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Sample records for fast ceramic reactor

  1. A US perspective on fast reactor fuel fabrication technology and experience. Part II: Ceramic fuels

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Fielding, Randall S.; Porter, Douglas L.; Meyer, Mitchell K.; Makenas, Bruce J.

    2009-08-01

    This paper is Part II of a review focusing on the United States experience with oxide, carbide, and nitride fast reactor fuel fabrication. Over 60 years of research in fuel fabrication by government, national laboratories, industry, and academia has culminated in a foundation of research and resulted in significant improvements to the technologies employed to fabricate these fuel types. This part of the review documents the current state of fuel fabrication technologies in the United States for each of these fuel types, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  2. 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.

  3. The Integral Fast Reactor

    SciTech Connect

    Till, C.E.; Chang, Y.I. ); Lineberry, M.J. )

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

  4. The Integral Fast Reactor

    SciTech Connect

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab.

  5. Integral Fast Reactor concept

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  6. Fast quench reactor method

    SciTech Connect

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.; Berry, Ray A.

    1999-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

  7. Fast quench reactor method

    DOEpatents

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

    1999-08-10

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

  8. Fast Reactor Technology Preservation

    SciTech Connect

    Wootan, David W.; Omberg, Ronald P.

    2008-01-11

    There is renewed worldwide interest in developing and implementing a new generation of advanced fast reactors. International cooperative efforts are underway such as the Global Nuclear Energy Partnership (GNEP). Advanced computer modeling and simulation efforts are a key part of these programs. A recognized and validated set of Benchmark Cases are an essential component of such modeling efforts. Testing documentation developed during the operation of the Fast Flux Test Facility (FFTF) provide the information necessary to develop a very useful set of Benchmark Cases.

  9. Spatial kinetics in fast reactors

    NASA Astrophysics Data System (ADS)

    Seleznev, E. F.; Belov, A. A.; Panova, I. S.; Matvienko, I. P.; Zhukov, A. M.

    2013-12-01

    The analysis of the solution to the spatial nonstationary equation of neutron transport is presented by the example of a fast reactor. Experiments in spatial kinetics conducted recently at the complex of critical assemblies (fast physical stand) and computations of their data using the TIMER code (for solving the nonstationary equation in multidimensional diffusion approximation for direct and inverse problems of reactor kinetics) have shown that kinetics of fast reactors substantially differs from kinetics of thermal reactors. The difference is connected with influence of the delayed neutron spectrum on rates of the process in a fast reactor.

  10. History of fast reactor fuel development

    NASA Astrophysics Data System (ADS)

    Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

    1993-09-01

    The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

  11. FAST NEUTRONIC REACTOR

    DOEpatents

    Snell, A.H.

    1957-12-01

    This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

  12. Fast reactors and nuclear nonproliferation

    SciTech Connect

    Avrorin, E.N.; Rachkov, V.I.; Chebeskov, A.N.

    2013-07-01

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  13. Fast quench reactor and method

    DOEpatents

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

    1998-05-12

    A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

  14. Fuel systems for compact fast space reactors

    SciTech Connect

    Cox, C.M.; Dutt, D.S.; Karnesky, R.A.

    1983-12-01

    About 200 refractory metal clad ceramic fuel pins have been irradiated in thermal reactors under the 1200 K to 1550 K cladding temperature conditions of primary relevance to space reactors. This paper reviews performance with respect to fissile atom density, operating temperatures, fuel swelling, fission gas release, fuel-cladding compatibility, and consequences of failure. It was concluded that UO/sub 2/ and UN fuels show approximately equal performance potential and that UC fuel has lesser potential. W/Re alloys have performed quite well as cladding materials, and Ta, Nb, and Mo/Re alloys, in conjunction with W diffusion barriers, show good promise. Significant issues to be addressed in the future include high burnup swelling of UN, effects of UO/sub 2/-Li coolant reaction in the event of fuel pin failure, and development of an irradiation performance data base with prototypically configured fuel pins irradiated in a fast neutron flux.

  15. Fast quench reactor and method

    SciTech Connect

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    2002-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

  16. Fast quench reactor and method

    SciTech Connect

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    1998-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

  17. Fast quench reactor and method

    SciTech Connect

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    2002-09-24

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

  18. Heterogeneous Transmutation Sodium Fast Reactor

    SciTech Connect

    S. E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even neutron number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both non-flattened and flattened core geometries. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of comparable size. A mass balance analysis revealed that the heterogeneous design may reduce the number of fast reactors needed to close the current once-through light water reactor fuel cycle.

  19. Exploratory evaluation of ceramics for automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1972-01-01

    An exploratory evaluation of ceramics for automobile thermal reactors was conducted. Potential ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance lasting over 800 hours in engine dynamometer tests and over 15,000 miles (24,200 km) of vehicle road tests. Reactors containing glass-ceramic components did not perform as well as silicon carbide. But the glass-ceramics still offer good potential for reactor use. The results of this study are considered to be a reasonable demonstration of the potential use of ceramics in thermal reactors.

  20. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  1. Fuel development for gas-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  2. On fast reactor kinetics studies

    SciTech Connect

    Seleznev, E. F.; Belov, A. A.; Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F.

    2012-07-01

    The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)

  3. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    SciTech Connect

    STAN, MARIUS; HECKER, SIEGFRIED S.

    2007-02-07

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  4. Design and evaluation of experimental ceramic automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1974-01-01

    The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

  5. Design and evaluation of experimental ceramic automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1974-01-01

    The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

  6. Heterogeneous Recycling in Fast Reactors

    SciTech Connect

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  7. Fast breeder reactor protection system

    DOEpatents

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  8. Advances by the Integral Fast Reactor Program

    SciTech Connect

    Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

    1991-01-01

    The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs.

  9. Fast pulse nonthermal plasma reactor

    DOEpatents

    Rosocha, Louis A.

    2005-06-14

    A fast pulsed nonthermal plasma reactor includes a discharge cell and a charging assembly electrically connected thereto. The charging assembly provides plural high voltage pulses to the discharge cell. Each pulse has a rise time between one and ten nanoseconds and a duration of three to twenty nanoseconds. The pulses create nonthermal plasma discharge within the discharge cell. Accordingly, the nonthermal plasma discharge can be used to remove pollutants from gases or break the gases into smaller molecules so that they can be more efficiently combusted.

  10. Fast Measurements Of Thermal Diffusivities Of Ceramics

    NASA Technical Reports Server (NTRS)

    Smith, Marnell; Goldstein, Howard E.

    1988-01-01

    Temperature rises of samples compared with reference sample. Apparatus quickly measures thermal diffusivities of ceramics at high temperatures. Produces data on relative thermal diffusivities of as many as six ceramic specimens per hour. Thermal-diffusivity tester makes it easy to determine thermal diffusivities of ceramics. Pronounced effects of processing parameters on thermal properties of ceramics evaluated quickly.

  11. Research Program of a Super Fast Reactor

    SciTech Connect

    Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki; Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki; GOTO, Shoji

    2006-07-01

    Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

  12. Advanced Safeguards Approaches for New Fast Reactors

    SciTech Connect

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  13. Fission energy: The integral fast reactor

    SciTech Connect

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements.

  14. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  15. COUPLED FAST-THERMAL POWER BREEDER REACTOR

    DOEpatents

    Avery, R.

    1961-07-18

    A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

  16. Advanced ceramic cladding for water reactor fuel

    SciTech Connect

    Feinroth, H.

    2000-07-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of {approximately}60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies {ge}50% would be examined.

  17. Fast-acting nuclear reactor control device

    DOEpatents

    Kotlyar, Oleg M.; West, Phillip B.

    1993-01-01

    A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

  18. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    NASA Astrophysics Data System (ADS)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  19. Stationary Liquid Fuel Fast Reactor

    SciTech Connect

    Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  20. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    SciTech Connect

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-09-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850ºC at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05).

  1. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  2. Integral data for fast reactors

    SciTech Connect

    Collins, P.J.; Poenitz, W.P.; McFarlane, H.F.

    1988-01-01

    Requirements at Argonne National Laboratory to establish the best estimates and uncertainties for LMR design parameters have lead to an extensive evaluation of the available critical experiment database. Emphasis has been put upon selection of a wide range of cores, including both benchmark, assemblies covering a range of spectra and compositions and power reactor mock-up assemblies with diverse measured parameters. The integral measurements have been revised, where necessary, using the most recent reference data and a covariance matrix constructed. A sensitivity database has been calculated, embracing all parameters, which enables quantification of the relevance of the integral data to parameters calculated with ENDF/B-V.2 cross sections.

  3. MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

    DOEpatents

    Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

    1962-06-26

    A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

  4. Nonlinear, inelastic fast reactor subassembly interaction analyses

    SciTech Connect

    Sutherland, W.H.; Bard, F.E.

    1983-01-01

    Liquid Metal Fast Breeder Reactor (LMFBR) core structural design is complicated by the trade-offs associated with keeping the subassemblies closely packed for the neutronic considerations and accommodating the volumetric changes associated with irradiation swelling. The environmental variation across the reactor core results in temperature and neutron flux gradients across the subassemblies which in turn cause the subassemblies to bow as well as dilate and grow volumetrically. These deformations in a tightly packed reactor core cause the subassemblies to interact and can potentially result in excessive withdrawal loads during the refueling operations. ABADAN, a general purpose, nonlinear, inelastic, multi-dimensional finite element structural analysis computer code, was developed for the express purpose of solving large nonlinear problems as typified by the above interaction problems. For the subassembly interaction problem ABADAN has been applied to the solution of an interacting radial row of Fast Flux Test Facility (FFTF) fuel assemblies.

  5. 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.

