2007-06-30

    It has been known that use of the hydrogen selective membrane as a reactor (MR) could potentially improve the efficiency of the water shift reaction (WGS), one of the least efficient unit operations for production of high purity hydrogen from syngas. However, no membrane reactor technology has been reduced to industrial practice thus far, in particular for a large-scale operation. This implementation and commercialization barrier is attributed to the lack of a commercially viable hydrogen selective membrane with (1) material stability under the application environment and (2) suitability for large-scale operation. Thus, in this project, we have focused on (1) the deposition of the hydrogen selective carbon molecular sieve (CMS) membrane we have developed on commercially available membranes as substrate, and (2) the demonstration of the economic viability of the proposed WGS-MR for hydrogen production from coal-based syngas. The commercial stainless steel (SS) porous substrate (i.e., ZrO{sub 2}/SS from Pall Corp.) was evaluated comprehensively as the 1st choice for the deposition of the CMS membrane for hydrogen separation. The CMS membrane synthesis protocol we developed previously for the ceramic substrate was adapted here for the stainless steel substrate. Unfortunately no successful hydrogen selective membranes had been prepared during Yr I of this project. The characterization results indicated two major sources of defect present in the SS substrate, which may have contributed to the poor CMS membrane quality. Near the end of the project period, an improved batch of the SS substrate (as the 2nd generation product) was received from the supplier. Our characterization results confirm that leaking of the crimp boundary no longer exists. However, the thermal stability of the ZrO{sub 2}/SS substrate through the CMS membrane preparation condition must be re-evaluated in the future. In parallel with the SS membrane activity, the preparation of the CMS membranes

  3. Sensitivity Studies of Advanced Reactors Coupled to High Temperature Electrolysis (HTE) Hydrogen Production Processes

    SciTech Connect

    Edwin A. Harvego; Michael G. McKellar; James E. O'Brien; J. Stephen Herring

    2007-04-01

    High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 °C to 950 °C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the steam or air sweep loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycle producing the highest efficiencies varied depending on the temperature range considered.

  4. Anaerobic biofilm reactors for dark fermentative hydrogen production from wastewater: A review.

    PubMed

    Barca, Cristian; Soric, Audrey; Ranava, David; Giudici-Orticoni, Marie-Thérèse; Ferrasse, Jean-Henry

    2015-06-01

    Dark fermentation is a bioprocess driven by anaerobic bacteria that can produce hydrogen (H2) from organic waste and wastewater. This review analyses a relevant number of recent studies that have investigated dark fermentative H2 production from wastewater using two different types of anaerobic biofilm reactors: anaerobic packed bed reactor (APBR) and anaerobic fluidized bed reactor (AFBR). The effect of various parameters, including temperature, pH, carrier material, inoculum pretreatment, hydraulic retention time, substrate type and concentration, on reactor performances was investigated by a critical discussion of the results published in the literature. Also, this review presents an in-depth study on the influence of the main operating parameters on the metabolic pathways. The aim of this review is to provide to researchers and practitioners in the field of H2 production key elements for the best operation of the reactors. Finally, some perspectives and technical challenges to improve H2 production were proposed.

  5. STAR - H2 : the secure transportable autonomous reactor for hydrogen production and desalinization.

    SciTech Connect

    Wade, D.C.; Doctor, R.; Peddicord, K.L.

    2002-02-26

    The Secure Transportable Autonomous Reactor for Hydrogen production is a modular fast reactor intended for the mid 21st century energy market wherein electricity and hydrogen are employed as complementary energy carriers and nuclear energy contributes to sustainable energy supply based on full transuranic recycle in a passively safe, environmentally friendly and proliferation-resistant manner suitable for widespread worldwide deployment.

  6. Liquid phase methanol reactor staging process for the production of methanol

    SciTech Connect

    Bonnell, Leo W.; Perka, Alan T.; Roberts, George W.

    1988-01-01

    The present invention is a process for the production of methanol from a syngas feed containing carbon monoxide, carbon dioxide and hydrogen. Basically, the process is the combination of two liquid phase methanol reactors into a staging process, such that each reactor is operated to favor a particular reaction mechanism. In the first reactor, the operation is controlled to favor the hydrogenation of carbon monoxide, and in the second reactor, the operation is controlled so as to favor the hydrogenation of carbon dioxide. This staging process results in substantial increases in methanol yield.

  7. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  8. CFD optimization of continuous stirred-tank (CSTR) reactor for biohydrogen production.

    PubMed

    Ding, Jie; Wang, Xu; Zhou, Xue-Fei; Ren, Nan-Qi; Guo, Wan-Qian

    2010-09-01

    There has been little work on the optimal configuration of biohydrogen production reactors. This paper describes three-dimensional computational fluid dynamics (CFD) simulations of gas-liquid flow in a laboratory-scale continuous stirred-tank reactor used for biohydrogen production. To evaluate the role of hydrodynamics in reactor design and optimize the reactor configuration, an optimized impeller design has been constructed and validated with CFD simulations of the normal and optimized impeller over a range of speeds and the numerical results were also validated by examination of residence time distribution. By integrating the CFD simulation with an ethanol-type fermentation process experiment, it was shown that impellers with different type and speed generated different flow patterns, and hence offered different efficiencies for biohydrogen production. The hydrodynamic behavior of the optimized impeller at speeds between 50 and 70 rev/min is most suited for economical biohydrogen production.

  9. Action Plan for updated Chapter 15 Accident Analysis in the SRS Production Reactor SAR

    SciTech Connect

    Hightower, N.T. III; Burnett, T.W.

    1989-11-15

    This report describes the Action Plan for the upgrade of the Chapter 15 Accident Analysis in the SRS Production Reactor SAR required for K-Restart. This Action Plan will be updated periodically to reflect task accomplishments and issue resolutions.

  10. Fuel pins with both target and fuel pellets in an isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

  11. Assemblies with both target and fuel pins in an isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

  12. Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

    SciTech Connect

    Asner, David M.; Burns, Kimberly A.; Campbell, Luke W.; Greenfield, Bryce A.; Kos, Marek S.; Orrell, John L.; Schram, Malachi; VanDevender, Brent A.; Wood, Lynn S.; Wootan, David W.

    2015-03-01

    The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

  13. Sustainability Considerations in Spent Light-water Nuclear Fuel Retrievability

    SciTech Connect

    Wood, Thomas W.; Rothwell, Geoffrey

    2012-01-10

    This paper examines long-term cost differences between two competing Light Water Reactor (LWR) fuels: Uranium Oxide (UOX) and Mixed Uranium Oxide-Plutonium Oxide (MOX). Since these costs are calculated on a life-cycle basis, expected savings from lower future MOX fuel prices can be used to value the option of substituting MOX for UOX, including the value of maintaining access to the used UOX fuel that could be reprocessed to make MOX. The two most influential cost drivers are the price of natural uranium and the cost of reprocessing. Significant and sustained reductions in reprocessing costs and/or sustained increases in uranium prices are required to give positive value to the retrievability of Spent Nuclear Fuel. While this option has positive economic value, it might not be exercised for 50 to 200 years. Therefore, there are many years for a program during which reprocessing technology can be researched, developed, demonstrated, and deployed. Further research is required to determine whether the cost of such a program would yield positive net present value and/or increases the sustainability of LWR energy systems.

  14. Practical reactor production of 41Ar from argon clathrate.

    PubMed

    Mercer, J R; Duke, M J; McQuarrie, S A

    2000-06-01

    The radionuclide 41Ar has many ideal properties as a gas flow tracer. However, the modest cross-section of 40Ar for thermal neutron activation makes preparation of suitable activities of 41Ar technically difficult particularly for low flux reactors. Argon can however be trapped in a molecular complex called a clathrate that can then be irradiated. We prepared argon clathrate and explored its irradiation and stability characteristics. Argon clathrate can be used to provide gigabecquerel quantities of 41Ar even with low power reactors.

  15. Optimization of a flat plate glass reactor for mass production of Nannochloropsis sp. outdoors.

    PubMed

    Richmond, A; Cheng-Wu, Z

    2001-02-23

    The relationships between areal (g m(-2) per day) and volumetric (g l(-1) per day) productivity of Nannochloropsis sp. as affected by the light-path (ranging from 1.3 to 17.0 cm) of a vertical flat plate glass photobioreactor were elucidated. In general, the shorter the length of the light-path (LP), the smaller the areal volume and the higher the volumetric productivity. The areal productivity in relation to the light-path, in contrast, yielded an optimum curve, the highest areal productivity was obtained in a 10 cm LP reactor, which is regarded, therefore, optimal for mass production of Nannochloropsis. An attempt was made to identify criteria by which to assess the efficiency of a photobioreactor in utilizing strong incident energy. Two basic factors which relate to reactor efficiency and its cost-effectiveness have been defined as (a) the total illuminated surface required to produce a set quantity of product and (b) culture volume required to produce that quantity. As a general guide line, the lower these values are, the more efficient and cost-effective the reactor would be. An interesting feature of this analysis rests with the fact that an open raceways is as effective in productivity per illuminated area as a flat-plate reactor with an optimal light path, both cultivation systems requiring ca. 85 m(2) of illuminated surface to produce 1 kg dry cell mass of Nannochloropsis sp. per day. The difference in light utilization efficiency between the two very different production systems involves three aspects - first, the open raceway requires ca. 6 times greater volume than the 10 cm flat plate reactor to produce the same quantity of cell-mass. Second, the total ground area (i.e. including the ground area between reactors) for the vertical flat plate reactor is less than one half of that occupied by an open raceway, indicating the former is more efficient, photosynthetically, compared with the latter. Finally, the harvested cell density is close to one order of

  16. Biological production of ethanol from coal. Task 4 report, Continuous reactor studies

    SciTech Connect

    Not Available

    1992-10-01

    The production of ethanol from synthesis gas by the anaerobic bacterium C. ljungdahlii has been demonstrated in continuous stirred tank reactors (CSTRs), CSTRs with cell recycle and trickle bed reactors. Various liquid media were utilized in these studies including basal medium, basal media with 1/2 B-vitamins and no yeast extract and a medium specifically designed for the growth of C. ljungdahlii in the CSTR. Ethanol production was successful in each of the three reactor types, although trickle bed operation with C. ljungdahlii was not as good as with the stirred tank reactors. Operation in the CSTR with cell recycle was particularly promising, producing 47 g/L ethanol with only minor concentrations of the by-product acetate.

  17. A packed bed membrane reactor for production of biodiesel using activated carbon supported catalyst.

    PubMed

    Baroutian, Saeid; Aroua, Mohamed K; Raman, Abdul Aziz A; Sulaiman, Nik M N

    2011-01-01

    In this study, a novel continuous reactor has been developed to produce high quality methyl esters (biodiesel) from palm oil. A microporous TiO2/Al2O3 membrane was packed with potassium hydroxide catalyst supported on palm shell activated carbon. The central composite design (CCD) of response surface methodology (RSM) was employed to investigate the effects of reaction temperature, catalyst amount and cross flow circulation velocity on the production of biodiesel in the packed bed membrane reactor. The highest conversion of palm oil to biodiesel in the reactor was obtained at 70 °C employing 157.04 g catalyst per unit volume of the reactor and 0.21 cm/s cross flow circulation velocity. The physical and chemical properties of the produced biodiesel were determined and compared with the standard specifications. High quality palm oil biodiesel was produced by combination of heterogeneous alkali transesterification and separation processes in the packed bed membrane reactor.

  18. Stable hydrogen production by methane steam reforming in a two zone fluidized bed reactor: Experimental assessment

    NASA Astrophysics Data System (ADS)

    Pérez-Moreno, L.; Soler, J.; Herguido, J.; Menéndez, M.

    2013-12-01

    The Two Zone Fluidized Bed Reactor concept is proposed for hydrogen production via the steam reforming of methane (SRM) including integrated catalyst regeneration. In order to study the effect of the contact mode, the oxidative SRM has been carried out over a Ni/Al2O3 catalyst using a fixed bed reactor (fBR), a conventional fluidized-bed reactor (FBR) and the proposed two-zone fluidized bed reactor (TZFBR). The technical feasibility of these reactors has been studied experimentally, investigating their performance (CH4 conversion, CO and H2 selectivity, and H2 global yield) and stability under different operating conditions. Coke generation in the process has been verified by several techniques. A stable performance was obtained in the TZFBR, where coke formation was counteracted with continuous catalyst regeneration. The viability of the TZFBR for carrying out this process with a valuable global yield to hydrogen is demonstrated.

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

    SciTech Connect

    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 in 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)

  20. Supplying the nuclear arsenal: Production reactor technology, management, and policy, 1942--1992

    SciTech Connect

    Carlisle, R.P.; Zenzen, J.M.

    1994-01-01

    This book focuses on the lineage of America`s production reactors, those three at Hanford and their descendants, the reactors behind America`s nuclear weapons. The work will take only occasional sideways glances at the collateral lines of descent, the reactor cousins designed for experimental purposes, ship propulsion, and electric power generation. Over the decades from 1942 through 1992, fourteen American production reactors made enough plutonium to fuel a formidable arsenal of more than twenty thousand weapons. In the last years of that period, planners, nuclear engineers, and managers struggled over designs for the next generation of production reactors. The story of fourteen individual machines and of the planning effort to replace them might appear relatively narrow. Yet these machines lay at the heart of the nation`s nuclear weapons complex. The story of these machines is the story of arming the winning weapon, supplying the nuclear arms race. This book is intended to capture the history of the first fourteen production reactors, and associated design work, in the face of the end of the Cold War.

  1. Conceptual design of a new homogeneous reactor for medical radioisotope Mo-99/Tc-99m production

    SciTech Connect

    Liem, Peng Hong; Tran, Hoai Nam; Sembiring, Tagor Malem; Arbie, Bakri

    2014-09-30

    To partly solve the global and regional shortages of Mo-99 supply, a conceptual design of a nitrate-fuel-solution based homogeneous reactor dedicated for Mo-99/Tc-99m medical radioisotope production is proposed. The modified LEU Cintichem process for Mo-99 extraction which has been licensed and demonstrated commercially for decades by BATAN is taken into account as a key design consideration. The design characteristics and main parameters are identified and the advantageous aspects are shown by comparing with the BATAN's existing Mo-99 supply chain which uses a heterogeneous reactor (RSG GAS multipurpose reactor)

  2. Reactor production and processing of radioisotopes for therapeutic applications in nuclear medicine

    SciTech Connect

    Knapp, F.F. Jr.; Mirzadeh, S.; Beets, A.L.

    1995-02-01

    Nuclear reactors continue to play an important role in providing radioisotopes for nuclear medicine. Many reactor-produced radioisotopes are ``neutron rich`` and decay by beta-emission and are thus of interest for therapeutic applications. This talk discusses the production and processing of a variety of reactor-produced radioisotopes of current interest, including those produced by the single neutron capture process, double neutron capture and those available from beta-decay of reactorproduced radioisotopes. Generators prepared from reactorproduced radioisotopes are of particular interest since repeated elution inexpensively provides many patient doses. The development of the alumina-based W-188/Re-188 generator system is discussed in detail.