  6. Behavior of actinides in the Integral Fast Reactor fuel cycle

    SciTech Connect

    Courtney, J.C.; Lineberry, M.J.

    1994-06-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  7. Gas-Fast Reactor Fuel Fabrication

    SciTech Connect

    Randall Fielding; Mitchell Meyer; Ramprashad Prabhakaran; Jim Miller; Sean McDeavitt

    2005-11-01

    The gas-cooled fast reactor is a high temperature helium cooled Generation IV reactor concept. Operating parameters for this type of reactor are well beyond those of current fuels so a novel fuel must be developed. One fuel concept calls for UC particles dispersed throughout a SiC matrix. This study examines a hybrid reaction bonding process as a possible fabrication route for this fuel. Processing parameters are also optimized. The process combines carbon and SiC powders and a carbon yielding polymer. In order to obtain dense reaction bonded SiC samples the porosity to carbon ratio in the preform must be large enough to accommodate SiC formation from the carbon present in the sample, however too much porosity reduces mechanical integrity which leads to poor infiltration properties . The porosity must also be of a suitable size to allow silicon transport throughout the sample but keep residual silicon to a minimum.

  8. Risk Management for Sodium Fast Reactors.

    SciTech Connect

    Denman, Matthew R; Groth, Katrina; Cardoni, Jeffrey N; Wheeler, Timothy A.

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

  9. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    SciTech Connect

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  10. Reactor vessel using metal oxide ceramic membranes

    DOEpatents

    Anderson, Marc A.; Zeltner, Walter A.

    1992-08-11

    A reaction vessel for use in photoelectrochemical reactions includes as its reactive surface a metal oxide porous ceramic membrane of a catalytic metal such as titanium. The reaction vessel includes a light source and a counter electrode. A provision for applying an electrical bias between the membrane and the counter electrode permits the Fermi levels of potential reaction to be favored so that certain reactions may be favored in the vessel. The electrical biasing is also useful for the cleaning of the catalytic membrane.

  11. Fast critical experiment data for space reactors

    NASA Astrophysics Data System (ADS)

    Collins, P. J.; McFarlane, H. F.; Olsen, D. N.; Atkinson, C. A.; Ross, J. R.

    Data from a number of previous critical experiments exist that are relevant to the design concepts being considered for SP-100 and MMW space reactors. Although substantial improvements in experiment techniques have since made some of the measured quantities somewhat suspect, the basic criticality data are still useful in most cases. However, experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, about 20 fast benchmark critical experiments were calculated with the latest ENDF/B data and modern transport codes. These calculations were undertaken as a part of the planning process for a new series of benchmark experiments aimed at supporting preliminary designs of SP-100 and MMW space reactors.

  12. Actinide management with commercial fast reactors

    NASA Astrophysics Data System (ADS)

    Ohki, Shigeo

    2015-12-01

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GWey if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

  13. Actinide management with commercial fast reactors

    SciTech Connect

    Ohki, Shigeo

    2015-12-31

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GW{sub e}y if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

  14. Accident analysis for US fast burst reactors

    SciTech Connect

    Paternoster, R.; Flanders, M.; Kazi, H.

    1994-09-01

    In the US fast burst reactor (FBR) community there has been increasing emphasis and scrutiny on safety analysis and understanding of possible accident scenarios. This paper summarizes recent work in these areas that is going on at the different US FBR sites. At this time, all of the FBR facilities have or in the process of updating and refining their accident analyses. This effort is driven by two objectives: to obtain a more realistic scenario for emergency response procedures and contingency plans, and to determine compliance with changing regulatory standards.

  15. High temperature ceramic membrane reactors for coal liquid upgrading

    SciTech Connect

    Tsotsis, T.T.

    1992-06-19

    In this project we well evaluate the performance of Sel-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated. (VC)

  16. High temperature ceramic membrane reactors for coal liquid upgrading

    SciTech Connect

    Tsotsis, T.T.

    1992-01-01

    In this project we intend to study a novel process concept, i.e, the use of ceramic membranes reactors in upgrading of coal derived liquids. Membrane reactors have been used in a number of catalytic reaction processes in order to overcome the limitations on conversion imposed by thermodynamic equilibrium. They have, furthermore, the inherent capability for combining reaction and separation in a single step. Thus they offer promise for improving and optimizing yield, selectivity and performance of processes involving complex liquids, as those typically found in coal liquid upgrading. Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we will evaluate the performance of Sol-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

  17. High temperature ceramic membrane reactors for coal liquid upgrading

    SciTech Connect

    Tsotsis, T.T.

    1992-01-01

    In this project we intend to study a novel process concept, i.e.,the use of ceramic membranes reactors in upgrading of coal derived liquids. Membrane reactors have been used in a number of catalytic reaction processes in order to overcome the limitations on conversion imposed by thermodynamic equilibrium. They have, furthermore, the inherent capability for combining reaction and separation in a single step. Thus they offer promise for improving and optimizing yield, selectivity and performance of processes involving complex liquids, as those typically found in coal liquid upgrading. Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we wig evaluate the performance of Sel-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

  18. A resting bottom sodium cooled fast reactor

    SciTech Connect

    Costes, D.

    2012-07-01

    This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

  19. LFR "Lead-Cooled Fast Reactor"

    SciTech Connect

    Cinotti, L; Fazio, C; Knebel, J; Monti, S; Abderrahim, H A; Smith, C; Suh, K

    2006-05-11

    The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small, transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology

  20. A CANDU-Based Fast Irradiation Reactor

    SciTech Connect

    Shatilla, Youssef

    2006-07-01

    A new steady-state fast neutron reactor is needed to satisfy the testing needs of Generation IV reactors, the Space Propulsion Program, and the Advanced Fuel Cycle Initiative. This paper presents a new concept for a CANDU-based fast irradiation reactor that is horizontal in orientation, with individual pressure tubes running the entire length of the scattering-medium tank (Calandria) filled with Lead-Bismuth-Eutectic (LBE). This approach for a test reactor will provide more flexibility in refueling, sample removal, and ability to completely re-configure the core to meet different users' requirements. Full core neutronic analysis of several fuel/coolant/geometry combinations showed a small hexagonal, LBE-cooled, U-Pu-10Zr fuel, with a core power of 100 MW{sub th} produced a fast flux (>0.1 MeV) of 1.5 x 10{sup 15} n/cm{sup 2} sec averaged over the whole length of six irradiation channels with a total testing volume of more than 77 liters. In-core breeding allowed the Pu-239 enrichment to be 15.3% which should result in core continuous operation for 180 effective full power days. Other coolants investigated included high pressure water steam and helium. An innovative shutdown/control system which consisted of the six outermost fuel channels was proven to be effective in shutting the core down when flooded with boric acid as a neutron absorber. The new shutdown/control system has the advantage of causing the minimum perturbation of the axial flux shape when the control channels are partially flooded with boric acid. This is because the acid is injected homogeneously along the control channel in contrast to regular control rods that are injected partially causing an axial perturbation in the core flux which in turn reduces safety analysis margins. The new shutdown/control system is not required to penetrate the core in a direction vertical to the fuel channels which allowed the freedom of changing core pitch as deemed necessary. A preliminary thermal hydraulic analysis

  1. The Industrial Sodium Cooled Fast Reactor

    SciTech Connect

    Samuel E. Bays; Haihua Zhao; Hongbin Zhang

    2009-04-01

    This paper investigates the use of enrichment and moderator zoning methods for optimizing the r-z power distribution within sodium cooled fast reactors. These methods allow overall greater fuel utilization in the core resulting in more fuel being irradiated near the maximum allowed thermal power. The peak-to-average power density was held to 1.18. This core design, in conjunction with a multiple-reheat Brayton power conversion system, has merit for producing an industrial level of electrical output (2400MWth, 1000MWe) from a relatively compact core size. The total core radius, including reflectors and shields, was held to 1.78m. Preliminary safety analysis suggests that positive reactivity insertion resulting from a leak between the sodium primary loop and helium power conversion system can be mitigated using simple gas-liquid centripetal separation strategies in the plant’s primary loop.

  2. High temperature ceramic membrane reactors for coal liquid upgrading

    SciTech Connect

    Tsotsis, T.T.

    1992-06-19

    Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we will evaluate the performance of Sel-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will also be investigated.

  3. High temperature ceramic membrane reactors for coal liquid upgrading

    SciTech Connect

    Tsotsis, T.T.

    1992-06-19

    Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we will evaluate the performance of Sel-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

  4. High temperature ceramic membrane reactors for coal liquid upgrading

    SciTech Connect

    Tsotsis, T.T.

    1992-01-01

    In this project we will study a novel process concept, i.e., the use of ceramic membrane reactors in upgrading of coal model compounds and coal derived liquids. In general terms, the USC research team is responsible for constructing and operating the membrane reactor apparatus and for testing various inorganic membranes for the upgrading of coal derived asphaltenes and coal model compounds. The USC effort will involve the principal investigator of this project and two graduate research assistants. The ALCOA team is responsible for the preparation of the inorganic membranes, for construction and testing of the ceramic membrane modules, and for measurement of their transport properties. The ALCOA research effort will involve Dr. Paul K. T. Liu, who is the project manager of the ALCOA research team, an engineer and a technician. UNOCAL's contribution will be limited to overall technical assistance in catalyst preparation and the operation of the laboratory upgrading membrane reactor and for analytical back-up and expertise in oil analysis and materials characterization. UNOCAL is a no-cost contractor but will be involved in all aspects of the project, as deemed appropriate.

  5. Implications of Fast Reactor Transuranic Conversion Ratio

    SciTech Connect

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

    2010-11-01

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

  6. BN-800 advanced nuclear power plant with fast reactor

    SciTech Connect

    Shishkin, A.N.; Kuzavkov, N.G.; Sobolev, V.A.; Shestakov, G.V.; Bagdasarov, Yu.E.; Kochetkov, L.A.; Matveyev, V.I.; Poplavsky, V.M.