  3. Conceptual design of a new homogeneous reactor for medical radioisotope Mo-99/Tc-99m production

    NASA Astrophysics Data System (ADS)

    Liem, Peng Hong; Tran, Hoai Nam; Sembiring, Tagor Malem; Arbie, Bakri

    2014-09-01

    To partly solve the global and regional shortages of Mo-99 supply, a conceptual design of a nitrate-fuel-solution based homogeneous reactor dedicated for Mo-99/Tc-99m medical radioisotope production is proposed. The modified LEU Cintichem process for Mo-99 extraction which has been licensed and demonstrated commercially for decades by BATAN is taken into account as a key design consideration. The design characteristics and main parameters are identified and the advantageous aspects are shown by comparing with the BATAN's existing Mo-99 supply chain which uses a heterogeneous reactor (RSG GAS multipurpose reactor).

  4. Development of alloy 718 tubular product for nuclear reactor internals

    SciTech Connect

    1981-01-01

    The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides mixing and flow direction of the core outlet flow. Alloy 718 tubes are the major components used in the UIS to provide this flow direction. The UIS is located directly above the reactor core and is exposed to a severe environment. This environment consists of high temperature sodium, alternating temperatures induced by mixing high temperature core assembly outlet flow with cooler core assembly outlet flow and rapid changes in temperature of the core outlet flow. The paper presents the UIS configuration, functions and environmental conditions that led to the selection of Alloy 718 as the material used to protect the basic UIS structure and to provide flow direction. The paper describes the tube fabrication process, the development of a finish sanding procedure and the results of high temperature thermal cycle testing.

  5. Outlook for the fusion hybrid and tritium-breeding fusion reactors

    NASA Astrophysics Data System (ADS)

    Richardson, J. M.; Cohen, R.; Simpson, J. W.

    The study examines the outlook for fusion hybrid reactors. The study evaluates the status of fusion hybrid technology in the United States and analyzes the circumstances under which such reactors might be deployed. The study also examines a related concept, the tritium-breeding fusion reactor. The study examined two potential applications for fusion hybrid technology: (1) the production of fissile material to fuel light-water reactors, and (2) the direct production of baseload electricity. For both applications, markets were sufficiently problematical or remote (mid-century or later) to warrant only modest current research and development emphasis on technology specific to the fusion hybrid reactor. For the tritium-breeding fusion reactor, a need for tritium for use in nuclear weapons might arise well before the middle of the next century, so that a program of design studies, experimentation, and evaluation should be undertaken.

  6. Biodiesel production in packed-bed reactors using lipase-nanoparticle biocomposite.

    PubMed

    Wang, Xia; Liu, Xueying; Zhao, Chuanming; Ding, Yi; Xu, Ping

    2011-05-01

    The development of appropriate reactors is crucial for the production of biodiesel. In this study, a packed-bed reactor system using lipase-Fe(3)O(4) nanoparticle biocomposite catalyst was successfully developed for biodiesel production based on soybean oil methanolysis. Emulsification before methanolysis improved the reaction rate. The lipase-nanoparticle biocomposite showed high activity and stability in the single-packed-bed reactor at an optimal flow rate (0.25 mL min(-1)). After 240 h of reaction, the conversion rate was sustained as high as 45%. The conversion rate and stability achieved using the four-packed-bed reactor were much higher than those achieved using the single-packed-bed reactor. The conversion of biodiesel was maintained at a high rate of over 88% for 192 h, and it only slightly declined to approximately 75% after 240 h of reaction. The packed-bed reactor system, therefore, has a great potential for achieving the design and operation of enzymatic biodiesel production on the industrial scale.

  7. Fast-quench reactor for hydrogen and elemental carbon production from natural gas and other hydrocarbons

    DOEpatents

    Detering, Brent A.; Kong, Peter C.

    2006-08-29

    A fast-quench reactor for production of diatomic hydrogen and unsaturated carbons is provided. During the fast quench in the downstream diverging section of the nozzle, such as in a free expansion chamber, the unsaturated hydrocarbons are further decomposed by reheating the reactor gases. More diatomic hydrogen is produced, along with elemental carbon. Other gas may be added at different stages in the process to form a desired end product and prevent back reactions. The product is a substantially clean-burning hydrogen fuel that leaves no greenhouse gas emissions, and elemental carbon that may be used in powder form as a commodity for several processes.

  8. Plasma-arc reactor for production possibility of powdered nano-size materials

    NASA Astrophysics Data System (ADS)

    Hadzhiyski, V.; Mihovsky, M.; Gavrilova, R.

    2011-01-01

    Nano-size materials of various chemical compositions find increasing application in life nowadays due to some of their unique properties. Plasma technologies are widely used in the production of a range of powdered nano-size materials (metals, alloys, oxides, nitrides, carbides, borides, carbonitrides, etc.), that have relatively high melting temperatures. Until recently, the so-called RF-plasma generated in induction plasma torches was most frequently applied [1-3]. The subject of this paper is the developments of a new type of plasma-arc reactor, operated with transferred arc system for production of disperse nano-size materials. The new characteristics of the PLASMALAB reactor are the method of feeding the charge, plasma arc control and anode design. The disperse charge is fed by a charge feeding system operating on gravity principle through a hollow cathode of an arc plasma torch situated along the axis of a water-cooled wall vertical tubular reactor. The powdered material is brought into the zone of a plasma space generated by the DC rotating transferred plasma arc. The arc is subjected to Auto-Electro-Magnetic Rotation (AEMR) by an inductor serially connected to the anode circuit. The anode is in the form of a water-cooled copper ring. It is mounted concentrically within the cylindrical reactor, with its lower part electrically insulated from it. The electric parameters of the arc in the reactor and the quantity of processed charge are maintained at a level permitting generation of a volumetric plasma discharge. This mode enables one to attain high mean mass temperature while the processed disperse material flows along the reactor axis through the plasma zone where the main physico-chemical processes take place. The product obtained leaves the reactor through the annular anode, from where it enters a cooling chamber for fixing the produced nano-structure. Experiments for AlN synthesis from aluminium power and nitrogen were carried out using the plasma reactor

  9. Modernization of the VVR-TS reactor core for the increasing of the radionuclides production

    NASA Astrophysics Data System (ADS)

    Fomin, R. V.; Kolesov, V. V.; Kochnov, O. Yu

    2017-01-01

    We have studied the possibility of increasing the 99Mo and 131I production through modernization of the VVR-TS reactor core. It has been found in particular that the introducing of beryllium reflector on the core periphery can significantly increase the starting reactivity margin. This, in turn, give us possibility to create an additional channel for increasing their production.

  10. Excerpt from {open_quotes}Summary of Near-Term Options for Russian Plutonium-Production Reactors{close_quotes}

    SciTech Connect

    1994-12-01

    The Russian Federation desires to stop producing weapons-grade plutonium. During the last several years, ten graphite-moderated, water-cooled, production reactors have been shut down. However, complete cessation of weapons-grade plutonium production is impeded by the fact that the last three operating Russian plutonium-production reactors supply electrical energy and district heat as well as produce plutonium. These reactors are major suppliers of heat in the Tomsk and Krasnoyarsk regions of Siberia.

  11. Stochastic modelling of power reactor fuel behavior

    NASA Astrophysics Data System (ADS)

    Mirza, Shahid Nawaz

    An understanding of the in-reactor behavior of nuclear fuel is essential to the safe and economic operation of a nuclear power plant. It is no longer possible to achieve this without computer code calculations. A state of art computer code, FRODO, for Fuel ROD Operation, has been developed to model the steady state behavior of fuel pins in a light water reactor and to do sensitivity analysis. FRODO concentrates on the thermal performance, fission product release and pellet-clad interaction and can be used to predict the fuel failure under the prevailing conditions. FRODO incorporates the numerous uncertainties involved in fuel behavior modeling, using statistical methods, to ascertain fuel failures and their causes. Sensitivity of fuel failure to different fuel parameters and reactor conditions can be easily evaluated. FRODO has been used to analyze the sensitivities of fuel failures to coolant flow reductions. It is found that the uncertainties have pronounced effects on conclusions about fuel failures and their causes.

  12. Effect of initiator concentration to low-density polyethylene production in a tubular reactor

    NASA Astrophysics Data System (ADS)

    Azmi, A.; Aziz, N.

    2016-11-01

    Low-density polyethylene (LDPE) is one of the most widely used polymers in the world, which is produced in high-capacity tubular and autoclave reactors. As the LDPE industry turn into more competitive and its market profit margins become tighter, manufacturers have to develop solutions to debottleneck the reactor output while abiding to the stringent product specification. A single polyolefin plant producing ten to forty grades of LDPE with various melt flow index (MFI), therefore understanding the reaction mechanism, the operating conditions as well as the dynamic behavior of tubular reactor is essential before any improvement can take place. In the present work, a steady state mathematical model representing a tubular reactor for the production of LDPE is simulated using MATLAB R2015a®. The model developed is a function of feed inlet, reactor jacket, single initiator injector and outlet stream. Analysis on the effect of initiator concentration (CI) shows sudden declining trend of initiator's concentration which indicates that all of the initiators are exhausted after polymerization reaction and no further reaction occur from this point onwards. Furthermore, the results demonstrate that the concentration of initiator gives significant impact on reactor temperature's profile and monomer conversion rate, since higher initiator concentration promotes greater polymerization rate, and therefore leads to higher monomer conversion throughput.

  13. Effect of catalytic cylinders on autothermal reforming of methane for hydrogen production in a microchamber reactor.

    PubMed

    Yan, Yunfei; Guo, Hongliang; Zhang, Li; Zhu, Junchen; Yang, Zhongqing; Tang, Qiang; Ji, Xin

    2014-01-01

    A new multicylinder microchamber reactor is designed on autothermal reforming of methane for hydrogen production, and its performance and thermal behavior, that is, based on the reaction mechanism, is numerically investigated by varying the cylinder radius, cylinder spacing, and cylinder layout. The results show that larger cylinder radius can promote reforming reaction; the mass fraction of methane decreased from 26% to 21% with cylinder radius from 0.25 mm to 0.75 mm; compact cylinder spacing corresponds to more catalytic surface and the time to steady state is decreased from 40 s to 20 s; alteration of staggered and aligned cylinder layout at constant inlet flow rates does not result in significant difference in reactor performance and it can be neglected. The results provide an indication and optimize performance of reactor; it achieves higher conversion compared with other reforming reactors.

  14. Effect of Catalytic Cylinders on Autothermal Reforming of Methane for Hydrogen Production in a Microchamber Reactor

    PubMed Central

    Yan, Yunfei; Guo, Hongliang; Zhang, Li; Zhu, Junchen; Yang, Zhongqing; Tang, Qiang; Ji, Xin

    2014-01-01

    A new multicylinder microchamber reactor is designed on autothermal reforming of methane for hydrogen production, and its performance and thermal behavior, that is, based on the reaction mechanism, is numerically investigated by varying the cylinder radius, cylinder spacing, and cylinder layout. The results show that larger cylinder radius can promote reforming reaction; the mass fraction of methane decreased from 26% to 21% with cylinder radius from 0.25 mm to 0.75 mm; compact cylinder spacing corresponds to more catalytic surface and the time to steady state is decreased from 40 s to 20 s; alteration of staggered and aligned cylinder layout at constant inlet flow rates does not result in significant difference in reactor performance and it can be neglected. The results provide an indication and optimize performance of reactor; it achieves higher conversion compared with other reforming reactors. PMID:25097877

  15. Photolytic treatment of atrazine-contaminated water: products, kinetics, and reactor design.

    PubMed

    Ye, Xuejun; Chen, Daniel; Li, Kuyen; Wang, Bin; Hopper, Jack

    2007-08-01

    This study investigates the products, kinetics, and reactor design of atrazine photolysis under 254-nm ultraviolet-C (UVC) irradiation. With an initial atrazine concentration of 60 microg/L (60 ppbm), only two products remain in detectable levels. Up to 77% of decomposed atrazine becomes hydroxyatrazine, the major product. Both atrazine and hydroxyatrazine photodecompose following the first-order rate equation, but the hydroxyatrazine photodecomposition rate is significantly slower than that of atrazine. For atrazine photodecomposition, the rate constant is proportional to the square of UVC output, but inversely proportional to the reactor volume. For a photochemical reactor design, a series of equations are proposed to calculate the needed UVC output power, water treatment capacity, and atrazine outlet concentration.

  16. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... light water nuclear power reactors which result from exposure of these materials to neutron irradiation..., “Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels”; ASTM E 185-79, “Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels”; and...

  17. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... light water nuclear power reactors which result from exposure of these materials to neutron irradiation..., “Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels”; ASTM E 185-79, “Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels”; and...

  18. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... light water nuclear power reactors which result from exposure of these materials to neutron irradiation..., “Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels”; ASTM E 185-79, “Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels”; and...

  19. Production of activated carbon from coconut shell char in a fluidized bed reactor

    SciTech Connect

    Sai, P.M.S.; Ahmed, J.; Krishnaiah, K.

    1997-09-01

    Activated carbon is produced from coconut shell char using steam or carbon dioxide as the reacting gas in a 100 mm diameter fluidized bed reactor. The effect of process parameters such as reaction time, fluidizing velocity, particle size, static bed height, temperature of activation, fluidizing medium, and solid raw material on activation is studied. The product is characterized by determination of iodine number and BET surface area. The product obtained in the fluidized bed reactor is much superior in quality to the activated carbons produced by conventional processes. Based on the experimental observations, the optimum values of process parameters are identified.

  20. Production of liquid fuels with a high-temperature gas-cooled reactor

    NASA Astrophysics Data System (ADS)

    Quade, R. N.; Vrable, D. L.; Green, L., Jr.

    An exploration is made of the technical, economic and environmental impact feasibility of integrating coal liquefaction methods directly and indirectly with a nuclear reactor source of process heat, with stress on the production of synthetic jet fuel. Production figures and operating costs are compared for indirect conventional and nuclear processes using Lurgi-Fischer-Tropsch technology with direct conventional and nuclear techniques employing the advanced SRC-II technology, and it is concluded that significant advantages in coal savings and environmental impact can be expected from nuclear reactor integration.

  1. Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

    SciTech Connect

    Gryzinski, M.A.; Wielgosz, M.

    2015-07-01

    The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel is additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with

  2. Energy analysis for the production of biodiesel in a spiral reactor using supercritical tert-butyl methyl ether (MTBE).

    PubMed

    Farobie, Obie; Matsumura, Yukihiko

    2015-11-01

    In this study, energy analysis was conducted for the production of biodiesel in a spiral reactor using supercritical tert-butyl methyl ether (MTBE). This study aims to determine the net energy ratio (NER) and energy efficiency for the production of biodiesel using supercritical MTBE and to verify the effectiveness of the spiral reactor in terms of heat recovery efficiency. The analysis results revealed that the NER for this process was 0.92. Meanwhile, the energy efficiency was 0.98, indicating that the production of biodiesel in a spiral reactor using supercritical MTBE is an energy-efficient process. By comparing the energy supply required for biodiesel production between spiral and conventional reactors, the spiral reactor was more efficient than the conventional reactor.

  3. Removal of anaerobic soluble microbial products in a biological activated carbon reactor.

    PubMed

    Dong, Xiaojing; Zhou, Weili; He, Shengbing

    2013-09-01

    The soluble microbial products (SMP) in the biological treatment effluent are generally of great amount and are poorly biodegradable. Focusing on the biodegradation of anaerobic SMP, the biological activated carbon (BAC) was introduced into the anaerobic system. The experiments were conducted in two identical lab-scale up-flow anaerobic sludge blanket (UASB) reactors. The high strength organics were degraded in the first UASB reactor (UASB1) and the second UASB (UASB2, i.e., BAC) functioned as a polishing step to remove SMP produced in UASB1. The results showed that 90% of the SMP could be removed before granular activated carbon was saturated. After the saturation, the SMP removal decreased to 60% on the average. Analysis of granular activated carbon adsorption revealed that the main role of SMP removal in BAC reactor was biodegradation. A strain of SMP-degrading bacteria, which was found highly similar to Klebsiella sp., was isolated, enriched and inoculated back to the BAC reactor. When the influent chemical oxygen demand (COD) was 10,000 mg/L and the organic loading rate achieved 10 kg COD/(m3 x day), the effluent from the BAC reactor could meet the discharge standard without further treatment. Anaerobic BAC reactor inoculated with the isolated Klebsiella was proved to be an effective, cheap and easy technical treatment approach for the removal of SMP in the treatment of easily-degradable wastewater with COD lower than 10,000 mg/L.

  4. Numerical simulation of vortex pyrolysis reactors for condensable tar production from biomass

    SciTech Connect

    Miller, R.S.; Bellan, J.

    1998-08-01

    A numerical study is performed in order to evaluate the performance and optimal operating conditions of vortex pyrolysis reactors used for condensable tar production from biomass. A detailed mathematical model of porous biomass particle pyrolysis is coupled with a compressible Reynolds stress transport model for the turbulent reactor swirling flow. An initial evaluation of particle dimensionality effects is made through comparisons of single- (1D) and multi-dimensional particle simulations and reveals that the 1D particle model results in conservative estimates for total pyrolysis conversion times and tar collection. The observed deviations are due predominantly to geometry effects while directional effects from thermal conductivity and permeability variations are relatively small. Rapid ablative particle heating rates are attributed to a mechanical fragmentation of the biomass particles that is modeled using a critical porosity for matrix breakup. Optimal thermal conditions for tar production are observed for 900 K. Effects of biomass identity, particle size distribution, and reactor geometry and scale are discussed.

  5. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    SciTech Connect

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A. Ignatiev, V. V.; Subbotin, S. A. Tsibulskiy, V. F.

    2015-12-15

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  6. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    NASA Astrophysics Data System (ADS)

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-12-01

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  7. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    SciTech Connect

    Lin, Jerry Y.S.

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

  8. Improvement in the bioreactor specific productivity by coupling continuous reactor with repeated fed-batch reactor for acetone-butanol-ethanol production.

    PubMed

    Setlhaku, Mpho; Brunberg, Sina; Villa, Eva Del Amor; Wichmann, Rolf

    2012-10-15

    In comparison to the different fermentation modes for the production of acetone, butanol and ethanol (ABE) researched to date, the continuous fermentation is the most economically favored. Continuous fermentation with two or more reactor cascade is reported to be the most efficient as it results in a more stable solvent production process. In this work, it is shown that a continuous (first-stage) reactor coupled to a repeated fed-batch (second stage) is superior to batch and fed-batch fermentations, including two-stage continuous fermentation. This is due to the efficient catalyst use, reported through the specific product rate and rapid glucose consumption rate. High solvents are produced at 19.4 g(ABE) l⁻¹, with volumetric productivities of 0.92 g(butanol) l⁻¹ h⁻¹ and 1.47 g(ABE) l ⁻¹ h⁻¹. The bioreactor specific productivities of 0.62 and 0.39 g g⁻¹(cdw) h⁻¹ obtained show a high catalyst activity. This new process mode has not been reported before in the development of ABE fermentation and it shows great potential and superiority to the existing fermentation methods.

  9. Effect of substrate concentration on dark fermentation hydrogen production using an anaerobic fluidized bed reactor.

    PubMed

    de Amorim, Eduardo Lucena Cavalcante; Sader, Leandro Takano; Silva, Edson Luiz

    2012-03-01

    The effect of substrate (glucose) concentration on the stability and yield of a continuous fermentative process that produces hydrogen was studied. Four anaerobic fluidized bed reactors (AFBRs) were operated with a hydraulic retention time (HRT) from 1 to 8 h and an influent glucose concentration from 2 to 25 g L(-1). The reactors were inoculated with thermally pre-treated anaerobic sludge and operated at a temperature of 30 °C with an influent pH around 5.5 and an effluent pH of about 3.5. The AFBRs with a HRT of 2 h and a feed strength of 2, 4, and 10 g L(-1) showed satisfactory H(2) production performance, but the reactor fed with 25 g L(-1) of glucose did not. The highest hydrogen yield value was obtained in the reactor with a glucose concentration of 2 g L(-1) when it was operated at a HRT of 2 h. The maximum hydrogen production rate value was achieved in the reactor with a HRT of 1 h and a feed strength of 10 g L(-1). The AFBRs operated with glucose concentrations of 2 and 4 g L(-1) produced greater amounts of acetic and butyric acids, while AFBRs with higher glucose concentrations produced a greater amount of solvents.

  10. A molten Salt Am242M Production Reactor for Space Applications

    NASA Technical Reports Server (NTRS)

    Emrich, William

    2005-01-01

    The use of Am242m holds great promise for increasing the efficiency nuclear thermal rocket engines. Because Am242m has the highest fission cross section of any known isotope (1000's of barns), its extremely high reactivity may be used to directly heat a propellant gas with fission fragments. Since this isotope does not occur naturally, it must be bred in special production reactors designed for that purpose. The primary advantage to using molten salt reactors for breeding Am242m is that the reactors can be reprocessed continually yielding a constant rate of production of the isotope. Once built and initially fueled, the reactor will continually breed the additional fuel it needs to remain critical. The only feedstock required is a salt of U238. No enriched fuel is required during normal operation and all fissile material, except the Am242m, is maintained in a closed loop. For a reactor operating at 200 MW several kilograms of Am242m may be bred each year.

  11. Production of a Biopolymer at Reactor Scale: A Laboratory Experience

    ERIC Educational Resources Information Center

    Genc, Rukan; Rodriguez-Couto, Susana

    2011-01-01

    Undergraduate students of biotechnology became familiar with several aspects of bioreactor operation via the production of xanthan gum, an industrially relevant biopolymer, by "Xanthomonas campestris" bacteria. The xanthan gum was extracted from the fermentation broth and the yield coefficient and productivity were calculated. (Contains 2 figures.)

  12. Venting of fission products and shielding in thermionic nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Salmi, E. W.

    1972-01-01

    Most thermionic reactors are designed to allow the fission gases to escape out of the emitter. A scheme to allow the fission gases to escape is proposed. Because of the low activity of the fission products, this method should pose no radiation hazards.

  13. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    SciTech Connect

    M. J. Russell

    2006-06-01

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

  14. Zero valent iron simultaneously enhances methane production and sulfate reduction in anaerobic granular sludge reactors.

    PubMed

    Liu, Yiwen; Zhang, Yaobin; Ni, Bing-Jie

    2015-05-15

    Zero valent iron (ZVI) packed anaerobic granular sludge reactors have been developed for improved anaerobic wastewater treatment. In this work, a mathematical model is developed to describe the enhanced methane production and sulfate reduction in anaerobic granular sludge reactors with the addition of ZVI. The model is successfully calibrated and validated using long-term experimental data sets from two independent ZVI-enhanced anaerobic granular sludge reactors with different operational conditions. The model satisfactorily describes the chemical oxygen demand (COD) removal, sulfate reduction and methane production data from both systems. Results show ZVI directly promotes propionate degradation and methanogenesis to enhance methane production. Simultaneously, ZVI alleviates the inhibition of un-dissociated H2S on acetogens, methanogens and sulfate reducing bacteria (SRB) through buffering pH (Fe(0) + 2H(+) = Fe(2+) + H2) and iron sulfide precipitation, which improve the sulfate reduction capacity, especially under deterioration conditions. In addition, the enhancement of ZVI on methane production and sulfate reduction occurs mainly at relatively low COD/ [Formula: see text] ratio (e.g., 2-4.5) rather than high COD/ [Formula: see text] ratio (e.g., 16.7) compared to the reactor without ZVI addition. The model proposed in this work is expected to provide support for further development of a more efficient ZVI-based anaerobic granular system.

  15. American National Standard: design basis for protection of light water nuclear power plants against effects of postulated pipe rupture

    SciTech Connect

    Not Available

    1980-12-31

    This standard addresses the design bases for light water reactor, nuclear power plant structures and components essential for the protection of public health and safety from the potential adverse effects of pipe whip, jet impingement, pressurization of compartments outside containment, environmental conditions and flooding associated with a postulated pipe rupture. The design bases for missile protection and the design bases for containment pressurization are not within this standard.

  16. Particle characteristics in the reactor and pelletizing areas of carbon black production.

    PubMed

    Kuhlbusch, T A J; Fissan, H

    2006-10-01

    Physical and chemical characteristics of airborne particles (ultrafine, PM1, PM2.5, and PM10) in reactor and pelletizing areas during carbon black production were measured to assess process related sources of particles in work areas. Results from bagging areas within the same three facilities have been previously published. Particle number and mass concentration measurements were conducted in these work areas and at ambient comparison sites at each of the three carbon black plants. No elevated ultrafine particle number concentrations (UFP, <100 nm) with respect to ambient were determined in the work areas of Plant 1, intermittently elevated concentrations at Plant 2, and permanently elevated concentrations at Plant 3. The intermittently elevated UFP concentrations in the pelletizer and reactor areas of Plant 2 could be related to nearby traffic emissions. The ultrafine particle number concentrations at Plant 2 are comparable to those determined at urban traffic sites. Both work areas of Plant 3 showed elevated UFP concentrations in the pelletizer reactor and areas. In the case of the reactor, which was the only enclosed reactor area investigated among the three facilities, the source of the elevated UFP number concentration was most likely attributable to grease and oil fumes from maintenance activities, a conclusion supported by carbon fractionation analysis. The elevated UFP number concentrations in the pelletizing area in this same plant are related to leaks in the production line, which allowed particulate matter to escape to the surrounding areas. Absolute PM10 mass concentrations were all within normal ambient concentrations except for the pelletizing area in Plant 3, which showed continuous levels above ambient. One additional source contributing to peak level PM10 mass concentrations at Plant 2 was due to wind dispersion from a carbon black spill incident the day prior to measurements. It is concluded from these measurements that no carbon black is released

  17. Design and construction of a cascading pressure reactor prototype for solar-thermochemical hydrogen production

    NASA Astrophysics Data System (ADS)

    Ermanoski, Ivan; Grobbel, Johannes; Singh, Abhishek; Lapp, Justin; Brendelberger, Stefan; Roeb, Martin; Sattler, Christian; Whaley, Josh; McDaniel, Anthony; Siegel, Nathan P.

    2016-05-01

    Recent work regarding the efficiency maximization for solar thermochemical fuel production in two step cycles has led to the design of a new type of reactor—the cascading pressure reactor—in which the thermal reduction step of the cycle is completed in multiple stages, at successively lower pressures. This approach enables lower thermal reduction pressures than in single-staged reactors, and decreases required pump work, leading to increased solar to fuel efficiencies. Here we report on the design and construction of a prototype cascading pressure reactor and testing of some of the key components. We especially focus on the technical challenges particular to the design, and their solutions.

  18. Fission-product data analysis from actinide samples exposed in the Dounreay Prototype Fast Reactor

    SciTech Connect

    Murphy, B.D.; Dickens, J.K.; Walker, R.L.; Newton, T.D.

    1994-12-31

    Since 1979 a cooperative agreement has been in effect between the United States and the United Kingdom to investigate the irradiation of various actinide species placed in the core of the Dounreay Prototype Fast Reactor (PFR). The irradiated species were isotopes of thorium, protactinium, uranium, neptunium, plutonium, americium, and curium. A set of actinide samples (mg quantities) was exposed to about 490 effective full power days (EFPD) of reactor operations. The fission-product results are reported here. The actinide results will be report elsewhere.

  19. Bio-oil production from palm fronds by fast pyrolysis process in fluidized bed reactor

    NASA Astrophysics Data System (ADS)

    Rinaldi, Nino; Simanungkalit, Sabar P.; Kiky Corneliasari, S.

    2017-01-01

    Fast pyrolysis process of palm fronds has been conducted in the fluidized bed reactor to yield bio-oil product (pyrolysis oil). The process employed sea sand as the heat transfer medium. The objective of this study is to design of the fluidized bed rector, to conduct fast pyrolysis process to product bio-oil from palm fronds, and to characterize the feed and bio-oil product. The fast pyrolysis process was conducted continuously with the feeding rate around 500 g/hr. It was found that the biomass conversion is about 35.5% to yield bio-oil, however this conversion is still minor. It is suggested due to the heating system inside the reactor was not enough to decompose the palm fronds as a feedstock. Moreover, the acids compounds ware mostly observed on the bio-oil product.

  20. Management of Hanford Site non-defense production reactor spent nuclear fuel, Hanford Site, Richland, Washington

    SciTech Connect

    1997-03-01

    The US Department of Energy (DOE) needs to provide radiologically, and industrially safe and cost-effective management of the non-defense production reactor spent nuclear fuel (SNF) at the Hanford Site. The proposed action would place the Hanford Site`s non-defense production reactor SNF in a radiologically- and industrially-safe, and passive storage condition pending final disposition. The proposed action would also reduce operational costs associated with storage of the non-defense production reactor SNF through consolidation of the SNF and through use of passive rather than active storage systems. Environmental, safety and health vulnerabilities associated with existing non-defense production reactor SNF storage facilities have been identified. DOE has determined that additional activities are required to consolidate non-defense production reactor SNF management activities at the Hanford Site, including cost-effective and safe interim storage, prior to final disposition, to enable deactivation of facilities where the SNF is now stored. Cost-effectiveness would be realized: through reduced operational costs associated with passive rather than active storage systems; removal of SNF from areas undergoing deactivation as part of the Hanford Site remediation effort; and eliminating the need to duplicate future transloading facilities at the 200 and 400 Areas. Radiologically- and industrially-safe storage would be enhanced through: (1) removal from aging facilities requiring substantial upgrades to continue safe storage; (2) utilization of passive rather than active storage systems for SNF; and (3) removal of SNF from some storage containers which have a limited remaining design life. No substantial increase in Hanford Site environmental impacts would be expected from the proposed action. Environmental impacts from postulated accident scenarios also were evaluated, and indicated that the risks associated with the proposed action would be small.