    1993-12-31

    Bn-800 reactor plant with fast reactor and sodium coolant in the primary and secondary circuits is designed for operation as part of the power units in the Yuzhno-Uralskaya nuclear power plant scheduled to be constructed in Chelyabinsk region and as part unit 4 in the Beloyarskaya nuclear power plant. Reactor operations are described.

  7. 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.

  8. 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.

  9. Cermet fuel for fast reactor - Fabrication and characterization

    NASA Astrophysics Data System (ADS)

    Mishra, Sudhir; Kutty, P. S.; Kutty, T. R. G.; Das, Shantanu; Dey, G. K.; Kumar, Arun

    2013-11-01

    (U, Pu)O2 ceramic fuel is the well-established fuel for the fast reactors and (U, Pu, Zr) metallic fuel is the future fuel. Both the fuels have their own merits and demerits. Optimal solution may lie in opting for a fuel which combines the favorable features of both fuel systems. The choice may be the use of cermet fuel which can be either (U, PuO2) or (Enriched U, UO2). In the present study, attempt has been made to fabricate (Natural U, UO2) cermet fuel by powder metallurgy route. Characterization of the fuel has been carried out using dilatometer, differential thermal analyzer, X-ray diffractometer, and Scanning Electron Microscope. The results show a high solidus temperature, high thermal expansion, presence of porosities, etc. in the fuel. The thermal conductivity of the fuel has also been measured. X-ray diffraction study on the fuel compact reveals presence of α U and UO2 phases in the matrix of the fuel.

  10. Fast biomass pyrolysis with an entrained-flow reactor

    NASA Astrophysics Data System (ADS)

    Bohn, M. S.; Benham, C.

    1982-02-01

    A tubular entrained flow reactor has been used to study the effect of process control variables on fast biomass pyrolysis. In this type of reactor, finely ground biomass particles are entrained by carrier gas and transported through a reactor tube which is heated to about 900 C. Biomass particles pyrolyze as a result of heat transfer from the reactor wall yielding a gas composed primarily of carbon monoxide, carbon dioxide, hydrogen, methane, and unsaturated hydrocarbons. In this experimental program three dependent variables, percent conversion to gas, gas composition, and process heat, have been measured as a function of several process control variables. These process variables include reactor temperature, carrier gas to biomass flow ratio, reactor residence time, biomass particle size, and reactor Reynolds number. The data allow one to design and predict the performance of large scale reactors and also elucidates heat transfer mechanisms in fast biomass pyrolysis.

  11. A fast spectrum dual path flow cermet reactor

    SciTech Connect

    Anghaie, S.; Feller, G.J. ); Peery, S.D.; Parsley, R.C. )

    1993-01-15

    A cermet fueled, dual path fast reactor for space nuclear propulsion applications is conceptually designed. The reactor utilizes an outer annulus core and an inner cylindrical core with radial and axial reflector. The dual path flow minimizes the impact of power peaking near the radial reflector. Basic neutronics and core design aspects of the reactor are discussed. The dual path reactor is integrated into a 25000 lbf thrust nuclear rocket.

  12. Sodium fast reactor safety and licensing research plan. Volume II.

    SciTech Connect

    Ludewig, H.; Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.; Lambert, J.; Hayes, S.; Sackett, J.; Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  13. Immobilization of Fast Reactor First Cycle Raffinate

    SciTech Connect

    Langley, K. F.; Partridge, B. A.; Wise, M.

    2003-02-26

    This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.

  14. High temperature ceramic membrane reactors for coal liquid upgrading

    SciTech Connect

    Tsotsis, T.T.

    1992-01-01

    Membrane reactors have been used in a number of catalytic reaction processes in order to overcome the limitations on conversion imposed by thermodynamic equilibrium. Having the inherent capability for combining reaction and separation in a single step, they offer promise for improving and optimizing yield, selectivity and performance of processes involving complex liquids, such as these typically found in coal liquid upgrading. Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. This project will evaluate the performance of Sol-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. Development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

  15. Technology Options for a Fast Spectrum Test Reactor

    SciTech Connect

    D. M. Wachs; R. W. King; I. Y. Glagolenko; Y. Shatilla

    2006-06-01

    Idaho National Laboratory in collaboration with Argonne National Laboratory has evaluated technology options for a new fast spectrum reactor to meet the fast-spectrum irradiation requirements for the USDOE Generation IV (Gen IV) and Advanced Fuel Cycle Initiative (AFCI) programs. The US currently has no capability for irradiation testing of large volumes of fuels or materials in a fast-spectrum reactor required to support the development of Gen IV fast reactor systems or to demonstrate actinide burning, a key element of the AFCI program. The technologies evaluated and the process used to select options for a fast irradiation test reactor (FITR) for further evaluation to support these programmatic objectives are outlined in this paper.

  16. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    SciTech Connect

    Seifritz, W.

    1983-11-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase.

  17. 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.

  18. Methods for quantifying uncertainty in fast reactor analyses.

    SciTech Connect

    Fanning, T. H.; Fischer, P. F.

    2008-04-07

    Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.

  19. Calculated analysis of experiments in fast neutron reactors

    SciTech Connect

    Davydov, V. K. Kalugina, K. M.; Gomin, E. A.

    2012-12-15

    In this paper, the results of computational simulation of experiments with the MK-I core of the JOYO fast neutron sodium-cooled reactor are presented. The MCU-KS code based on the Monte Carlo method was used for calculations. The research was aimed at additional verification of the MCU-KS code for systems with a fast neutron spectrum.

  20. Calculated analysis of experiments in fast neutron reactors

    NASA Astrophysics Data System (ADS)

    Davydov, V. K.; Kalugina, K. M.; Gomin, E. A.

    2012-12-01

    In this paper, the results of computational simulation of experiments with the MK-I core of the JOYO fast neutron sodium-cooled reactor are presented. The MCU-KS code based on the Monte Carlo method was used for calculations. The research was aimed at additional verification of the MCU-KS code for systems with a fast neutron spectrum.

  1. Composite nuclear fuel fabrication methodology for gas fast reactors

    NASA Astrophysics Data System (ADS)

    Vasudevamurthy, Gokul

    An advanced fuel form for use in Gas Fast Reactors (GFR) was investigated. Criteria for the fuel includes operation at high temperature (˜1400°C) and high burnup (˜150 MWD/MTHM) with effective retention of fission products even during transient temperatures exceeding 1600°C. The GFR fuel is expected to contain up to 20% transuranics for a closed fuel cycle. Earlier evaluations of reference fuels for the GFR have included ceramic-ceramic (cercer) dispersion type composite fuels of mixed carbide or nitride microspheres coated with SiC in a SiC matrix. Studies have indicated that ZrC is a potential replacement for SiC on account of its higher melting point, increased fission product corrosion resistance and better chemical stability. The present work investigated natural uranium carbide microspheres in a ZrC matrix instead of SiC. Known issues of minor actinide volatility during traditional fabrication procedures necessitated the investigation of still high temperature but more rapid fabrication techniques to minimize these anticipated losses. In this regard, fabrication of ZrC matrix by combustion synthesis from zirconium and graphite powders was studied. Criteria were established to obtain sufficient matrix density with UC microsphere volume fractions up to 30%. Tests involving production of microspheres by spark erosion method (similar to electrodischarge machining) showed the inability of the method to produce UC microspheres in the desired range of 300 to 1200 mum. A rotating electrode device was developed using a minimum current of 80A and rotating at speeds up to 1500 rpm to fabricate microspheres between 355 and 1200 mum. Using the ZrC process knowledge, UC electrodes were fabricated and studied for use in the rotating electrode device to produce UC microspheres. Fabrication of the cercer composite form was studied using microsphere volume fractions of 10%, 20%, and 30%. The macrostructure of the composite and individual components at various stages were

  2. Simulator platform for fast reactor operation and safety technology demonstration

    SciTech Connect

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  3. 70 Group Neutron Fast Reactor Cross Section Set and 25 Group Neutron Fast Reactor Cross Section Set.

    Energy Science and Technology Software Center (ESTSC)

    1984-10-29

    Version 00 These multigroup cross sections are used in fast reactor calculations. The benchmark calculations for the 23 fast critical assemblies used in the benchmark tests of JFS-2 were performed with one-dimensional diffusion theory by using the JFS-3-J2 set.

  4. CERAMIC FUEL ELEMENT MATERIAL FOR A NEUTRONIC REACTOR AND METHOD OF FABRICATING SAME

    DOEpatents

    Duckworth, W.H.

    1957-12-01

    This patent relates to ceramic composition, and to neutronic reactor fuel elements formed therefrom. These ceramic elements have high density and excellent strength characteristics and are formed by conventional ceramic casting and sintering at a temperature of about 2700 deg F in a nitrogen atmosphere. The composition consists of silicon carbide, silicon, uranium oxide and a very small percentage of molybdenum. Compositions containing molybdenum are markedly stronger than those lacking this ingredient.