  1. Production of {sup 99}Mo using LEU and molybdenum targets in a 1 MW Triga reactor

    SciTech Connect

    Mo, S.C.

    1993-12-31

    The production of {sup 99}Mo using Low Enriched Uranium (LEU) and natural molybdenum targets in a 1 MW Triga reactor is investigated. The successive linear programming technique is applied to minimize the target loadings for different yield constraints. The irradiation time is related to the kinetics of the growth and decay of {sup 99}Mo. The feasibility of a neutron generated based {sup 99}Mo production system is discussed.

  2. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    SciTech Connect

    Licht, J. R.; Bergeron, A.; Dionne, B.; Van den Branden, G.; Kalcheva, S.; Sikik, E.; Koonen, E.

    2015-12-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

  3. Toward continuous and scalable production of colloidal nanocrystals by switching from batch to droplet reactors.

    PubMed

    Niu, Guangda; Ruditskiy, Aleksey; Vara, Madeline; Xia, Younan

    2015-08-21

    Colloidal nanocrystals are finding widespread use in a wide variety of applications ranging from catalysis to photonics, electronics, energy harvesting/conversion/storage, environment protection, information storage, and biomedicine. Despite the large number of successful demonstrations, there still exists a significant gap between academic studies and industrial applications owing to the lack of an ability to produce colloidal nanocrystals in large quantities without losing control over their properties. Droplet reactors have shown great potential for the continuous and scalable production of colloidal nanocrystals with uniform and well-controlled sizes, shapes, structures, and compositions. In this tutorial review, we begin with rationales for the use of droplet reactors as a new platform to scale up the production of colloidal nanocrystals, followed by discussions of the general concepts and technical challenges in applying droplet reactors to the synthesis of nanocrystals, including droplet formation, introduction and mixing of reagents, management of gaseous species, and interfacial adsorption. At the end, we use a set of examples to highlight the unique capabilities of droplet reactors for the high-volume production of colloidal nanocrystals in the setting of both homogeneous nucleation and seed-mediated growth.

  4. Methane production enhancement by an independent cathode in integrated anaerobic reactor with microbial electrolysis.

    PubMed

    Cai, Weiwei; Han, Tingting; Guo, Zechong; Varrone, Cristiano; Wang, Aijie; Liu, Wenzong

    2016-05-01

    Anaerobic digestion (AD) represents a potential way to achieve energy recovery from waste organics. In this study, a novel bioelectrochemically-assisted anaerobic reactor is assembled by two AD systems separated by anion exchange membrane, with the cathode placing in the inside cylinder (cathodic AD) and the anode on the outside cylinder (anodic AD). In cathodic AD, average methane production rate goes up to 0.070 mL CH4/mL reactor/day, which is 2.59 times higher than AD control reactor (0.027 m(3) CH4/m(3)/d). And COD removal is increased ∼15% over AD control. When changing to sludge fermentation liquid, methane production rate has been further increased to 0.247 mL CH4/mL reactor/day (increased by 51.53% comparing with AD control). Energy recovery efficiency presents profitable gains, and economic revenue from increased methane totally self-cover the cost of input electricity. The study indicates that cathodic AD could cost-effectively enhance methane production rate and degradation of glucose and fermentative liquid.

  5. Design and analysis of an immobilized cell reactor with simultaneous product separation: ethanol from whey lactose

    SciTech Connect

    Dale, M.C.

    1983-01-01

    The simultaneous separation of volatile fermentation products from product inhibited fermentations can increase the productivity of a bioreactor by reducing the product concentration in the bioreactor. In this work, a simultaneous tubular reactor separator is developed in which the volatile product is removed from the reacting broth by an inert gas phase. The immobilized cell reactor separator (ICRS) consists of two column reactors: a cocurrent enriching column followed by an countercurrent stripping column. The application of the ICRS concept to the ethanol from whey lactose fermentation was investigated using the yeast Kluyveromyces fragilis 2415. An equilibrium stage model of the ICRS was developed including a surface renewal term for an adsorbed monolayer of reacting cells. This model demonstrated the effect of important operational parameters including temperature, pressure, and gas flow rates. Experimental results using yeast adsorbed to 1/4'' ceramic saddles were somewhat unsatisfactory but very high productivities, cell densities, and separation efficiency were obtained using an absorbant column packing in a gas continuous operating mode.

  6. Biodiesel production from palm oil using combined mechanical stirred and ultrasonic reactor.

    PubMed

    Choedkiatsakul, I; Ngaosuwan, K; Cravotto, G; Assabumrungrat, S

    2014-07-01

    This paper investigates the production of biodiesel from palm oil using a combined mechanical stirred and ultrasonic reactor (MS-US). The incorporation of mechanical stirring into the ultrasonic reactor explored the further improvement the transesterification of palm oil. Initial reaction rate values were 54.1, 142.9 and 164.2 mmol/L min for the mechanical-stirred (MS), ultrasonic (US) and MS-US reactors, respectively. Suitable methanol to oil molar ratio and the catalyst loading values were found to be 6 and 1 of oil, respectively. The effect of ultrasonic operating parameters; i.e. frequency, location, and number of transducer, has been investigated. Based on the conversion yield at the reactor outlet after 1 h, the number of transducers showed a relevant role in the reaction rate. Frequency and transducer location would appear to have no significant effect. The properties of the obtained biodiesel (density, viscosity, pour point, and flash point) satisfy the ASTM standard. The combined MS-US reactors improved the reaction rate affording the methyl esters in higher yield.

  7. Temperature effect towards methane gas production and performances of anaerobic fixed bed reactors

    NASA Astrophysics Data System (ADS)

    Budiastuti, H.; Widyabudiningsih, D.; Kurnia, D. R. D.

    2016-11-01

    This study was conducted to observe the effect of temperature towards methane gas production and performances of Anaerobic Fixed Bed Reactors (AFBR) in degrading artificial wastewater. During seeding and acclimatization as well as operation at normal loading rates the reactor was controlled at 35 ± 1 °C. The reactor temperature was then switched to room temperature to compare its performance with performance of the temperature controlled reactor. Seeding was performed by wastewater feeding at concentration of 5,300 mgCOD/L. Seeding was continued to acclimatization after COD efficiency reached about 30% and mixed liquor volatile suspended solids (MLVSS) in the reactor were in the range of 28 to 29 g/L. Acclimatization was conducted by step increased feeding while obtaining organic loading rates (OLR) of 530 mgCOD/L/day and 40 days HRT, and it was stopped when COD efficiencies almost constant at about 80% were achieved. Normal loads at OLR of 530 mgCOD/L/day and HRT of 40 days resulted in COD efficiencies in the range of 80 to 92%, produced methane gas ranged from 80 to 170 ml/day, at pH around 7 during controlled temperature. At room temperature, COD efficiencies decreased to the fluctuated range of 78 to 84%, methane gas dropped to the highest of 144 ml/day but pH range was still at around 7. At OLR 1.5 times normal OLR (795 mgCOD/L/day) and HRT of 40 days, the controlled temperature reactor shows superiority by producing efficiencies in the range of 84 to 94% and 260 mL/day of CH4 as the highest gas produced during observation. Efficiencies of COD obtained during uncontrolled temperature dropped to the range of 60 to 80%. The uncontrolled temperature reactor only produced the highest CH4 production of 48 mL/day. pH from both reactors still ranged in the normal pH range about 7. It shows that AFBR has to be controlled at around 35 °C to produce higher COD efficiencies and methane gas production.

  8. Fast Spectrum Molten Salt Reactor Options

    SciTech Connect

    Gehin, Jess C; Holcomb, David Eugene; Flanagan, George F; Patton, Bruce W; Howard, Rob L; Harrison, Thomas J

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  9. Production of Biodiesel at Kinetic Limit Achieved in a Centrifugal Reactor/Separator

    SciTech Connect

    McFarlane, Joanna; Tsouris, Costas; Birdwell Jr, Joseph F; Lee, Denise L; Jennings, Hal L; Pahmer Boitrago, Amy M; Terpstra, Sarah M

    2010-01-01

    The kinetics of the transesterification of soybean oil has been investigated in a centrifugal reactor at temperatures from 45 to 80 C and pressures up to 2.6 bar using gas chromatography flame ionization detection (GC-FID) and infrared (IR) spectroscopy. The yields of product methyl esters were quantified using IR, proton Nuclear Magnetic Resonance (H1NMR), and viscosity measurements and were found to achieve 90% of the yield in 2 min; however, to meet ASTM specifications with one pass through the reactor, a 15 min residence time was needed. Performance was improved by sequential reactions, allowing separation of by-product glycerine and injection of additional small aliquots of methanol. The kinetics was modeled using a three-step mechanism of reversible reactions, which was used to predict performance at commercial scale. The mechanism correctly predicted the exponential decline in reaction rate as the concentration of the products allowed significant reverse reactions to occur.

  10. Use of LEU in the aqueous homogeneous medical isotope production reactor

    SciTech Connect

    Ball, R.M.

    1997-08-01

    The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution.

  11. Kinetic study on the effect of temperature on biogas production using a lab scale batch reactor.

    PubMed

    Deepanraj, B; Sivasubramanian, V; Jayaraj, S

    2015-11-01

    In the present study, biogas production from food waste through anaerobic digestion was carried out in a 2l laboratory-scale batch reactor operating at different temperatures with a hydraulic retention time of 30 days. The reactors were operated with a solid concentration of 7.5% of total solids and pH 7. The food wastes used in this experiment were subjected to characterization studies before and after digestion. Modified Gompertz model and Logistic model were used for kinetic study of biogas production. The kinetic parameters, biogas yield potential of the substrate (B), the maximum biogas production rate (Rb) and the duration of lag phase (λ), coefficient of determination (R(2)) and root mean square error (RMSE) were estimated in each case. The effect of temperature on biogas production was evaluated experimentally and compared with the results of kinetic study. The results demonstrated that the reactor with operating temperature of 50°C achieved maximum cumulative biogas production of 7556ml with better biodegradation efficiency.

  12. Continuous fermentative hydrogen production using a two-phase reactor system with recycle.

    PubMed

    Kraemer, Jeremy T; Bagley, David M

    2005-05-15

    The effects of effluent recycle were examined in a two-phase anaerobic system where the first phase was operated for fermentative hydrogen production and the second for methanogenesis. The hydrogen reactor was operated as a chemostat at 35 degrees C and pH 5.5 with a 10 h hydraulic retention time, and the methane reactor was operated as an up-flow reactor at 28 degrees C and pH between 6.9 and 7.2. Two recycle ratios were examined: 0 and 0.98. Effluent recycle reduced the required alkalinity for pH control by approximately 40%. The H2 productivity metric, with a basis in electrons and incorporating both gaseous and dissolved H2, was developed as a more fundamental reporting method than the molar H2 yield. Without recycle, the H2 productivity was 0.115 g of H2 COD/g of feed COD, but decreased to 0.015 q of H2 COD/g of feed COD with recycle (COD = chemical oxygen demand). Mass balances indicated the lower H2 productivity during recycle was due to electrons being partitioned to methane and less-oxidized soluble constituents such as propionic acid, ethanol, and butanol. The results indicated that achieving high H2 productivity with nonsterile wastewaters will be challenging and membrane filtration of the recycle liquid may be required to exclude the return of hydrogen-consuming organisms.

  13. [Hydrolyzed lactose contained in the ultrafiltrate of milk or milk products in an enzymatic membrane reactor].

    PubMed

    Roger, L; Maubois, J L; Thapon, J L; Brule, G

    1978-01-01

    Milk and milk by-products with a low lactose content, very interesting from a nutritional and technological point of view, were obtained by the application of the enzymatic membrane reactor technique. A previous separation of the aqueous phase of milk or ultrafiltrate was necessary and realized by ultrafiltration. The enzyme, a commercial beta-galactosidase, was maintained in solution in the retentate part of the membrane reactor. The optimal conditions of the lactose hydrolysis in milk and whey ultrafiltrates were determined. The behaviour of the aqueous phase of milk in membrane reactor, specially of mineral salts, was studied. Three possibilities were proposed to avoid a calcium-phosphate deposit on the surface of (and in) the reactor membranes: a precipitation of calcium salts by heating, a partial demineralization by electrodialysis or ion exchange, a calcium complexation by addition of sodium citrate. A continuous process for the lactose hydrolysis of milk and demineralized whey or milk ultrafiltrate was proposed. The organoleptic quality of low lactose milk, before and after heat treatment, was evaluated by a tasting panel. High sweeting syrup, were obtained by concentration of lactose hydrolyzed and demineralized ultrafiltrates. Nutritional aspects of these products are discussed specially from the toxicological point of view of galactose.

  14. Scalable synthesis of palladium icosahedra in plug reactors for the production of oxygen reduction reaction catalysts

    DOE PAGES

    Wang, Helan; Niu, Guangda; Zhou, Ming; ...

    2016-03-10

    We have synthesized Pd icosahedra with uniform, controllable sizes in plug reactors separated by air. The oxygen contained in the air segments not only contributed to the generation of a reductant from diethylene glycol in situ, but also oxidized elemental Pd back to the ionic form by oxidative etching and thus slowed down the reduction kinetics. Compared to droplet reactors involving silicone oil or fluorocarbon, the use of air as a carrier phase could reduce the production cost by avoiding additional procedures for the separation of products from the oil. The average diameters of the Pd icosahedra could be readilymore » controlled in the range of 12–20 nm. The Pd icosahedra were further employed as seeds for the production of Pd@Pt2–3L core-shell icosahedra, which could serve as a catalyst toward the oxygen reduction reaction with greatly enhanced activity. As a result, we believe that the plug reactors could be extended to other types of noble-metal nanocrystals for their scale-up production.« less

  15. Scalable synthesis of palladium icosahedra in plug reactors for the production of oxygen reduction reaction catalysts

    SciTech Connect

    Wang, Helan; Niu, Guangda; Zhou, Ming; Wang, Xue; Park, Jinho; Bao, Shixiong; Chi, Miaofang; Cai, Zaisheng; Xia, Younan

    2016-03-10

    We have synthesized Pd icosahedra with uniform, controllable sizes in plug reactors separated by air. The oxygen contained in the air segments not only contributed to the generation of a reductant from diethylene glycol in situ, but also oxidized elemental Pd back to the ionic form by oxidative etching and thus slowed down the reduction kinetics. Compared to droplet reactors involving silicone oil or fluorocarbon, the use of air as a carrier phase could reduce the production cost by avoiding additional procedures for the separation of products from the oil. The average diameters of the Pd icosahedra could be readily controlled in the range of 12–20 nm. The Pd icosahedra were further employed as seeds for the production of Pd@Pt2–3L core-shell icosahedra, which could serve as a catalyst toward the oxygen reduction reaction with greatly enhanced activity. As a result, we believe that the plug reactors could be extended to other types of noble-metal nanocrystals for their scale-up production.