  5. High temperature ceramic membrane reactors for coal liquid upgrading

    SciTech Connect

    Tsotsis, T.T. . Dept. of Chemical Engineering); Liu, P.K.T. ); Webster, I.A. )

    1992-01-01

    Membrane reactors are today finding extensive applications for gas and vapor phase catalytic reactions (see discussion in the introduction and recent reviews by Armor [92], Hsieh [93] and Tsotsis et al. [941]). There have not been any published reports, however, of their use in high pressure and temperature liquid-phase applications. The idea to apply membrane reactor technology to coal liquid upgrading has resulted from a series of experimental investigations by our group of petroleum and coal asphaltene transport through model membranes. Coal liquids contain polycyclic aromatic compounds, which not only present potential difficulties in upgrading, storage and coprocessing, but are also bioactive. Direct coal liquefaction is perceived today as a two-stage process, which involves a first stage of thermal (or catalytic) dissolution of coal, followed by a second stage, in which the resulting products of the first stage are catalytically upgraded. Even in the presence of hydrogen, the oil products of the second stage are thought to equilibrate with the heavier (asphaltenic and preasphaltenic) components found in the feedstream. The possibility exists for this smaller molecular fraction to recondense with the unreacted heavy components and form even heavier undesirable components like char and coke. One way to diminish these regressive reactions is to selectively remove these smaller molecular weight fractions once they are formed and prior to recondensation. This can, at least in principle, be accomplished through the use of high temperature membrane reactors, using ceramic membranes which are permselective for the desired products of the coal liquid upgrading process. An additional incentive to do so is in order to eliminate the further hydrogenation and hydrocracking of liquid products to undesirable light gases.

  6. Interim Status Report on the Design of the Gas-Cooled Fast Reactor (GFR)

    SciTech Connect

    Weaver, K. D.

    2005-01-31

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above.

  7. Interim Status Report on the Design of the Gas-Cooled Fast Reactor (GFR)

    SciTech Connect

    Kevan D. Weaver

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850ºC at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above.

  8. Development of a novel ceramic microchannel reactor for methane steam reforming

    NASA Astrophysics Data System (ADS)

    Murphy, Danielle M.

    Microchannel heat exchanger and reactor technology has recently gained interest as an innovative way to improve heat-exchanger efficiency, reduce size and weight, and utilize thermal management capabilities to improve conversion, yield, selectivity, and catalyst life. Among many other possible applications, this technology is suitable for advanced recuperated engines, oxy-fired combustion processes for oxygen separation, gas-cooled nuclear reactors, recuperative heat exchanger and reformer units for solid oxide fuel cell systems, and chemical processing. This work presents the design, fabrication, and performance of novel ceramic microchannel reactors in heat-exchanger and fuel-reforming applications. Although most microchannel devices are made of metal materials, ceramics offer an alternative which enables significantly higher operating temperatures, improved tolerance to harsh chemical environments, and improved adherence of ceramic-based catalyst washcoats. Significant cost savings in materials and manufacturing methods for high-volume manufacturing can also be achieved. High-temperature performance of the ceramic microchannel reactor is measured through non-reactive heat-exchanger experiments within a dedicated test stand. Heat-exchanger effectiveness of up to 88% is experimentally established. After coating catalyst material over half of the reactor layers, use of the ceramic microchannel reactor in methane fuel-processing applications is demonstrated. As a fuel reformer, the ceramic microchannel reactor achieves process intensification by combining heat-exchanger and catalytic-reactor functions to produce syngas. Gas hourly space velocities (GHSV) up to 50,000 hr-1 with methane conversion higher than 85% are achieved. A complete computational fluid dynamics (CFD) model, as well as a geometrically simplified hybrid CFD/chemical kinetics model, is used in conjunction with experimentation to examine heat transfer, fluid flow, and chemical kinetics within the

  9. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    SciTech Connect

    Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel; Zhang, Yongfeng; Novascone, Stephen Rhead; Medvedev, Pavel G.

    2015-09-01

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

  10. A fast shutdown system for SRS (Savannah River Site) reactors

    SciTech Connect

    Baumann, N.P.

    1990-01-01

    Power has been sharply reduced at Savannah River Site (SRS) reactors in large part to ensure that no bulk boiling occurs during hypothesized loss of coolant accidents. A fast shutdown system is essential to regain much of this lost power. Computations and experiments indicate that a He-3 injection system will serve this function. Instrumented tests of a full system are planned for early 1991 for one of the SRS reactors. 4 refs., 7 figs., 1 tab.

  11. Spectrophotometric Procedure for Fast Reactor Advanced Coolant Manufacture Control

    NASA Astrophysics Data System (ADS)

    Andrienko, O. S.; Egorov, N. B.; Zherin, I. I.; Indyk, D. V.

    2016-01-01

    The paper describes a spectrophotometric procedure for fast reactor advanced coolant manufacture control. The molar absorption coefficient of dimethyllead dibromide with dithizone was defined as equal to 68864 ± 795 l·mole-1·cm-1, limit of detection as equal to 0.583 · 10-6 g/ml. The spectrophotometric procedure application range was found to be equal to 37.88 - 196.3 g. of dimethyllead dibromide in the sample. The procedure was used within the framework of the development of the method of synthesis of the advanced coolant for fast reactors.

  12. Instrumentation, Monitoring and NDE for New Fast Reactors

    SciTech Connect

    Bond, Leonard J.; Doctor, Steven; Bunch, Kyle; Good, Morris; Waltar, Alan E.

    2007-07-01

    The Global Nuclear Energy Partnership (GNEP) will require the development of actinide transmutation, which can most effectively be accomplished in a fast-spectrum reactor. To achieve higher standards of safety and reliability, operate with longer intervals between outages, and achieve high operating capacity factors, new instrumentation and on-line monitoring capabilities will be required-- during both fabrication and operation. This paper reports parts of a knowledge capture and technology state-of-the-art assessment for fast-reactor instrumentation and controls, monitoring and diagnostics. (authors)

  13. Impact of Including Higher Actinides in Fast Reactor Transmutation Analyses

    SciTech Connect

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

    2007-09-01

    Previous fast reactor transmutation studies generally disregarded higher mass minor actinides beyond Cm-246 due to various considerations including deficiencies in nuclear cross-section data. Although omission of these higher mass actinides does not significantly impact the neutronic calculations and fuel cycle performance parameters follow-on neutron dose calculations related to fuel recycling, transportation and handling are significantly impacted. This report shows that including the minor actinides in the equilibrium fast reactor calculations will increase the predicted neutron emission by about 30%. In addition a sensitivity study was initiated by comparing the impact of different cross-section evaluation file for representing these minor actinides.

  14. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    SciTech Connect

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  15. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  16. Metallic Fast Reactor Fuel Fabrication for Global Nuclear Energy Partnership

    SciTech Connect

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter

    2009-07-01

    Fast reactors are once again being considered for nuclear power generation, in addition to transmutation of long-lived fission products resident in spent nuclear fuels. This re-consideration follows with intense developmental programs for both fuel and reactor design. One of the two leading candidates for next generation fast reactor fuel is metal alloys, resulting primarily from the successes achieved in the 1960s to early 1990s with both the experimental breeding reactor-II and the fast flux test facility. The goal of the current program is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional, fast-spectrum nuclear fuel while destroying recycled actinides, thereby closing the nuclear fuel cycle. In order to meet this goal, the program must develop efficient and safe fuel fabrication processes designed for remote operation. This paper provides an overview of advanced casting processes investigated in the past, and the development of a gaseous diffusion calculation that demonstrates how straightforward process parameter modification can mitigate the loss of volatile minor actinides in the metal alloy melt.

  17. Integral Fast Reactor Program annual progress report, FY 1991

    SciTech Connect

    Not Available

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

  18. Integral Fast Reactor Program annual progress report, FY 1991

    SciTech Connect

    Not Available

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R D.

  19. Packed rod neutron shield for fast nuclear reactors

    DOEpatents

    Eck, John E.; Kasberg, Alvin H.

    1978-01-01

    A fast neutron nuclear reactor including a core and a plurality of vertically oriented neutron shield assemblies surrounding the core. Each assembly includes closely packed cylindrical rods within a polygonal metallic duct. The shield assemblies are less susceptible to thermal stresses and are less massive than solid shield assemblies, and are cooled by liquid coolant flow through interstices among the rods and duct.

  20. Detailed calculations of minor actinide transmutation in a fast reactor

    SciTech Connect

    Takeda, Toshikazu

    2015-12-31

    The transmutation of minor actinides in a fast reactor is investigated by a new method to investigate the transmutation behavior of individual minor actinides. It is found that Np-237 and Am-241 mainly contributes to the transmutation rate though the transmutation behaviors are very different.

  1. Integral Fast Reactor Program annual progress report, FY 1994

    SciTech Connect

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, J.J.

    1994-12-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1994. Technical accomplishments are presented in the following areas of the IFR technology development activities: metal fuel performance; pyroprocess development; safety experiments and analyses; core design development; fuel cycle demonstration; and LMR technology R&D.

  2. Detailed calculations of minor actinide transmutation in a fast reactor

    NASA Astrophysics Data System (ADS)

    Takeda, Toshikazu

    2015-12-01

    The transmutation of minor actinides in a fast reactor is investigated by a new method to investigate the transmutation behavior of individual minor actinides. It is found that Np-237 and Am-241 mainly contributes to the transmutation rate though the transmutation behaviors are very different.

  3. Integral Fast Reactor Program. Annual progress report, FY 1992

    SciTech Connect

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1993-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

  4. Integral Fast Reactor Program. Annual progress report, FY 1993

    SciTech Connect

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D.

  5. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    SciTech Connect

    R. Wigeland; K. Hamman

    2009-09-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving

  6. Physics Characterization of a Heterogeneous Sodium Fast Reactor Transmutation System

    SciTech Connect

    Samuel E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even mass number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both a non-flattened and a pancake core geometry. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of the same size.

  7. Neutron initiation probability in fast burst reactor

    SciTech Connect

    Liu, X.; Du, J.; Xie, Q.; Fan, X.