  16. Fission products from the damaged Fukushima reactor observed in Hungary.

    PubMed

    Bihari, Árpád; Dezső, Zoltán; Bujtás, Tibor; Manga, László; Lencsés, András; Dombóvári, Péter; Csige, István; Ranga, Tibor; Mogyorósi, Magdolna; Veres, Mihály

    2014-01-01

    Fission products, especially (131)I, (134)Cs and (137)Cs, from the damaged Fukushima Dai-ichi nuclear power plant (NPP) were detected in many places worldwide shortly after the accident caused by natural disaster. To observe the spatial and temporal variation of these isotopes in Hungary, aerosol samples were collected at five locations from late March to early May 2011: Institute of Nuclear Research, Hungarian Academy of Sciences (ATOMKI, Debrecen, East Hungary), Paks NPP (Paks, South-Central Hungary) as well as at the vicinity of Aggtelek (Northeast Hungary), Tapolca (West Hungary) and Bátaapáti (Southwest Hungary) settlements. In addition to the aerosol samples, dry/wet fallout samples were collected at ATOMKI, and airborne elemental iodine and organic iodide samples were collected at Paks NPP. The peak in the activity concentration of airborne (131)I was observed around 30 March (1-3 mBq m(-3) both in aerosol samples and gaseous iodine traps) with a slow decline afterwards. Aerosol samples of several hundred cubic metres of air showed (134)Cs and (137)Cs in detectable amounts along with (131)I. The decay-corrected inventory of (131)I fallout at ATOMKI was 2.1±0.1 Bq m(-2) at maximum in the observation period. Dose-rate contribution calculations show that the radiological impact of this event at Hungarian locations was of no considerable concern.

  17. Fast Breeder Reactor studies

    SciTech Connect

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  18. Methane production by treating vinasses from hydrous ethanol using a modified UASB reactor

    PubMed Central

    2012-01-01

    Background A modified laboratory-scale upflow anaerobic sludge blanket (UASB) reactor was used to obtain methane by treating hydrous ethanol vinasse. Vinasses or stillage are waste materials with high organic loads, and a complex composition resulting from the process of alcohol distillation. They must initially be treated with anaerobic processes due to their high organic loads. Vinasses can be considered multipurpose waste for energy recovery and once treated they can be used in agriculture without the risk of polluting soil, underground water or crops. In this sense, treatment of vinasse combines the elimination of organic waste with the formation of methane. Biogas is considered as a promising renewable energy source. The aim of this study was to determine the optimum organic loading rate for operating a modified UASB reactor to treat vinasse generated in the production of hydrous ethanol from sugar cane molasses. Results The study showed that chemical oxygen demand (COD) removal efficiency was 69% at an optimum organic loading rate (OLR) of 17.05 kg COD/m3-day, achieving a methane yield of 0.263 m3/kg CODadded and a biogas methane content of 84%. During this stage, effluent characterization presented lower values than the vinasse, except for potassium, sulfide and ammonia nitrogen. On the other hand, primers used to amplify the 16S-rDNA genes for the domains Archaea and Bacteria showed the presence of microorganisms which favor methane production at the optimum organic loading rate. Conclusions The modified UASB reactor proposed in this study provided a successful treatment of the vinasse obtained from hydrous ethanol production. Methanogen groups (Methanobacteriales and Methanosarcinales) detected by PCR during operational optimum OLR of the modified UASB reactor, favored methane production. PMID:23167984

  19. Modeling of the HiPco process for carbon nanotube production. II. Reactor-scale analysis

    NASA Technical Reports Server (NTRS)

    Gokcen, Tahir; Dateo, Christopher E.; Meyyappan, M.

    2002-01-01

    The high-pressure carbon monoxide (HiPco) process, developed at Rice University, has been reported to produce single-walled carbon nanotubes from gas-phase reactions of iron carbonyl in carbon monoxide at high pressures (10-100 atm). Computational modeling is used here to develop an understanding of the HiPco process. A detailed kinetic model of the HiPco process that includes of the precursor, decomposition metal cluster formation and growth, and carbon nanotube growth was developed in the previous article (Part I). Decomposition of precursor molecules is necessary to initiate metal cluster formation. The metal clusters serve as catalysts for carbon nanotube growth. The diameter of metal clusters and number of atoms in these clusters are some of the essential information for predicting carbon nanotube formation and growth, which is then modeled by the Boudouard reaction with metal catalysts. Based on the detailed model simulations, a reduced kinetic model was also developed in Part I for use in reactor-scale flowfield calculations. Here this reduced kinetic model is integrated with a two-dimensional axisymmetric reactor flow model to predict reactor performance. Carbon nanotube growth is examined with respect to several process variables (peripheral jet temperature, reactor pressure, and Fe(CO)5 concentration) with the use of the axisymmetric model, and the computed results are compared with existing experimental data. The model yields most of the qualitative trends observed in the experiments and helps to understanding the fundamental processes in HiPco carbon nanotube production.

  20. Accelerator Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

    SciTech Connect

    Brown, Nicholas R.; Heidet, Florent; Haj Tahar, Malek

    2016-01-01

    This article is a review of several accelerator–reactor interface issues and nuclear fuel cycle applications of acceleratordriven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systems on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.

  1. Accelerator Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

    DOE PAGES

    Brown, Nicholas R.; Heidet, Florent; Haj Tahar, Malek

    2016-01-01

    This article is a review of several accelerator–reactor interface issues and nuclear fuel cycle applications of acceleratordriven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systemsmore » on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.« less

  2. Transport properties of fission product vapors

    SciTech Connect

    Im, K.H.; Ahluwalia, R.K.

    1983-07-01

    Kinetic theory of gases is used to calculate the transport properties of fission product vapors in a steam and hydrogen environment. Provided in tabular form is diffusivity of steam and hydrogen, viscosity and thermal conductivity of the gaseous mixture, and diffusivity of cesium iodide, cesium hydroxide, diatomic tellurium and tellurium dioxide. These transport properties are required in determining the thermal-hydraulics of and fission product transport in light water reactors.

  3. Nitrous Oxide Production in a Granule-based Partial Nitritation Reactor: A Model-based Evaluation.

    PubMed

    Peng, Lai; Sun, Jing; Liu, Yiwen; Dai, Xiaohu; Ni, Bing-Jie

    2017-04-03

    Sustainable wastewater treatment has been attracting increasing attentions over the past decades. However, the production of nitrous oxide (N2O), a potent GHG, from the energy-efficient granule-based autotrophic nitrogen removal is largely unknown. This study applied a previously established N2O model, which incorporated two N2O production pathways by ammonia-oxidizing bacteria (AOB) (AOB denitrification and the hydroxylamine (NH2OH) oxidation). The two-pathway model was used to describe N2O production from a granule-based partial nitritation (PN) reactor and provide insights into the N2O distribution inside granules. The model was evaluated by comparing simulation results with N2O monitoring profiles as well as isotopic measurement data from the PN reactor. The model demonstrated its good predictive ability against N2O dynamics and provided useful information about the shift of N2O production pathways inside granules for the first time. The simulation results indicated that the increase of oxygen concentration and granule size would significantly enhance N2O production. The results further revealed a linear relationship between N2O production and ammonia oxidation rate (AOR) (R(2) = 0.99) under the conditions of varying oxygen levels and granule diameters, suggesting that bulk oxygen and granule size may exert an indirect effect on N2O production by causing a change in AOR.

  4. Nitrous Oxide Production in a Granule-based Partial Nitritation Reactor: A Model-based Evaluation

    PubMed Central

    Peng, Lai; Sun, Jing; Liu, Yiwen; Dai, Xiaohu; Ni, Bing-Jie

    2017-01-01

    Sustainable wastewater treatment has been attracting increasing attentions over the past decades. However, the production of nitrous oxide (N2O), a potent GHG, from the energy-efficient granule-based autotrophic nitrogen removal is largely unknown. This study applied a previously established N2O model, which incorporated two N2O production pathways by ammonia-oxidizing bacteria (AOB) (AOB denitrification and the hydroxylamine (NH2OH) oxidation). The two-pathway model was used to describe N2O production from a granule-based partial nitritation (PN) reactor and provide insights into the N2O distribution inside granules. The model was evaluated by comparing simulation results with N2O monitoring profiles as well as isotopic measurement data from the PN reactor. The model demonstrated its good predictive ability against N2O dynamics and provided useful information about the shift of N2O production pathways inside granules for the first time. The simulation results indicated that the increase of oxygen concentration and granule size would significantly enhance N2O production. The results further revealed a linear relationship between N2O production and ammonia oxidation rate (AOR) (R2 = 0.99) under the conditions of varying oxygen levels and granule diameters, suggesting that bulk oxygen and granule size may exert an indirect effect on N2O production by causing a change in AOR. PMID:28367960

  5. Effects of key operational parameters on biohydrogen production via anaerobic fermentation in a sequencing batch reactor.

    PubMed

    Won, S G; Lau, A K

    2011-07-01

    In this study, a series of tests were conducted in a 6L anaerobic sequencing batch reactor (ASBR) to investigate the effect of pH, hydraulic retention time (HRT) and organic loading rate on biohydrogen production at 28°C. Sucrose was used as the main substrate to mimic carbohydrate-rich wastewater and inoculum was prepared from anaerobic digested sludge without pretreatment. The reactor was operated initially with nitrogen sparging to form anaerobic condition. Results showed that methanogens were effectively suppressed. The optimum pH value would vary depending on the HRT. Maximum hydrogen production rate and yield of 3.04 L H(2)/L reactor d and 2.16 mol H(2)/mol hexose respectively were achieved at pH 4.5, HRT 30 h, and OLR 11.0 kg/m(3)d. Two relationships involving the propionic acid/acetic acid ratio and ethanol/acetic acid ratio were derived from the analysis of the metabolites of fermentation. Ethanol/acetic acid ratio of 1.25 was found to be a threshold value for higher hydrogen production.

  6. Microbial community composition and reactor performance during hydrogen production in a UASB reactor fed with raw cheese whey inoculated with compost.

    PubMed

    Castelló, E; Perna, V; Wenzel, J; Borzacconi, L; Etchebehere, C

    2011-01-01

    This study investigated the microbial community developed in a UASB reactor for hydrogen production and correlated it to reactor performance. The reactor was inoculated with kitchen waste compost and fed with raw cheese whey at two organic loading rates, 20 gCOD/Ld and 30 gCOD/Ld. Hydrogen production was very variable, using an OLR of 30 gCOD/Ld averaged 1.0 LH(2)/Ld with no methane produced under these conditions. The hydrogen yield was also very variable and far from the theoretical. This low yield could be explained by selection of a mixed fermentative population with presence of hydrogen producing organisms (Clostridium, Ruminococcus and Enterobacter) and other non-hydrogen producing fermenters (Lactobacillus, Dialister and Prevotella). The molecular analysis of the raw cheese whey used for feeding revealed the presence of three predominant organisms that are affiliated with the genera Buttiauxella (a low-yield hydrogen producer) and Streptococcus (a lactic acid-producing fermenter). Although these organisms did not persist in the reactor, the continuous addition of these fermenters could decrease the reactor's hydrogen yield.

  7. Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

    NASA Technical Reports Server (NTRS)

    Sibille, Laurent; Dominques, Jesus A.

    2012-01-01

    The maturation of Molten Regolith Electrolysis (MRE) as a viable technology for oxygen and metals production on explored planets relies on the realization of the self-heating mode for the reactor. Joule heat generated during regolith electrolysis creates thermal energy that should be able to maintain the molten phase (similar to electrolytic Hall-Heroult process for aluminum production). Self-heating via Joule heating offers many advantages: (1) The regolith itself is the crucible material, it protects the vessel walls (2) Simplifies the engineering of the reactor (3) Reduces power consumption (no external heating) (4) Extends the longevity of the reactor. Predictive modeling is a tool chosen to perform dimensional analysis of a self-heating reactor: (1) Multiphysics modeling (COMSOL) was selected for Joule heat generation and heat transfer (2) Objective is to identify critical dimensions for first reactor prototype.

  8. Feasibility study Part I - Thermal hydraulic analysis of LEU target for {sup 99}Mo production in Tajoura reactor

    SciTech Connect

    Bsebsu, F.M.; Abotweirat, F. E-mail: abutweirat@yahoo.com; Elwaer, S.

    2008-07-15

    The Renewable Energies and Water Desalination Research Center (REWDRC), Libya, will implement the technology for {sup 99}Mo isotope production using LEU foil target, to obtain new revenue streams for the Tajoura nuclear research reactor and desiring to serve the Libyan hospitals by providing the medical radioisotopes. Design information is presented for LEU target with irradiation device and irradiation Beryllium (Be) unit in the Tajoura reactor core. Calculated results for the reactor core with LEU target at different level of power are presented for steady state and several reactivity induced accident situations. This paper will present the steady state thermal hydraulic design and transient analysis of Tajoura reactor was loaded with LEU foil target for {sup 99}Mo production. The results of these calculations show that the reactor with LEU target during the several cases of transient are in safe and no problems will occur. (author)

  9. Modular Hybrid Plasma Reactor for Low Cost Bulk Production of Nanomaterials

    SciTech Connect

    Peter C. Kong

    2011-12-01

    INL developed a bench scale modular hybrid plasma system for gas phase nanomaterials synthesis. The system was being optimized for WO3 nanoparticles production and scale model projection to a 300 kW pilot system. During the course of technology development many modifications had been done to the system to resolve technical issues that had surfaced and also to improve the performance. All project tasks had been completed except 2 optimization subtasks. These 2 subtasks, a 4-hour and an 8-hour continuous powder production runs at 1 lb/hr powder feeding rate, were unable to complete due to technical issues developed with the reactor system. The 4-hour run had been attempted twice and both times the run was terminated prematurely. The modular electrode for the plasma system was significantly redesigned to address the technical issues. Fabrication of the redesigned modular electrodes and additional components had been completed at the end of the project life. However, not enough resource was available to perform tests to evaluate the performance of the new modifications. More development work would be needed to resolve these problems prior to scaling. The technology demonstrated a surprising capability of synthesizing a single phase of meta-stable delta-Al2O3 from pure alpha-phase large Al2O3 powder. The formation of delta-Al2O3 was surprising because this phase is meta-stable and only formed between 973-1073 K, and delta-Al2O3 is very difficult to synthesize as a single phase. Besides the specific temperature window to form this phase, this meta-stable phase may have been stabilized by nanoparticle size formed in a high temperature plasma process. This technology may possess the capability to produce unusual meta-stable nanophase materials that would be otherwise difficult to produce by conventional methods. A 300 kW INL modular hybrid plasma pilot scale model reactor had been projected using the experimental data from PPG Industries 300 kW hot wall plasma reactor. The

  10. Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

    SciTech Connect

    E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

    2008-08-01

    A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540°C and 900°C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

  11. Production of polygalacturonases by Aspergillus oryzae in stirred tank and internal- and external-loop airlift reactors.