    2012-07-01

    Based on the probability balance of neutron random events in multiply system, the four random process of neutron in prompt super-critical is described and then the equation of neutron initiation probability W(r,E,{Omega},t) is deduced. On the assumption of static, slightly prompt super-critical and the two factorial approximation, the formula of the average probability of 'one' neutron is derived which is the same with the result derived from the point model. The MC simulation using point model is applied in Godiva- II and CFBR-II, and the simulation result of one neutron initiation is well consistent with the theory that the initiation probability of Godiva- II inverted commas CFBR-II burst reactor are 0.00032, 0.00027 respectively on the ordinary burst operation. (authors)

  8. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  9. Exploratory development of a glass ceramic automobile thermal reactor. [anti-pollution devices

    NASA Technical Reports Server (NTRS)

    Gould, R. E.; Petticrew, R. W.

    1973-01-01

    This report summarizes the design, fabrication and test results obtained for glass-ceramic (CER-VIT) automotive thermal reactors. Several reactor designs were evaluated using both engine-dynamometer and vehicle road tests. A maximum reactor life of about 330 hours was achieved in engine-dynamometer tests with peak gas temperatures of about 1065 C (1950 F). Reactor failures were mechanically induced. No evidence of chemical degradation was observed. It was concluded that to be useful for longer times, the CER-VIT parts would require a mounting system that was an improvement over those tested in this program. A reactor employing such a system was designed and fabricated.

  10. Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs.

  11. Fast Breeder Reactors in Sweden: Vision and Reality.

    PubMed

    Fjaestad, Maja

    2015-01-01

    The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and '60s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy. Sweden had advanced plans for a nuclear breeder program, but canceled them in the middle of the 1970s with the rise of nuclear skepticism. The article investigates the nuclear breeder as a technological vision. The nuclear breeder reactor is an example of a technological future that did not meet its industrial expectations. But that does not change the fact that the breeder was an influential technology. Decisions about the contemporary reactors were taken with the idea that in a foreseeable future they would be replaced with the efficient breeder. The article argues that general themes in the history of the breeder reactor can deepen our understanding of the mechanisms behind technological change. PMID:26334698

  12. The IAEA international conference on fast reactors and related fuel cycles: highlights and main outcomes

    SciTech Connect

    Monti, S.; Toti, A.

    2013-07-01

    The 'International Conference on Fast Reactors and Related Fuel Cycles', which is regularly held every four years, represents the main international event dealing with fast reactors technology and related fuel cycles options. Main topics of the conference were new fast reactor concepts, design and simulation capabilities, safety of fast reactors, fast reactor fuels and innovative fuel cycles, analysis of past experience, fast reactor knowledge management. Particular emphasis was put on safety aspects, considering the current need of developing and harmonizing safety standards for fast reactors at the international level, taking also into account the lessons learned from the accident occurred at the Fukushima- Daiichi nuclear power plant in March 2011. Main advances in the several key areas of technological development were presented through 208 oral presentations during 41 technical sessions which shows the importance taken by fast reactors in the future of nuclear energy.

  13. Instrumentation, Monitoring and NDE for New Fast Reactors

    SciTech Connect

    Bond, Leonard J.; Doctor, Steven R.; Bunch, Kyle J.; Good, Morris S.; Waltar, Alan E.

    2007-07-28

    The Global Nuclear Energy Partnership (GNEP) has been proposed as a viable system in which to close the fuel cycle in a manner consistent with markedly expanding the global role of nuclear power while significantly reducing proliferation risks. A key part of this system relies on the development of actinide transmutation, which can only be effectively accomplished in a fast-spectrum reactor. The fundamental physics for fast reactors is well established. However, to achieve higher standards of safety and reliability, operate with longer intervals between outages, and achieve high operating capacity factors, new instrumentation and on-line monitoring capabilities will be required--during both fabrication and operation. Since the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor – II (EBR-II) reactors were operational in the USA, there have been major advances in instrumentation, not the least being the move to digital systems. Some specific capabilities have been developed outside the USA, but new or at least re-established capabilities will be required. In many cases the only available information is in reports and papers. New and improved sensors and instrumentation will be required. Advanced instrumentation has been developed for high-temperature/high-flux conditions in some cases, but most of the original researchers and manufacturers are retired or no longer in business.

  14. Small LBE-Cooled Fast Reactor for Expanding Market

    SciTech Connect

    Hiroshi Sekimoto; Shinichi Makino; Kunihiko Nakamura; Yoshio Kamishima; Takashi Kawakita

    2002-07-01

    A long-life safe simple small portable proliferation-resistant reactor is expected to solve many problems associating future energy and globally environmental problems. From discussions on mainly neutronics and safety points it has been shown that the heavy liquid metal cooled fast reactor is the best candidate to satisfy the above requirements. A lead-bismuth-eutectic (LBE) cooled fast reactor LSPR (LBE-Cooled Long-Life Safe Simple Small Portable Proliferation-Resistant Reactor) has been designed and continues to be improved. In the present paper a recent version of LSPR is presented. The total power of the present design is 150 MWt (53 MWe). During whole reactor life of 12 years the excess reactivity required for burnup is very low, and negative coolant dilatation coefficient is confirmed. This characteristic together with some other characteristics makes unprotected loss of flow (ULOF) accident inherently safe. It can survive even simultaneous rod run-out transient over power (UTOP), ULOF and unprotected loss of heat sink (ULOHS) accident without the help of an operator or active device. (authors)

  15. A Comparison of Long-Lived, Prolieration Resistant Fast Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-09-01

    Nuclear power is expected to play a significant role in meeting future electricity needs, and in significantly reducing emissions compared to fossil-fueled power plants. However, the next generation of nuclear power plants will be expected to demonstrate significant advancements in economics, safety, waste disposal, and proliferation resistance. Many reactor types have been proposed for “Generation IV”, some of which have been fast reactors. The work discussed in here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal of the entire project is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The goal of the work presented in this paper is to investigate and compare a variety of possible fuel types, looking for optimum economics for an actinide burning, low cost of electricity, reactor design using sodium or lead-bismuth as the coolant.

  16. A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors.

    Energy Science and Technology Software Center (ESTSC)

    1987-05-14

    Version 00 SLAROM solves the neutron integral transport equations to determine the flux distribution and spectra in a fast reactor lattice and calculates cell averaged effective cross sections. The code uses multigroup data of the type in DLC-111/JFS that use Bondarenko factors for resonance effects.

  17. Progress in reliability of fast reactor operation and new trends to increased inherent safety

    SciTech Connect

    Merk, Bruno; Stanculescu, Alexander; Chellapandi, Perumal; Hill, Robert

    2015-06-01

    The reasons for the renewed interest in fast reactors and an overview of the progress in sodium cooled fast reactor operation in the last ten years are given. The excellent operational performance of sodium cooled fast reactors in this period is highlighted as a sound basis for the development of new fast reactors. The operational performance of the BN-600 is compared and evaluated against the performance of German light water reactors to assess the reliability. The relevance of feedback effects for safe reactor design is described, and a new method for the enhancement of feedback effects in fast reactors is proposed. Experimental reactors demonstrating the inherent safety of advanced sodium cooled fast reactor designs are described and the potential safety improvements resulting from the use of fine distributed moderating material are discussed.

  18. Proposed fuel cycle for the Integral Fast Reactor

    SciTech Connect

    Burris, L.; Walters, L.C.

    1985-01-01

    One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor.

  19. Steam generator for liquid metal fast breeder reactor

    DOEpatents

    Gillett, James E.; Garner, Daniel C.; Wineman, Arthur L.; Robey, Robert M.

    1985-01-01

    Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

  20. Computational Neutronics Methods and Transmutation Performance Analyses for Fast Reactors

    SciTech Connect

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

    2007-03-01

    The once-through fuel cycle strategy in the United States for the past six decades has resulted in an accumulation of Light Water Reactor (LWR) Spent Nuclear Fuel (SNF). This SNF contains considerable amounts of transuranic (TRU) elements that limit the volumetric capacity of the current planned repository strategy. A possible way of maximizing the volumetric utilization of the repository is to separate the TRU from the LWR SNF through a process such as UREX+1a, and convert it into fuel for a fast-spectrum Advanced Burner Reactor (ABR). The key advantage in this scenario is the assumption that recycling of TRU in the ABR (through pyroprocessing or some other approach), along with a low capture-to-fission probability in the fast reactor’s high-energy neutron spectrum, can effectively decrease the decay heat and toxicity of the waste being sent to the repository. The decay heat and toxicity reduction can thus minimize the need for multiple repositories. This report summarizes the work performed by the fuel cycle analysis group at the Idaho National Laboratory (INL) to establish the specific technical capability for performing fast reactor fuel cycle analysis and its application to a high-priority ABR concept. The high-priority ABR conceptual design selected is a metallic-fueled, 1000 MWth SuperPRISM (S-PRISM)-based ABR with a conversion ratio of 0.5. Results from the analysis showed excellent agreement with reference values. The independent model was subsequently used to study the effects of excluding curium from the transuranic (TRU) external feed coming from the LWR SNF and recycling the curium produced by the fast reactor itself through pyroprocessing. Current studies to be published this year focus on analyzing the effects of different separation strategies as well as heterogeneous TRU target systems.

  1. A Comparison of Long-Lived, Proliferation Resistant Fast Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-09-01

    Various methods have been proposed to transmute and thus consume the current inventory of trans-uranic waste that exists in spent light-water-reactor fuel. These methods include both critical and sub-critical systems. The neutronics of metallic and nitride fuels loaded with 20-30wt% light-water-reactor plutonium plus minor actinides for use in a lead-bismuth and sodium cooled fast reactor are discussed, with an emphasis on the fuel cycle life and isotopic content. Calculations show that core life can extend beyond 20 years, and the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from 0.5 to 0.9 g/MWd.