    PubMed

    Fontana, Roselei Claudete; da Silveira, Maurício Moura

    2012-11-01

    The production of endo- and exo-polygalacturonase (PG) by Aspergillus oryzae was assessed in stirred tank reactors (STRs), internal-loop airlift reactors (ILARs) and external-loop airlift reactors (ELARs). For STR production, we compared culture media formulated with either pectin (WBE) or partially hydrolyzed pectin. The highest enzyme activities were obtained in medium that contained 50% pectin in hydrolyzed form (WBE5). PG production in the three reactor types was compared for WBE5 and low salt WBE medium, with additional salts added at 48, 60 and 72h (WBES). The ELARs performed better than the ILARs in WBES medium where the exo-PG was the same concentration as for STRs and the endo-PG was 20% lower. These results indicate that PG production is higher under experimental conditions that result in higher cell growth with minimum pH values less than 3.0.

  12. Mass Transfer of Corrosion Products in the Nonisothermal Sodium Loop of a Fast Reactor

    NASA Astrophysics Data System (ADS)

    Varseev, E. V.; Alekseev, V. V.

    2014-11-01

    The mass transfer of the products of corrosion of the steel surface of the sodium loop of a fast nuclear power reactor was investigated for the purpose of optimization of its parameters. The problem of deposition of the corrosion products on the surface of the heat-exchange unit of the indicated loop was considered. Experimental data on the rate of accumulation of deposits in the channel of this unit and results of the dispersion analysis of the suspensions contained in the sodium coolant are presented.

  13. A strategy for intensive production of molybdenum-99 isotopes for nuclear medicine using CANDU reactors.

    PubMed

    Morreale, A C; Novog, D R; Luxat, J C

    2012-01-01

    Technetium-99m is an important medical isotope utilized worldwide in nuclear medicine and is produced from the decay of its parent isotope, molybdenum-99. The online fueling capability and compact fuel of the CANDU(®)(1) reactor allows for the potential production of large quantities of (99)Mo. This paper proposes (99)Mo production strategies using modified target fuel bundles loaded into CANDU fuel channels. Using a small group of channels a yield of 89-113% of the weekly world demand for (99)Mo can be obtained.

  14. Relative fission product yield determination in the USGS TRIGA Mark I reactor

    NASA Astrophysics Data System (ADS)

    Koehl, Michael A.

    Fission product yield data sets are one of the most important and fundamental compilations of basic information in the nuclear industry. This data has a wide range of applications which include nuclear fuel burnup and nonproliferation safeguards. Relative fission yields constitute a major fraction of the reported yield data and reduce the number of required absolute measurements. Radiochemical separations of fission products reduce interferences, facilitate the measurement of low level radionuclides, and are instrumental in the analysis of low-yielding symmetrical fission products. It is especially useful in the measurement of the valley nuclides and those on the extreme wings of the mass yield curve, including lanthanides, where absolute yields have high errors. This overall project was conducted in three stages: characterization of the neutron flux in irradiation positions within the U.S. Geological Survey TRIGA Mark I Reactor (GSTR), determining the mass attenuation coefficients of precipitates used in radiochemical separations, and measuring the relative fission products in the GSTR. Using the Westcott convention, the Westcott flux, modified spectral index, neutron temperature, and gold-based cadmium ratios were determined for various sampling positions in the USGS TRIGA Mark I reactor. The differential neutron energy spectrum measurement was obtained using the computer iterative code SAND-II-SNL. The mass attenuation coefficients for molecular precipitates were determined through experiment and compared to results using the EGS5 Monte Carlo computer code. Difficulties associated with sufficient production of fission product isotopes in research reactors limits the ability to complete a direct, experimental assessment of mass attenuation coefficients for these isotopes. Experimental attenuation coefficients of radioisotopes produced through neutron activation agree well with the EGS5 calculated results. This suggests mass attenuation coefficients of molecular

  15. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-12-31

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  16. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A. ); Blake, J.E.; Rush, G.C. )

    1990-01-01

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  17. Production and modeling of radioactive gold nanoparticles in Tehran research reactor.

    PubMed

    Hosseini, Seyedeh Fatemeh; Sadeghi, Mahdi; Aboudzadeh, Mohammad Reza; Mohseni, Morteza

    2016-12-01

    Gold has two medically useful radioactive isotopes, (198)Au and (199)Au, for locally irradiating and killing tumor cells. (198)Au radionuclide has been produced through the irradiation of the pure gold via (197)Au(n,γ)(198)Au reaction in the Tehran Research Reactor at a thermal neutron flux of 4.5×10(13)ncm(-2)s(-1) for the different irradiation times. In this paper, the activity of (198)Au radionuclide has been determined using MCNPX-2.6 and TALYS-1.6 codes and also the theoretical approach. The calculated results were compared with the corresponding experimental values. The calculated results were in good agreement with the experimental data, thus the used codes can be used as a powerful tool to predict and optimize production conditions in reactor.

  18. A fuzzy-logic-based controller for methane production in anaerobic fixed-film reactors.

    PubMed

    Robles, A; Latrille, E; Ruano, M V; Steyer, J P

    2017-01-01

    The main objective of this work was to develop a controller for biogas production in continuous anaerobic fixed-bed reactors, which used effluent total volatile fatty acids (VFA) concentration as control input in order to prevent process acidification at closed loop. To this aim, a fuzzy-logic-based control system was developed, tuned and validated in an anaerobic fixed-bed reactor at pilot scale that treated industrial winery wastewater. The proposed controller varied the flow rate of wastewater entering the system as a function of the gaseous outflow rate of methane and VFA concentration. Simulation results show that the proposed controller is capable to achieve great process stability even when operating at high VFA concentrations. Pilot results showed the potential of this control approach to maintain the process working properly under similar conditions to the ones expected at full-scale plants.

  19. Ammonium recovery from reject water combined with hydrogen production in a bioelectrochemical reactor.

    PubMed

    Wu, Xue; Modin, Oskar

    2013-10-01

    In this study, a bioelectrochemical reactor was investigated for simultaneous hydrogen production and ammonium recovery from reject water, which is an ammonium-rich side-stream produced from sludge treatment processes at wastewater treatment plants. In the anode chamber of the reactor, microorganisms converted organic material into electrical current. The electrical current was used to generate hydrogen gas at the cathode with 96±6% efficiency. Real or synthetic reject water was fed to the cathode chamber where proton reduction into hydrogen gas resulted in a pH increase which led to ammonium being converted into volatile ammonia. The ammonia could be stripped from the solution and recovered in acid. Overall, ammonium recovery efficiencies reached 94% with synthetic reject water and 79% with real reject water. This process could potentially be used to make wastewater treatment plants more resource-efficient and further research is warranted.

  20. Continuous production of Cu2ZnSnS4 nanocrystals in a flow reactor.

    PubMed

    Shavel, Alexey; Cadavid, Doris; Ibáñez, Maria; Carrete, Alex; Cabot, Andreu

    2012-01-25

    A procedure for the continuous production of Cu(2)ZnSnS(4) (CZTS) nanoparticles with controlled composition is presented. CZTS nanoparticles were prepared through the reaction of the metals' amino complexes with elemental sulfur in a continuous-flow reactor at moderate temperatures (300-330 °C). High-resolution transmission electron microscopy and X-ray diffraction analysis showed the nanocrystals to have a crystallographic structure compatible with that of the kesterite. Chemical characterization of the materials showed the presence of the four elements in each individual nanocrystal. Composition control was achieved by adjusting the solution flow rate through the reactor and the proper choice of the nominal precursor concentration within the flowing solution. Single-particle analysis revealed a composition distribution within each sample, which was optimized at the highest synthesis temperatures used.

  1. Assessment of fission product yields data needs in nuclear reactor applications

    SciTech Connect

    Kern, K.; Becker, M.; Broeders, C.

    2012-07-01

    Studies on the build-up of fission products in fast reactors have been performed, with particular emphasis on the effects related to the physics of the nuclear fission process. Fission product yields, which are required for burn-up calculations, depend on the proton and neutron number of the target nucleus as well as on the incident neutron energy. Evaluated nuclear data on fission product yields are available for all relevant target nuclides in reactor applications. However, the description of their energy dependence in evaluated data is still rather rudimentary, which is due to the lack of experimental fast fission data and reliable physical models. Additionally, physics studies of evaluated JEFF-3.1.1 fission yields data have shown potential improvements, especially for various fast fission data sets of this evaluation. In recent years, important progress in the understanding of the fission process has been made, and advanced model codes are currently being developed. This paper deals with the semi-empirical approach to the description of the fission process, which is used in the GEF code being developed by K.-H. Schmidt and B. Jurado on behalf of the OECD Nuclear Energy Agency, and with results from the corresponding author's diploma thesis. An extended version of the GEF code, supporting the calculation of spectrum weighted fission product yields, has been developed. It has been applied to the calculation of fission product yields in the fission rate spectra of a MOX fuelled sodium-cooled fast reactor. Important results are compared to JEFF-3.1.1 data and discussed in this paper. (authors)

  2. Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors.

    PubMed

    Zhang, Yifeng; Angelidaki, Irini

    2012-05-15

    A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 mL/L/d at the initial acetate concentration of 410 mg/L (5 mM), while the cathodic hydrogen recovery ( [Formula: see text] ) and overall systemic coulombic efficiency (CE(os)) were 93% and 28%, respectively, and the systemic hydrogen yield ( [Formula: see text] ) peaked at 1.27 mol-H(2)/mol-acetate. The hydrogen production increased along with acetate and buffer concentration. The highest hydrogen production rate of 32.2 mL/L/d and [Formula: see text] of 1.43 mol-H(2)/mol-acetate were achieved at 1640 mg/L (20 mM) acetate and 100 mM phosphate buffer. Further evaluation of the reactor under single electricity-generating or hydrogen-producing mode indicated that further improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens. Furthermore, 16S rRNA genes analysis showed that this special operation strategy resulted same microbial community structures in the anodic biofilms of the two cell units. The simple, compact and in situ applicable SMEC offers new opportunities for reactor design for a microbial electricity-assisted biohydrogen production system.

  3. Production of volatile fatty acids from wastewater screenings using a leach-bed reactor.

    PubMed

    Cadavid-Rodríguez, Luz Stella; Horan, Nigel J

    2014-09-01

    Screenings recovered from the inlet works of wastewater treatment plants were digested without pre-treatment or dilution using a lab-scale, leach-bed reactor. Variations in recirculation ratio of the leachate of 4 and 8 l/lreactor/day and pH values of 5 and 6 were evaluated in order to determine the optimal operating conditions for maximum total volatile fatty acids (VFA) production. By increasing the recirculation ratio of the leachate from 4 to 8 l/lreactor/day it was possible to increase VFA production (11%) and soluble COD (17%) and thus generate up to 264 g VFA/kg-dry screenings. These VFA were predominantly acetic acid with some propionic and butyric acid. The optimum pH for VFA production was 6.0, when the methanogenic phase was inhibited. Below pH 5.0, acid-producing fermentation was inhibited and some alcohols were produced. Ammonia release during the hydrolysis of screenings provided adequate alkalinity; consequently, a digestion process without pH adjustment could be recommended. The leach-bed reactor was able to achieve rapid rates of screenings degradation with the production of valuable end-products that will reduce the carbon footprint associated with current screenings disposal techniques.

  4. Non-catalytic alcoholysis process for production of biodiesel fuel by using bubble column reactor

    NASA Astrophysics Data System (ADS)

    Hagiwara, S.; Nabetani, H.; Nakajima, M.

    2015-04-01

    Biodiesel fuel is a replacement for diesel as a fuel produced from biomass resources. It is usually defined as a fatty acid methyl ester (FAME) derived from vegetable oil or animal fat. In European countries, such as Germany and France, biodiesel fuel is commercially produced mainly from rapeseed oil, whereas in the United States and Argentina, soybean oil is more frequently used. In many other countries such as Japan and countries in Southeast Asia, lipids that cannot be used as a food source could be more suitable materials for the production of biodiesel fuel because its production from edible oils could result in an increase in the price of edible oils, thereby increasing the cost of some foodstuffs. Therefore, used edible oil, lipids contained in waste effluent from the oil milling process, byproducts from oil refining process and crude oils from industrial crops such as jatropha could be more promising materials in these countries. The materials available in Japan and Southeast Asia for the production of biodiesel fuel have common characteristics; they contain considerable amount of impurities and are high in free fatty acids (FFA). Superheated methanol vapor (SMV) reactor might be a promising method for biodiesel fuel production utilizing oil feedstock containing FFA such as waste vegetable oil and crude vegetable oil. In the conventional method using alkaline catalyst, FFA contained in waste vegetable oil is known to react with alkaline catalyst such as NaOH and KOH generating saponification products and to inactivate it. Therefore, the FFA needs to be removed from the feedstock prior to the reaction. Removal of the alkaline catalyst after the reaction is also required. In the case of the SMV reactor, the processes for removing FFA prior to the reaction and catalyst after the reaction can be omitted because it requires no catalyst. Nevertheless, detailed study on the productivity of biodiesel fuel produced from waste vegetable oils and other non

  5. Fermentative hydrogen production in packed-bed and packing-free upflow reactors.

    PubMed

    Li, C; Zhang, T; Fang, H H P

    2006-01-01

    Fermentative hydrogen production from a synthetic wastewater containing 10 g/L of sucrose was studied in two upflow reactors at 26 degrees C for 400 days. One reactor was filled with packing rings (RP) and the other was packing free (RF). The effect of hydraulic retention time (HRT) from 2 h to 24 h was investigated. Results showed that, under steady state, the hydrogen production rate significantly increased from 0.63 L/L/d to 5.35 L/L/d in the RF when HRT decreased from 24 h to 2 h; the corresponding rates were 0.56 L/L/d to 6.17 L/L/d for the RP. In the RF, the hydrogen yield increased from 0.96 mol/mol-sucrose at 24 h of HRT to the maximum of 1.10 mol/mol-sucrose at 8 h of HRT, and then decreased to 0.68 mol/mol-sucrose at 2 h. In the RP, the yield increased from 0.86 mol/mol-sucrose at 24 h of HRT to the maximum of 1.22 mol/mol-sucrose at 14 h of HRT, and then decreased to 0.78 mol/mol-sucrose at 2 h. Overall, the reactor with packing was more effective than the one free of packing. In both reactors, sludge agglutinated into granules. The microbial community of granular sludge in RP was investigated using 16S rDNA based techniques. The distribution of bacterial cells and extracellular polysaccharides in hydrogen-producing granules was investigated by fluorescence-based techniques. Results indicated that most of the N-acetyl-galactosamine/galactose-containing extracellular polysaccharides were distributed on the outer layer of the granules with a filamentous structure.

  6. Accuracy Based Generation of Thermodynamic Properties for Light Water in RELAP5-3D

    SciTech Connect

    Cliff B. Davis

    2010-09-01

    RELAP5-3D interpolates to obtain thermodynamic properties for use in its internal calculations. The accuracy of the interpolation was determined for the original steam tables currently used by the code. This accuracy evaluation showed that the original steam tables are generally detailed enough to allow reasonably accurate interpolations in most areas needed for typical analyses of nuclear reactors cooled by light water. However, there were some regions in which the original steam tables were judged to not provide acceptable accurate results. Revised steam tables were created that used a finer thermodynamic mesh between 4 and 21 MPa and 530 and 640 K. The revised steam tables solved most of the problems observed with the original steam tables. The accuracies of the original and revised steam tables were compared throughout the thermodynamic grid.

  7. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    SciTech Connect

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  8. Scale-up and performance of an immobilized cell reactor separator for the production of ethanol from whey lactose

    SciTech Connect

    Dale, M.C.; Okos, M.R.

    1988-01-01

    An adsorbed microbial cell, trickle flow, reactor incorporating a circulating gas phase to simultaneously separate a volatile inhibitory product has shown high fermentation rates and effective product separation on a 1 in. diameter lab bench scale. Carbon dioxide generated by the fermentation is recirculated to strip the volatile fermentation product from the fermentation broth. Random and structured packings were evaluated and a parallel type structured packing was developed consisting of a plates of absorbant material to which cells are adsorbed and spaced to allow passage of the circulating gas stream. The substrate solution flows by gravity in a capillary flow fashion down through the plates in which the cells are immobilized. The theoretical reactor separation and fermentation performance as a function of plate thickness and gas channel spacing was evaluated. This reactor has been scaled up to a 6 in. ID reactor with a total reactor volume of about 50 liters using a spiral wound version of the parallel plates, and then to a 24 in. ID reactor with a reactor volume of about 1300 liters. 16 refs., 12 figs., 1 tab.

  9. Biohydrogen production from waste bread in a continuous stirred tank reactor: A techno-economic analysis.

    PubMed

    Han, Wei; Hu, Yun Yi; Li, Shi Yi; Li, Fei Fei; Tang, Jun Hong

    2016-12-01

    Biohydrogen production from waste bread in a continuous stirred tank reactor (CSTR) was techno-economically assessed. The treating capacity of the H2-producing plant was assumed to be 2 ton waste bread per day with lifetime of 10years. Aspen Plus was used to simulate the mass and energy balance of the plant. The total capital investment (TCI), total annual production cost (TAPC) and annual revenue of the plant were USD931020, USD299746/year and USD639920/year, respectively. The unit hydrogen production cost was USD1.34/m(3) H2 (or USD14.89/kg H2). The payback period and net present value (NPV) of the plant were 4.8years and USD1266654, respectively. Hydrogen price and operators cost were the most important variables on the NPV. It was concluded that biohydrogen production from waste bread in the CSTR was feasible for practical application.

  10. Semicontinuous Production of Lactic Acid From Cheese Whey Using Integrated Membrane Reactor

    NASA Astrophysics Data System (ADS)

    Li, Yebo; Shahbazi, Abolghasem; Coulibaly, Sekou; Mims, Michele M.

    Semicontinuous production of lactic acid from cheese whey using free cells of Bifidobacterium longum with and without nanofiltration was studied. For the semicontinuous fermentation without membrane separation, the lactic acid productivity of the second and third runs is much lower than the first run. The semicontinuous fermentation with nanoseparation was run semicontinuously for 72 h with lactic acid to be harvested every 24 h using a nanofiltration membrane unit. The cells and unutilized lactose were kept in the reactor and mixed with newly added cheese whey in the subsequent runs. Slight increase in the lactic acid productivity was observed in the second and third runs during the semicontinuous fermentation with nanofiltration. It can be concluded that nanoseparation could improve the lactic acid productivity of the semicontinuous fermentation process.

  11. A model for nonvolatile fission product release during reactor accident conditions

    SciTech Connect

    Lewis, B.J.; Andre, B.; Ducros, G.; Maro, D.

    1996-10-01

    An analytical model has been developed to describe the release kinetics of nonvolatile fission products (e.g., molybdenum, cerium, ruthenium, and barium) from uranium dioxide fuel under severe reactor accident conditions. This treatment considers the rate-controlling process of release in accordance with diffusional transport in the fuel matrix and fission product vaporization from the fuel surface into the surrounding gas atmosphere. The effect of the oxygen potential in the gas atmosphere on the chemical form and volatility of the fission product is considered. A correlation is also developed to account for the trapping effects of antimony and tellurium in the Zircaloy cladding. This model interprets the release behavior of fission products observed in Commissariat a l`Energie Atomique experiments conducted in the HEVA/VERCORS facility at high temperature in a hydrogen and steam atmosphere.

  12. Hybrid adsorptive membrane reactor

    NASA Technical Reports Server (NTRS)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  13. Hybrid adsorptive membrane reactor

    SciTech Connect

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  14. Modelling Methane Production and Sulfate Reduction in Anaerobic Granular Sludge Reactor with Ethanol as Electron Donor

    NASA Astrophysics Data System (ADS)

    Sun, Jing; Dai, Xiaohu; Wang, Qilin; Pan, Yuting; Ni, Bing-Jie

    2016-10-01

    In this work, a mathematical model based on growth kinetics of microorganisms and substrates transportation through biofilms was developed to describe methane production and sulfate reduction with ethanol being a key electron donor. The model was calibrated and validated using experimental data from two case studies conducted in granule-based Upflow Anaerobic Sludge Blanket reactors. The results suggest that the developed model could satisfactorily describe methane and sulfide productions as well as ethanol and sulfate removals in both systems. The modeling results reveal a stratified distribution of methanogenic archaea, sulfate-reducing bacteria and fermentative bacteria in the anaerobic granular sludge and the relative abundances of these microorganisms vary with substrate concentrations. It also indicates sulfate-reducing bacteria can successfully outcompete fermentative bacteria for ethanol utilization when COD/SO42‑ ratio reaches 0.5. Model simulation suggests that an optimal granule diameter for the maximum methane production efficiency can be achieved while the sulfate reduction efficiency is not significantly affected by variation in granule size. It also indicates that the methane production and sulfate reduction can be affected by ethanol and sulfate loading rates, and the microbial community development stage in the reactor, which provided comprehensive insights into the system for its practical operation.

  15. Modelling Methane Production and Sulfate Reduction in Anaerobic Granular Sludge Reactor with Ethanol as Electron Donor

    PubMed Central

    Sun, Jing; Dai, Xiaohu; Wang, Qilin; Pan, Yuting; Ni, Bing-Jie

    2016-01-01

    In this work, a mathematical model based on growth kinetics of microorganisms and substrates transportation through biofilms was developed to describe methane production and sulfate reduction with ethanol being a key electron donor. The model was calibrated and validated using experimental data from two case studies conducted in granule-based Upflow Anaerobic Sludge Blanket reactors. The results suggest that the developed model could satisfactorily describe methane and sulfide productions as well as ethanol and sulfate removals in both systems. The modeling results reveal a stratified distribution of methanogenic archaea, sulfate-reducing bacteria and fermentative bacteria in the anaerobic granular sludge and the relative abundances of these microorganisms vary with substrate concentrations. It also indicates sulfate-reducing bacteria can successfully outcompete fermentative bacteria for ethanol utilization when COD/SO42− ratio reaches 0.5. Model simulation suggests that an optimal granule diameter for the maximum methane production efficiency can be achieved while the sulfate reduction efficiency is not significantly affected by variation in granule size. It also indicates that the methane production and sulfate reduction can be affected by ethanol and sulfate loading rates, and the microbial community development stage in the reactor, which provided comprehensive insights into the system for its practical operation. PMID:27731395

  16. Enhanced Hydrogen Production Integrated with CO2 Separation in a Single-Stage Reactor

    SciTech Connect

    Mahesh Iyer; Himanshu Gupta; Danny Wong; Liang-Shih Fan

    2005-09-30

    Hydrogen production from coal gasification can be enhanced by driving the equilibrium limited Water Gas Shift reaction forward by incessantly removing the CO{sub 2} by-product via the carbonation of calcium oxide. This project aims at using the OSU patented high-reactivity mesoporous precipitated calcium carbonate sorbent for removing the CO{sub 2} product. Preliminary experiments demonstrate the show the superior performance of the PCC sorbent over other naturally occurring calcium sorbents. Gas composition analyses show the formation of 100% pure hydrogen. Novel calcination techniques could lead to smaller reactor footprint and single-stage reactors that can achieve maximum theoretical H{sub 2} production for multicyclic applications. Sub-atmospheric calcination studies reveal the effect of vacuum level, diluent gas flow rate, thermal properties of the diluent gas and the sorbent loading on the calcination kinetics which play an important role on the sorbent morphology. Steam, which can be easily separated from CO{sub 2}, is envisioned to be a potential diluent gas due to its enhanced thermal properties. Steam calcination studies at 700-850 C reveal improved sorbent morphology over regular nitrogen calcination. A mixture of 80% steam and 20% CO{sub 2} at ambient pressure was used to calcine the spent sorbent at 820 C thus lowering the calcination temperature. Regeneration of calcium sulfide to calcium carbonate was achieved by carbonating the calcium sulfide slurry by bubbling CO{sub 2} gas at room temperature.

  17. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  18. Membrane reactor microstructure for polymer grade olefin production and hydrogen cogeneration.

    PubMed

    Ziaka, Zoe D; Navrozidou, Anna; Paraschopoulou, Louisa; Vasileiadis, Savvas P

    2010-09-01

    The current communication describes research work on effective membrane reactor nanostructures and nanoreaction-nanoseparation technology for polymer grade olefin production via catalytic paraffin dehydrogenation reactions. Emphasis is given in systems of permreactors and permeators to perform the described reactive and separative operations. We elaborate on new membrane microstructure designs for paraffin dehydrogenations including the design of experiments, operation, and best parameter selection and optimization of such systems. The described processes are of current significance in the area of new microreactor design and operation including hydrocarbon processing and conversion to valuable fuels and chemicals such as hydrogen, olefins, and polyolefins. These improved results are because of the unique design characteristics of the examined microreactor systems to perform accurate multiphase and heterogeneous functions into one unit operation. A number of membrane reactor configurations were made and tested on stream for the catalytic propane dehydrogenation reaction to propylene with successful results. Some of the results are discussed below which show the better performance of nanostructured membrane reactors for the specific dehydrogenation.

  19. Computational and experimental prediction of dust production in pebble bed reactors -- Part I

    SciTech Connect

    Maziar Rostamian; Gannon Johnson; Mie Hiruta; Gabriel P. Potirniche; Abderrafi M. Ougouag; Joshua J. Cogliati; Akira Tokuhiro

    2013-10-01

    This paper describes the computational modeling and simulation, and experimental testing of graphite moderators in frictional contacts as anticipated in a pebble bed reactor. The potential of carbonaceous particulate generation due to frictional contact at the surface of pebbles and the ensuing entrainment and transport into the gas coolant are safety concerns at elevated temperatures under accident scenarios such as air ingress in the high temperature gas-cooled reactor. The safety concerns are due to the documented ability of carbonaceous particulates to adsorb fission products and transport them in the primary circuit of the pebble bed reactor, thus potentially giving rise to a relevant source term under accident scenarios. Here, a finite element approach is implemented to develop a nonlinear wear model in air environment. In this model, material wear coefficient is related to the changes in asperity height during wear. The present work reports a comparison between the finite element simulations and the experimental results obtained using a custom-designed tribometer. The experimental and computational results are used to estimate the quantity of nuclear grade graphite dust produced from a typical anticipated configuration. In Part II, results from a helium environment at higher temperatures and pressures are experimentally studied.

  20. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John b. Walter

    2010-10-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  1. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  2. Study and comparison of two enzyme membrane reactors for fatty acids and glycerol production

    SciTech Connect

    Molinari, R.; Santoro, M.E.; Drioli, E. . Dept. of Chemical Engineering and Materials Inst. on Membranes and Chemical Reactors-CNR, Arcavacata di Rende )

    1994-11-01

    Two enzyme membrane reactors (EMR), (1) with one substrate (olive oil) in an oil-in-water emulsion (E-EMR) and (2) with two separated liquid phases (oil and water) (TSLP-EMR), have been studied for the conversion of the triglycerides to fatty acids and glycerol. The enzyme was Candida cylindracea lipase confined on the pressurized face or entrapped in the sponge side of capillary ultrafiltration membranes. Two methods for immobilizing the enzyme in the TSLP-EMR were used: ultrafiltration on a virgin membrane and ultrafiltration on glutaraldehyde pretreated membranes. A multiple use of the reactor was obtained immobilizing the enzyme on the membrane preactivated with glutaraldehyde. The TSLP-EMR showed a specific activity of 0.529 mmol/(mg[center dot]h) versus a specific activity of 0.170 mmol/(mg[center dot]h) of the E-EMR. The rate of fatty acid production in the TSLP-EMR was linear with time showing no enzyme deactivation in an operating time of 80 h. The kinetics observed in the two reactors was different: an equilibrium reaction product-inhibited for the E-EMR and an apparent irreversible reaction of zero order for the TSLP-EMR. Taking into account that in the TSLP-EMR, compared to the E-EMR, (1) the specific activity was higher, (2) the specific rate was constant with the time, and (3) the two products were already separated after the reaction, the TSLP-EMR configuration seems the more convenient.

  3. High-performance recombinant protein production with Escherichia coli in continuously operated cascades of stirred-tank reactors.

    PubMed

    Schmideder, Andreas; Weuster-Botz, Dirk

    2017-03-01

    The microbial expression of intracellular, recombinant proteins in continuous bioprocesses suffers from low product concentrations. Hence, a process for the intracellular production of photoactivatable mCherry with Escherichia coli in a continuously operated cascade of two stirred-tank reactors was established to separate biomass formation (first reactor) and protein expression (second reactor) spatially. Cascades of miniaturized stirred-tank reactors were implemented, which enable the 24-fold parallel characterization of cascade processes and the direct scale-up of results to the liter scale. With PAmCherry concentrations of 1.15 g L(-1) cascades of stirred-tank reactors improved the process performance significantly compared to production processes in chemostats. In addition, an optimized fed-batch process was outperformed regarding space-time yield (149 mg L(-1) h(-1)). This study implicates continuous cascade processes to be a promising alternative to fed-batch processes for microbial protein production and demonstrates that miniaturized stirred-tank reactors can reduce the timeline and costs for cascade process characterization.