  2. Multiple reheat helium Brayton cycles for sodium fast reactors

    SciTech Connect

    Haihua Zhao; Per F. Peterson

    2008-07-01

    Sodium fast reactors (SFR) traditionally adopt the steam Rankine cycle for power conversion. The resulting potential for water-sodium reaction remains a continuing concern which at least partly delays the SFR technology commercialization and is a contributor to higher capital cost. Supercritical CO2 provides an alternative, but is also capable of sustaining energetic chemical reactions with sodium. Recent development on advanced inert-gas Brayton cycles could potentially solve this compatibility issue, increase thermal efficiency, and bring down the capital cost close to light water reactors. In this paper, helium Brayton cycles with multiple reheat and intercooling states are presented for SFRs with reactor outlet temperatures in the range of 510°C to 650°C. The resulting thermal efficiencies range from 39% and 47%, which is comparable with supercritical recompression CO2 cycles (SCO2 cycle). A systematic comparison between multiple reheat helium Brayton cycle and the SCO2 cycle is given, considering compatibility issues, plant site cooling temperature effect on plant efficiency, full plant cost optimization, and other important factors. The study indicates that the multiple reheat helium cycle is the preferred choice over SCO2 cycle for sodium fast reactors.

  3. How Small Can Fast-Spectrum Space Reactors Get?

    SciTech Connect

    Hatton, Steven A.; El-Genk, Mohamed S.

    2006-01-20

    Fast neutron spectrum space reactors are an appropriate choice for high thermal powers, but for low powers, they may not satisfy the excess reactivity requirement while remaining sub-critical when immersed in wet sand and flooded with seawater following a launch abort accident. This paper identifies the smallest size fast spectrum, Sectored, Compact Reactor loaded with Single UN fuel pins (SCoRe-S7), which satisfy the requirements of cold clean excess reactivity > $4.00 and remains at least $1.00 subcritical at shutdown and in submersion conditions. Results indicate that increasing the diameter of the SCoRe-S core reduces its active height and the UN fuel enrichment, but increases the Spectrum-Shift Absorber (SSA) of 157GdN additive to the fuel. All SCoRe-S cores also have a 0.1 mm thick 157Gd2O3 SSA coating on the outer surface of the reactor vessel to reduce the effect of the wet sand reflector, while the SSA fuel additive reduces the effect on the criticality of the flooded reactor caused by thermal neutron fission. The active core height decreases from 42.4 cm for the smallest SCoRe-S7 to as much as to 37.4 cm for the largest core of SCoRe-S11. For a 1.8 MWth reactor thermal power the UN fuel specific power decreases from 17.0 in the SCoRe-S7 to 11.5 Wth/kg in the -S11. The corresponding reactor total mass, including the BeO reflector, increases from 440 kg to 512 kg.

  4. Practical Combinations of Light-Water Reactors and Fast-Reactors for Future Actinide Transmutation

    SciTech Connect

    Collins, Emory D; Renier, John-Paul

    2007-01-01

    Multicycle partitioning-transmutation (P-T) studies continue to show that use of existing light-water reactors (LWRs) and new advanced light-water reactors (ALWRs) can effectively transmute transuranic (TRU) actinides, enabling initiation of full actinide recycle much earlier than waiting for the development and deployment of sufficient fast reactor (FR) capacity. The combination of initial P-T cycles using LWRs/ALWRs in parallel with economic improvements to FR usage for electricity production, and a follow-on transition period in which FRs are deployed, is a practical approach to near-term closure of the nuclear fuel cycle with full actinide recycle.

  5. Planar ceramic membrane assembly and oxidation reactor system

    DOEpatents

    Carolan, Michael Francis; Dyer, legal representative, Kathryn Beverly; Wilson, Merrill Anderson; Ohrn, Ted R.; Kneidel, Kurt E.; Peterson, David; Chen, Christopher M.; Rackers, Keith Gerard; Dyer, Paul Nigel

    2009-04-07

    Planar ceramic membrane assembly comprising a dense layer of mixed-conducting multi-component metal oxide material, wherein the dense layer has a first side and a second side, a porous layer of mixed-conducting multi-component metal oxide material in contact with the first side of the dense layer, and a ceramic channeled support layer in contact with the second side of the dense layer. The planar ceramic membrane assembly can be used in a ceramic wafer assembly comprising a planar ceramic channeled support layer having a first side and a second side; a first dense layer of mixed-conducting multi-component metal oxide material having an inner side and an outer side, wherein the inner side is in contact with the first side of the ceramic channeled support layer; a first outer support layer comprising porous mixed-conducting multi-component metal oxide material and having an inner side and an outer side, wherein the inner side is in contact with the outer side of the first dense layer; a second dense layer of mixed-conducting multi-component metal oxide material having an inner side and an outer side, wherein the inner side is in contact with the second side of the ceramic channeled layer; and a second outer support layer comprising porous mixed-conducting multi-component metal oxide material and having an inner side and an outer side, wherein the inner side is in contact with the outer side of the second dense layer.

  6. Planar ceramic membrane assembly and oxidation reactor system

    DOEpatents

    Carolan, Michael Francis; Dyer, legal representative, Kathryn Beverly; Wilson, Merrill Anderson; Ohm, Ted R.; Kneidel, Kurt E.; Peterson, David; Chen, Christopher M.; Rackers, Keith Gerard; Dyer, deceased, Paul Nigel

    2007-10-09

    Planar ceramic membrane assembly comprising a dense layer of mixed-conducting multi-component metal oxide material, wherein the dense layer has a first side and a second side, a porous layer of mixed-conducting multi-component metal oxide material in contact with the first side of the dense layer, and a ceramic channeled support layer in contact with the second side of the dense layer. The planar ceramic membrane assembly can be used in a ceramic wafer assembly comprising a planar ceramic channeled support layer having a first side and a second side; a first dense layer of mixed-conducting multi-component metal oxide material having an inner side and an outer side, wherein the inner side is in contact with the first side of the ceramic channeled support layer; a first outer support layer comprising porous mixed-conducting multi-component metal oxide material and having an inner side and an outer side, wherein the inner side is in contact with the outer side of the first dense layer; a second dense layer of mixed-conducting multi-component metal oxide material having an inner side and an outer side, wherein the inner side is in contact with the second side of the ceramic channeled layer; and a second outer support layer comprising porous mixed-conducting multi-component metal oxide material and having an inner side and an outer side, wherein the inner side is in contact with the outer side of the second dense layer.

  7. "Ultra"-Fast Fracture Strength of Advanced Structural Ceramic Materials Studied at Elevated Temperatures

    NASA Technical Reports Server (NTRS)

    Choi, Sung R.; Gyekenyesi, John P.

    1999-01-01

    The accurate determination of inert strength is important in reliable life prediction of structural ceramic components. At ambient temperature, the inert strength of a brittle material is typically regarded as free of the effects of slow crack growth due to stress corrosion. Therefore, the inert strength can be determined either by eliminating active species, especially moisture, with an appropriate inert medium, or by using a very high test rate. However, at elevated temperatures, the concept or definition of the inert strength of brittle ceramic materials is not clear, since temperature itself is a degrading environment, resulting in strength degradation through slow crack growth and/or creep. Since the mechanism to control strength is rate-dependent viscous flow, the only conceivable way to determine the inert strength at elevated temperatures is to utilize a very fast test rate that either minimizes the time for or eliminates slow crack growth. Few experimental studies have measured the elevated-temperature, inert (or "ultra"-fast fracture) strength of advanced ceramics. At the NASA Lewis Research Center, an experimental study was initiated to better understand the "ultra"-fast fracture strength behavior of advanced ceramics at elevated temperatures. Fourteen advanced ceramics - one alumina, eleven silicon nitrides, and two silicon carbides - have been tested using constant stress-rate (dynamic fatigue) testing in flexure with a series of stress rates including the "ultra"-fast stress rate of 33 000 MPa/sec with digitally controlled test frames. The results for these 14 advanced ceramics indicate that, notwithstanding possible changes in flaw populations as well as flaw configurations because of elevated temperatures, the strength at 33 000 MPa/sec approached the room-temperature strength or reached a higher value than that determined at the conventional test rate of 30 MPa/sec. On the basis of the experimental data, it can be stated that the elevated

  8. Comparative breeding characteristics of fusion and fast reactors.

    PubMed

    Fortescue, P

    1977-06-17

    Expressions are developed to allow ready comparison of a hybrid fission-fusion plant and a fast breeder with respect to the number of thermal reactors that their fissile production could support, both for their feed requirements and for the new inventory needs of an expanding industry. These relations are expressed in terms of the neutron multiplication factor obtained in the fusion blanket, and the analogous quantities represented by the conversion ratios of the fast and thermal fission associated with the comparison. Results are presented graphically both for the steady state and for industries of arbitrary growth rate, and include the influence of tritium production requirements. Even a modest blanket neutron multiplication factor could enable the hybrid fusion system greatly to outperform the fast breeder on this simple basis of material balances. PMID:17831749

  9. Model and Mockup of DEMO Ceramic Breeding Zone for Testing in the IVV-2M Reactor

    SciTech Connect

    Kapychev, V.; Gorokhov, V.; Davydov, D.

    2001-01-15

    A model of a ceramic breeding zone has been designed and manufactured for testing in the IVV-2M fission reactor. The model contains both lithium orthosilicate and beryllium spheroid particles 1 to 2 mm in diameter. The structural material is ferritic-martensitic stainless steel. In addition, a mockup of the model for verification of neutronics calculation methods has been designed and manufactured. A special facility (Functional In-Reactor Investigations of Tritium-Breeding Models) has been designed, fabricated, and assembled at the reactor for studying the kinetics of tritium extraction from the ceramic material under irradiation. Neutronic and thermo-hydraulic calculations and an initial measurement of tritium release from the model under reactor irradiation have been performed.

  10. Properties of plutonium and its alloys for use as fast reactor fuels

    NASA Astrophysics Data System (ADS)

    Hecker, Siegfried S.; Stan, Marius

    2008-12-01

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher melting U-Pu-Zr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  11. Reactivity Control Schemes for Fast Spectrum Space Nuclear Reactors

    SciTech Connect

    Craft, Aaron E.; King, Jeffrey C.

    2008-01-21

    Several different reactivity control schemes are considered for future space nuclear reactor power systems. Each of these control schemes uses a combination of boron carbide absorbers and/or beryllium oxide reflectors to achieve sufficient reactivity swing to keep the reactor subcritical during launch and to provide sufficient excess reactivity to operate the reactor over its expected 7-15 year lifetime. The size and shape of the control system directly impacts the size and mass of the space reactor's reflector and shadow shield, leading to a tradeoff between reactivity swing and total system mass. This paper presents a trade study of drum, shutter, and petal control schemes based on reactivity swing and mass effects for a representative fast-spectrum, gas-cooled reactor. For each control scheme, the dimensions and composition of the core are constant, and the reflector is sized to provide $5 of cold-clean excess reactivity with each configuration in its most reactive state. The advantages and disadvantages of each configuration are discussed, along with optimization techniques and novel geometric approaches for each scheme.

  12. EXTENDING SODIUM FAST REACTOR DRIVER FUEL USE TO HIGHER TEMPERATURES

    SciTech Connect

    Douglas L. Porter

    2011-02-01

    Calculations of potential sodium-cooled fast reactor fuel temperatures were performed to estimate the effects of increasing the outlet temperature of a given fast reactor design by increasing pin power, decreasing assembly flow, or increasing inlet temperature. Based upon experience in the U.S., both metal and mixed oxide (MOX) fuel types are discussed in terms of potential performance effects created by the increased operating temperatures. Assembly outlet temperatures of 600, 650 and 700 °C were used as goal temperatures. Fuel/cladding chemical interaction (FCCI) and fuel melting, as well as challenges to the mechanical integrity of the cladding material, were identified as the limiting phenomena. For example, starting with a recent 1000 MWth fast reactor design, raising the outlet temperature to 650 °C through pin power increase increased the MOX centerline temperature to more than 3300 °C and the metal fuel peak cladding temperature to more than 700 °C. These exceeded limitations to fuel performance; fuel melting was limiting for MOX and FCCI for metal fuel. Both could be alleviated by design ‘fixes’, such as using a barrier inside the cladding to minimize FCCI in the metal fuel, or using annular fuel in the case of MOX. Both would also require an advanced cladding material with improved stress rupture properties. While some of these are costly, the benefits of having a high-temperature reactor which can support hydrogen production, or other missions requiring high process heat may make the extra costs justified.

  13. 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.

  14. Fast Pyrolysis of Agricultural Wastes in a Fluidized Bed Reactor

    NASA Astrophysics Data System (ADS)

    Wang, X. H.; Chen, H. P.; Yang, H. P.; Dai, X. M.; Zhang, S. H.

    Solid biomass can be converted into liquid fuel through fast pyrolysis, which is convenient to be stored and transported with potential to be used as a fossil oil substitute. In China, agricultural wastes are the main biomass materials, whose pyrolysis process has not been researched adequately compared to forestry wastes. As the representative agricultural wastes in China, peanut shell and maize stalk were involved in this paper and pine wood sawdust was considered for comparing the different pyrolysis behaviors of agricultural wastes and forestry wastes. Fast pyrolysis experiments were carried out in a bench-scale fluidized-bed reactor. The bio-oil yieldsof peanut shell and maize stalk were obviously lower than that ofpine sawdust. Compared with pine sawdust, the char yields of peanut shell and maize stalk were higher but the heating value of uncondensable gaswas lower. This means that the bio-oil cost will be higher for agricultural wastes if taking the conventional pyrolysis technique. And the characteristic and component analysis resultsof bio-oil revealed that the quality of bio-oil from agricultural wastes, especially maize stalk, was worse than that from pine wood. Therefore, it is important to take some methods to improve the quality of bio-oilfrom agricultural wastes, which should promote the exploitation of Chinese biomass resources through fast pyrolysis in afluidized bed reactor.

  15. Pre-Combustion Carbon Dioxide Capture by a New Dual Phase Ceramic-Carbonate Membrane Reactor

    SciTech Connect

    Lin, Jerry

    2014-09-30

    This report documents synthesis, characterization and carbon dioxide permeation and separation properties of a new group of ceramic-carbonate dual-phase membranes and results of a laboratory study on their application for water gas shift reaction with carbon dioxide separation. A series of ceramic-carbonate dual phase membranes with various oxygen ionic or mixed ionic and electronic conducting metal oxide materials in disk, tube, symmetric, and asymmetric geometric configurations was developed. These membranes, with the thickness of 10 μm to 1.5 mm, show CO2 permeance in the range of 0.5-5×10-7 mol·m-2·s-1·Pa-1 in 500-900oC and measured CO2/N2 selectivity of up to 3000. CO2 permeation mechanism and factors that affect CO2 permeation through the dual-phase membranes have been identified. A reliable CO2 permeation model was developed. A robust method was established for the optimization of the microstructures of ceramic-carbonate membranes. The ceramic-carbonate membranes exhibit high stability for high temperature CO2 separations and water gas shift reaction. Water gas shift reaction in the dual-phase membrane reactors was studied by both modeling and experiments. It is found that high temperature syngas water gas shift reaction in tubular ceramic-carbonate dual phase membrane reactor is feasible even without catalyst. The membrane reactor exhibits good CO2 permeation flux, high thermal and chemical stability and high thermal shock resistance. Reaction and separation conditions in the membrane reactor to produce hydrogen of 93% purity and CO2 stream of >95% purity, with 90% CO2 capture have been identified. Integration of the ceramic-carbonate dual-phase membrane reactor with IGCC process for carbon dioxide capture was analyzed. A methodology was developed to identify optimum operation conditions for a membrane tube of given dimensions that would treat coal syngas with targeted performance. The calculation results show that the dual-phase membrane reactor could

  16. Current status and directions for fast reactor reprocessing

    SciTech Connect

    Burch, W.D.

    1983-01-01

    The development of fast breeder reactors (FBRs) for commercial electric power production has been under way in several countries for more than 20 years. In the United States as elsewhere, early work was centered on small reactors to prove the feasibility of concepts and later was followed by larger reactors to test engineering features and to develop fuel technology. In the early 1970s, with the perceived crisis in electrical generation expected late in this century, major efforts were mounted to plan and carry out comprehensive development programs to ensure the capability to develop and begin using this new form of nuclear power by the end of this century. This comprehensive effort included the first serious efforts directed toward the supporting fuel cycle activities. However, because of the effects of the oil price rise and resulting conservation, a slowdown of industrial growth, and cut-backs in energy needs, there has been a decline in program activities. Unlike the fuel cycle for light-water reactors (LWRs), where supply and the back-end recycle and/or waste disposal activities can largely be uncoupled, recovery and recycle of fissile materials in spent fuel must be accomplished in one or two years in a practical breeder system. 3 references.

  17. Innovative Fast Reactors: Impact of Fuel Composition on Reactivity Coefficients

    SciTech Connect

    G. Palmiotti; M. Salvatores; M. Asswaroongruengchot

    2009-12-01

    A major challenge for future Fast Reactors could be the recycling of minor actinides (MA) in the core fuel, in order to minimize wastes and contribute to meet both the sustainability objective and the reduction of the burden on a geological disposal. Although the most outstanding issues will be found in the development and validation of the appropriate fuels, the presence of MA in the core can potentially deteriorate the core reactivity coefficients. In the present paper we will show however that there is no physical limit to the amount of MA in the core fuel, but that a careful physics analysis can indicate the most appropriate measures to reduce the MA impact on the reactivity coefficients, and in particular, for Na cooled reactors, on the Na void reactivity coefficient.

  18. METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

    DOEpatents

    Koch, L.J.

    1959-01-20

    A specific arrangement of the fertile material and fissionable material in the active portion of a fast reactor to achieve improvement in performance and to effectively lower the operating temperatures in the center of the reactor is described. According to this invention a group of fuel elements containing fissionable material are assembled to form a hollow fuel core. Elements containing a fertile material, such as depleted uranium, are inserted into the interior of the fuel core to form a central blanket. Additional elemenis of fertile material are arranged about the fuel core to form outer blankets which in tunn are surrounded by a reflector. This arrangement of fuel core and blankets results in substantial flattening of the flux pattern.

  19. Decay heat removal in GEN IV gas cooled fast reactors.

    SciTech Connect

    Cheng, L. Y.; Wei, T. Y. C.

    2009-08-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  20. Fast reactor core concepts to improve transmutation efficiency

    SciTech Connect

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-12-31

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

  1. Steam generator tubing development for commercial fast breeder reactors

    SciTech Connect

    Sessions, C.E.; Uber, C.F.

    1981-11-01

    The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs.

  2. Five Requirements for Nuclear Energy and CANDLE Fast Reactor

    NASA Astrophysics Data System (ADS)

    Sekimoto, Hiroshi

    2010-06-01

    The Center for Research into Innovative Nuclear Energy Systems (CRINES) was established in order to succeed the COE-INES mission after finishing this program in Tokyo Tech. CRINES considers nuclear energy should satisfy 5 requirements; sustainability as basic energy, solving 3 problems inherent to accidents, radioactive waste and nuclear bomb, and economical acceptance. Characteristics of CANDLE fast reactor are discussed for these requirements. It satisfies 4 requirements; sustainability and solving 3 inherent problems. For the remaining requirement for economy, a high potential to satisfy this requirement is also shown.