  4. Environmental characterization of two potential locations at Hanford for a new production reactor

    SciTech Connect

    Watson, E.C.; Becker, C.D.; Fitzner, R.E.; Gano, K.A.; Imhoff, K.L.; McCallum, R.F.; Myers, D.A.; Page, T.L.; Price, K.R.; Ramsdell, J.V.; Rice D.G.; Schreiber D.L.; Skumatz L.A.; Sommer D.J.; Tawil J.J.; Wallace R.W.; Watson D.G.

    1984-09-01

    This report describes various environmental aspects of two areas on the Hanford Site that are potential locations for a New Production Reactor (NPR). The area known as the Skagit Hanford Site is considered the primary or reference site. The second area, termed the Firehouse Site, is considered the alternate site. The report encompasses an environmental characterization of these two potential NPR locations. Eight subject areas are covered: geography and demography; ecology; meteorology; hydrology; geology; cultural resources assessment; economic and social effects of station construction and operation; and environmental monitoring. 80 refs., 68 figs., 109 tabs.

  5. Functionally gradient material for membrane reactors to convert methane gas into value-added products

    DOEpatents

    Balachandran, U.; Dusek, J.T.; Kleefisch, M.S.; Kobylinski, T.P.

    1996-11-12

    A functionally gradient material for a membrane reactor for converting methane gas into value-added-products includes an outer tube of perovskite, which contacts air; an inner tube which contacts methane gas, of zirconium oxide, and a bonding layer between the perovskite and zirconium oxide layers. The bonding layer has one or more layers of a mixture of perovskite and zirconium oxide, with the layers transitioning from an excess of perovskite to an excess of zirconium oxide. The transition layers match thermal expansion coefficients and other physical properties between the two different materials. 7 figs.

  6. Functionally gradient material for membrane reactors to convert methane gas into value-added products

    DOEpatents

    Balachandran, Uthamalingam; Dusek, Joseph T.; Kleefisch, Mark S.; Kobylinski, Thadeus P.

    1996-01-01

    A functionally gradient material for a membrane reactor for converting methane gas into value-added-products includes an outer tube of perovskite, which contacts air; an inner tube which contacts methane gas, of zirconium oxide, and a bonding layer between the perovskite and zirconium oxide layers. The bonding layer has one or more layers of a mixture of perovskite and zirconium oxide, with the layers transitioning from an excess of perovskite to an excess of zirconium oxide. The transition layers match thermal expansion coefficients and other physical properties between the two different materials.

  7. Analysis of fission product revaporization in a BWR Reactor Coolant System during a station blackout accident

    SciTech Connect

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This paper presents an analysis of fission product revaporization from the Reactor Coolant System (RCS) following the Reactor Pressure Vessel (RPV) failure. The station blackout accident in a BWR Mark I Power Plant was considered. The TRAPMELT3 models for vaporization, chemisorption, and the decay heating of RCS structures and gases were used and extended beyond the RPV failure in the analysis. The RCS flow models based on the density-difference or pressure-difference between the RCS and containment pedestal region were developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP was developed for the analysis. The REVAP code was incorporated with the MARCH, TRAPMELT3 and NAUA codes from the Source Term Code Package (STCP) to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors in determining the magnitude of revaporization and subsequent release of the volatile fission product into the environment. 6 refs., 8 figs.

  8. A self-catalytic mixed-conducting membrane reactor for effective production of hydrogen from methane

    NASA Astrophysics Data System (ADS)

    Dong, Xueliang; Liu, Zhengkun; Jin, Wanqin; Xu, Nanping

    A supported mixed-conducting LNCO membrane with a pore-gradient structure LNO support was successfully prepared via a dry pressing and co-sintering route. The match of sintering behaviors between membrane and support was realized by the preparation of an ABB‧B‧‧B‧BA-type membrane. A scanning electron microscopy (SEM) test demonstrated that the surface of the supported membrane was dense and crack-free and the pore-gradient structure of the support can be observed clearly. The oxygen flux of the supported membrane was about 5.6 times that of the symmetric LNCO membrane. A self-catalytic mixed-conducting membrane reactor was constructed using the prepared membrane for hydrogen production from methane. It was found that this membrane reactor exhibited high performance and good stability for hydrogen production. At 1123 K, the CH 4 conversion, hydrogen selectivity and hydrogen production remained at about 60%, 89% and 8.0 ml(STP) cm -2 min -1, respectively, for more than 120 h.

  9. The kinetics of nitrogen removal and biogas production in an anammox non-woven membrane reactor.

    PubMed

    Ni, Shou-Qing; Lee, Po-Heng; Sung, Shihwu

    2010-08-01

    The anammox non-woven membrane reactor (ANMR) is a novel reactor configuration to culture the slowly growing anammox bacteria. Different mathematical models were used to study the process kinetics of the nitrogen removal in the ANMR. The kinetics of nitrogen gas production of anammox process was first evaluated in this paper. For substrate removal kinetics, the modified Stover-Kincannon model and the Grau second-order model were more applicable to the ANMR than the first-order model and the Monod model. For nitrogen gas production kinetics, the Van der Meer and Heertjes model was more appropriate than the modified Stover-Kincannon model. Model evaluation was carried out by comparing experimental data with predicted values calculated from suitable models. Both model kinetics study and model testing showed that the Grau second-order model and the Van der Meer and Heertjes model seemed to be the best models to describe the nitrogen removal and nitrogen gas production in the ANMR, respectively.

  10. Conceptual Design of Low-Temperature Hydrogen Production and High-Efficiency Nuclear Reactor Technology

    NASA Astrophysics Data System (ADS)

    Fukushima, Kimichika; Ogawa, Takashi

    Hydrogen, a potential alternative energy source, is produced commercially by methane (or LPG) steam reforming, a process that requires high temperatures, which are produced by burning fossil fuels. However, as this process generates large amounts of CO2, replacement of the combustion heat source with a nuclear heat source for 773-1173K processes has been proposed in order to eliminate these CO2 emissions. In this paper, a novel method of nuclear hydrogen production by reforming dimethyl ether (DME) with steam at about 573K is proposed. From a thermodynamic equilibrium analysis of DME steam reforming, the authors identified conditions that provide high hydrogen production fraction at low pressure and temperatures of about 523-573K. By setting this low-temperature hydrogen production process upstream from a turbine and nuclear reactor at about 573K, the total energy utilization efficiency according to equilibrium mass and heat balance analysis is about 50%, and it is 75%for a fast breeder reactor (FBR), where turbine is upstream of the reformer.

  11. Catalog of experimental projects for a fissioning plasma reactor

    NASA Technical Reports Server (NTRS)

    Lanzo, C. D.

    1973-01-01

    Experimental and theoretical investigations were carried out to determine the feasibility of using a small scale fissioning uranium plasma as the power source in a driver reactor. The driver system is a light water cooled and moderated reactor of the MTR type. The eight experiments and proposed configurations for the reactor are outlined.

  12. Advanced Reactors Thermal Energy Transport for Process Industries

    SciTech Connect

    P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

    2014-07-01

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

  13. Method of production H/sub 2/ using a rotating drum reactor with a pulse jet heat source

    DOEpatents

    Paulson, L.E.

    1988-05-13

    A method of producing hydrogen by an endothermic steam-carbon reaction using a rotating drum reactor and a pulse jet combustor. The pulse jet combustor uses coal dust as a fuel to provide reaction temperatures of 1300/degree/ to 1400/degree/F. Low-rank coal, water, limestone and catalyst are fed into the drum reactor where they are heated, tumbled and reacted. Part of the reaction product from the rotating drum reactor is hydrogen which can be utilized in suitable devices. 1 fig.

  14. Isotope production target irradiation experience at the annular core research reactor

    SciTech Connect

    Talley, D.G.

    1997-02-01

    As a result of an Environmental Impact Statement (EIS) recently issued by the Department of Energy, Sandia National Laboratories (SNL) has been selected as the {open_quotes}most appropriate facility{close_quotes} for the production of {sup 99}Mo. The daughter product of {sup 99}Mo is {sup 99m}Tc, a radioisotope used in 36,000 medical procedures per day in the U.S.{close_quote} At SNL, the {sup 99}Mo would be created by the fission process in UO{sub 2} coated {open_quotes}targets{close_quotes} and chemically separated in the SNL Hot Cell Facility (HCF). SNL has recently completed the irradiation of five production targets at its Annular Core Research Reactor (ACRR). Following irradiation, four of the targets were chemically processed in the HCF using the Cintichem process.

  15. Assessing optimal fermentation type for bio-hydrogen production in continuous-flow acidogenic reactors.

    PubMed

    Ren, N Q; Chua, H; Chan, S Y; Tsang, Y F; Wang, Y J; Sin, N

    2007-07-01

    In this study, the optimal fermentation type and the operating conditions of anaerobic process in continuous-flow acidogenic reactors was investigated for the maximization of bio-hydrogen production using mixed cultures. Butyric acid type fermentation occurred at pH>6, propionic acid type fermentation occurred at pH about 5.5 with E(h) (redox potential) >-278mV, and ethanol-type fermentation occurred at pH<4.5. The representative strains of these fermentations were Clostridium sp., Propionibacterium sp. and Bacteriodes sp., respectively. Ethanol fermentation was optimal type by comparing the operating stabilities and hydrogen production capacities between the fermentation types, which remained stable when the organic loading rate (OLR) reached the highest OLR at 86.1kgCOD/m(3)d. The maximum hydrogen production reached up to 14.99L/d.

  16. Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

    SciTech Connect

    Wegst, Ulrike G.K.; Allen, Todd; Sridharan, Kumar

    2014-04-07

    Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

  17. Process development and modeling of fluidized-bed reactor with coimmobilized biocatalyst for fuel ethanol production

    NASA Astrophysics Data System (ADS)

    Sun, May Yongmei

    This research focuses on two steps of commercial fuel ethanol production processes: the hydrolysis starch process and the fermentation process. The goal of this research is to evaluate the performance of co-immobilized biocatalysts in a fluidized bed reactor with emphasis on economic and engineering aspects and to develop a predictive mathematical model for this system. The productivity of an FBR is higher than productivity of a traditional batch reactor or CSTR. Fluidized beds offer great advantages over packed beds for immobilized cells when small particles are used or when the reactant feed contains suspended solids. Plugging problems, excessive pressure drops (and thus attrition), or crushing risks may be avoided. No mechanical stirring is required as mixing occurs due to the natural turbulence in the fluidized process. Both enzyme and microorganism are immobilized in one catalyst bead which is called co-immobilization. Inside this biocatalyst matrix, starch is hydrolyzed by the enzyme glucoamylase to form glucose and then converted to ethanol and carbon dioxide by microorganisms. Two biocatalysts were evaluated: (1) co-immobilized yeast strain Saccharomyces cerevisiae and glucoamylase. (2) co-immobilized Zymomonas mobilis and glucoamylase. A co-immobilized biocatalyst accomplishes the simultaneous saccharification and fermentation (SSF process). When compared to a two-step process involving separate saccharification and fermentation stages, the SSF process has productivity values twice that given by the pre-saccharified process when the time required for pre-saccharification (15--25 h) was taken into account. The SSF process should also save capital cost. The information about productivity, fermentation yield, concentration profiles along the bed, ethanol inhibition, et al., was obtained from the experimental data. For the yeast system, experimental results showed that: no apparent decrease of productivity occurred after two and half months, the productivity

  18. A physical description of fission product behavior fuels for advanced power reactors.

    SciTech Connect

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  19. Hydrogen production by plasma electrolysis reactor of KOH-ethanol solution

    NASA Astrophysics Data System (ADS)

    Saksono, N.; Batubara, T.; Bismo, S.

    2016-11-01

    Plasma electrolysis has great potential in industrial hydrogen production, chlor-alkali production, and waste water treatment. Plasma electrolysis produces more hydrogen with less energy consumption than hydrocarbon or Faraday electrolysis. This paper investigated the hydrogen production by plasma electrolysis of KOH-ethanol solution at 80 °C and 1 atm. The effects of voltage, KOH solution, ethanol addition, and cathode deep on plasma electrolysis performance were studied. The hydrogen production was analyzed using bubble flow meter and hydrogen analyzer. The electrical energy consumption was measured by a digital multimeter. The effectiveness of plasma electrolysis in terms of hydrogen production was evaluated by comparing it with Faraday Electrolysis. The results showed that hydrogen produced by plasma electrolysis is 149 times higher than the hydrogen produced by Faraday electrolysis. The optimum hydrogen production was 50.71 mmol/min, obtained at 700 V with 0.03 M KOH, 10% vol ethanol and 6.6 cm cathode deep, with energy consumption 1.49 kJ/mmol. The result demonstrates a promising path for hydrogen production by utilizing plasma electrolysis reactor.

  20. Azadirachtin production by hairy root cultivation of Azadirachta indica in a modified stirred tank reactor.

    PubMed

    Srivastava, Smita; Srivastava, A K

    2012-11-01

    Present investigation involves hairy root cultivation of Azadirachta indica in a modified stirred tank reactor under optimized culture conditions for maximum volumetric productivity of azadirachtin. The selected hairy root line (Az-35) was induced via Agrobacterium rhizogenes LBA 920-mediated transformation of A. indica leaf explants (Coimbatore variety, India). Liquid culture of the hairy roots was developed in a modified Murashige and Skoog medium (MM2). To further enhance the productivity of azadirachtin, selected growth regulators (1.0 mg/l IAA and 0.025 mg/l GA(3)), permeabilizing agent (0.5 % v/v DNBP), a biotic elicitor (1 % v/v Curvularia (culture filtrate)) and an indirectly linked biosynthetic precursor (50 mg/l cholesterol) were added in the growth medium on 15th day of the hairy root cultivation period in shake flask. Highest azadirachtin production (113 mg/l) was obtained on 25th day of the growth cycle with a biomass of 21 g/l DW. Further, batch cultivation of hairy roots was carried out in a novel liquid-phase bioreactor configuration (modified stirred tank reactor with polyurethane foam as root support) to investigate the possible scale-up of the established A. indica hairy root culture. A biomass production of 15.2 g/l with azadirachtin accumulation in the hairy roots of 6.4 mg/g (97.28 mg/l) could be achieved after 25 days of the batch cultivation period, which was ~27 and ~14 % less biomass and azadirachtin concentration obtained respectively, in shake flasks. An overall volumetric productivity of 3.89 mg/(l day) of azadirachtin was obtained in the bioreactor.