  3. Five Requirements for Nuclear Energy and CANDLE Fast Reactor

    SciTech Connect

    Sekimoto, Hiroshi

    2010-06-22

    The Center for Research into Innovative Nuclear Energy Systems (CRINES) was established in order to succeed the COE-INES mission after finishing this program in Tokyo Tech. CRINES considers nuclear energy should satisfy 5 requirements; sustainability as basic energy, solving 3 problems inherent to accidents, radioactive waste and nuclear bomb, and economical acceptance. Characteristics of CANDLE fast reactor are discussed for these requirements. It satisfies 4 requirements; sustainability and solving 3 inherent problems. For the remaining requirement for economy, a high potential to satisfy this requirement is also shown.

  4. Fast reactor core concepts to improve transmutation efficiency

    NASA Astrophysics Data System (ADS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-12-01

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

  5. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    SciTech Connect

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  6. Development of ceramic membrane reactors for high temperature gas cleanup. Final report

    SciTech Connect

    Roberts, D.L.; Abraham, I.C.; Blum, Y.; Gottschlich, D.E.; Hirschon, A.; Way, J.D.; Collins, J.

    1993-06-01

    The objective of this project was to develop high temperature, high pressure catalytic ceramic membrane reactors and to demonstrate the feasibility of using these membrane reactors to control gaseous contaminants (hydrogen sulfide and ammonia) in integrated gasification combined cycle (IGCC) systems. Our strategy was to first develop catalysts and membranes suitable for the IGCC application and then combine these two components as a complete membrane reactor system. We also developed a computer model of the membrane reactor and used it, along with experimental data, to perform an economic analysis of the IGCC application. Our results have demonstrated the concept of using a membrane reactor to remove trace contaminants from an IGCC process. Experiments showed that NH{sub 3} decomposition efficiencies of 95% can be achieved. Our economic evaluation predicts ammonia decomposition costs of less than 1% of the total cost of electricity; improved membranes would give even higher conversions and lower costs.

  7. Modular Code and Data System for Fast Reactor Neutronics Analyses

    SciTech Connect

    RIMPAULT, G.

    2008-06-30

    Version 00. The European Reactor ANalysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R&D organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core follow-up modules (connected in the

  8. Modular Code and Data System for Fast Reactor Neutronics Analyses

    Energy Science and Technology Software Center (ESTSC)

    2008-06-30

    Version 00. The European Reactor ANalysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R&Dmore » organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core follow-up modules (connected

  9. Shape optimization of a sodium cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Schmitt, Damien; Allaire, Grégoire; Pantz, Olivier; Pozin, Nicolas

    2014-06-01

    Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth.(1, 2) Usual optimization methods for core conception are based on a parametric description of a given core design(3).(4) New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints.(5, 6) First studies show that these methods could be applied to sodium cooled core conception.(7) In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a get realistic core layout. Its caracteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas.

  10. The search for advanced remote technology in fast reactor reprocessing

    SciTech Connect

    Burch, W.D.; Herndon, J.N.; Stradley, J.G.

    1990-01-01

    Research and development in fast reactor reprocessing has been under way about 20 years in several countries throughout the world. During the past decade in France and the United Kingdom, active development programs have been carried out in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the EBR-II facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. Germany and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in all of these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper will focus principally on the search for improved facility concepts and better maintenance systems in the CFRP and, in turn, on how developments at ORNL have influenced the technology elsewhere.

  11. Ceramic membranes for catalytic membrane reactors with high ionic conductivities and low expansion properties

    DOEpatents

    Mackay, Richard; Sammells, Anthony F.

    2000-01-01

    Ceramics of the composition: Ln.sub.x Sr.sub.2-x-y Ca.sub.y B.sub.z M.sub.2-z O.sub.5+.delta. where Ln is an element selected from the fblock lanthanide elements and yttrium or mixtures thereof; B is an element selected from Al, Ga, In or mixtures thereof; M is a d-block transition element of mixtures thereof; 0.01.ltoreq.x.ltoreq.1.0; 0.01.ltoreq.y.ltoreq.0.7; 0.01.ltoreq.z.ltoreq.1.0 and .delta. is a number that varies to maintain charge neutrality are provided. These ceramics are useful in ceramic membranes and exhibit high ionic conductivity, high chemical stability under catalytic membrane reactor conditions and low coefficients of expansion. The materials of the invention are particularly useful in producing synthesis gas.

  12. The development and application of an improved reactor analysis model for fast reactors

    NASA Astrophysics Data System (ADS)

    Hou, Jia

    Accuracy in neutron cross sections calculation and consistency in reactor physics are fundamental requirements in advanced nuclear reactor design and analysis. The work presented in this dissertation focuses on the development and advanced application of a reactor analysis model with updated cross section libraries that is suitable for online cross section generation for fast reactors. Research has been performed in two areas of interest in reactor physics. The first target of the research is to develop effcient modeling capacity of the 1- D lattice code MICROX-2 for its neutron spectrum calculation based on Collision Probability Method (CPM). Expanded master cross section libraries have been generated based on updated nuclear data and optimized fine-group energy structure to accommodate both thermal and fast reactor spectra as well as to comply with the need for advanced fuel cycle analysis. After verifying the new libraries, the solution methods have been reviewed and updated, including the update of interpolation scheme for resonance self-shielding factors and improvement of spatial self-shielding models for various fuel assembly geometries. The assessment of the updated lattice calculation models has shown that the prediction accuracy of lattice properties represented by the eigenvalue and reaction rate ratios is improved, especially for fast neutron spectrum lattices of which the importance of neutrons in the unresolved energy range is high. The second target of the research is to improve the accuracy of few-group nuclear cross section generation for the reactor core calculation. A 2-D pin-by-pin lattice model has been developed based on embedded CPM within the framework of the Nodal Expansion Method (NEM), which is capable of modeling the heterogeneity of the fuel assembly. Then, an online cross section generation methodology along with discontinuity factors has been developed based on Iterative Diffusion- Diffusion Methodology (IDDM), which can minimize the

  13. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    SciTech Connect

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2002-07-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B{sub 4}C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

  14. Fuels for sodium-cooled fast reactors: US perspective

    NASA Astrophysics Data System (ADS)

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.

    2007-09-01

    The US experience with mixed oxide, metal, and mixed carbide fuels is substantial, comprised of irradiation of over 50 000 MOX rods, over 130 000 metal rods, and 600 mixed carbide rods, in EBR-II and FFTF alone. All three types have been demonstrated capable of fuel utilization at or above 200 GWd/MTHM. To varying degrees, life-limiting phenomena for each type have been identified and investigated, and there are no disqualifying safety-related fuel behaviors. All three fuel types appear capable of meeting requirements of sodium-cooled fast reactor fuels, with reliability of mixed oxide and metal fuel well established. Improvements in irradiation performance of cladding and duct alloys have been a key development in moving these fuel designs toward higher-burnup potential. Selection of one fuel system over another will depend on circumstances particular to the application and on issues other than fuel performance, such as fabrication cost or overall system safety performance.

  15. Sodium fast reactor fuels and materials : research needs.

    SciTech Connect

    Denman, Matthew R.; Porter, Douglas; Wright, Art; Lambert, John; Hayes, Steven; Natesan, Ken; Ott, Larry J.; Garner, Frank; Walters, Leon; Yacout, Abdellatif

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  16. Limitations of eddy current testing in a fast reactor environment

    NASA Astrophysics Data System (ADS)

    Wu, Tao; Bowler, John R.

    2016-02-01

    The feasibility of using eddy current probes for detecting flaws in fast nuclear reactor structures has been investigated with the aim of detecting defects immersed in electrically conductive coolant including under liquid sodium during standby. For the inspections to be viable, there is a need to use an encapsulated sensor system that can be move into position with the aid of visualization tools. The initial objective being to locate the surface to be investigated using, for example, a combination of electromagnetic sensors and sonar. Here we focus on one feature of the task in which eddy current probe impedance variations due to interaction with the external surface of a tube are evaluated in order to monitor the probe location and orientation during inspection.

  17. Designing a SCADA system simulator for fast breeder reactor

    NASA Astrophysics Data System (ADS)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  18. FFTF (Fast Flux Test Facility) reactor shutdown system reliability reevaluation

    SciTech Connect

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations.

  19. Materials development for a fast breeder reactor steam generator concept

    SciTech Connect

    Sessions, C.E.; Reynolds, S.D. Jr.; Hebbar, M.A.; Lewis, J.F.; Kiefer, J.H.

    1981-11-01

    The progress achieved since 1977 in the important area of materials and processes development of fast reactor steam generator development is summarized. The two distinguishing features of the proposed Westinghouse-Tampa steam generator concept are the convoluted shell expansion joint (CSEJ) and the double-wall tubing with a third fluid leak detection capability. A 2/one quarter/ Cr-1 Mo low alloy steel will be used for all important parts of the generator including the CSEJ and the tubes. Other areas in which progress was made include tube-to-tubesheet (T/TS) welding, post-weld heat treatment (PWHT), tube expansion, and development of materials specifications for prototype and future plant materials. 8 refs.

  20. 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.

  1. Advanced sodium fast reactor accident source terms : research needs.

    SciTech Connect

    Powers, Dana Auburn; Clement, Bernard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  2. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    SciTech Connect

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  3. Reactor Physics Characterization of Transmutation Targeting Options in a Sodium Fast Reactor

    SciTech Connect

    Samuel E. Bays

    2007-04-01

    In sodium fast reactor designs, the fuel related inherent negative reactivity feedback is accomplished mainly through parasitic capture in U-238. However for an efficient minor actinide burning system, it is desirable to reduce or eliminate U-238 entirely to suppress further transuranic actinide generation. Consequently, reactivity feedback is accomplished by enhancing axial neutron streaming during a loss of coolant void situation. This is done by flattening “pancake” the active core geometry. Flattening the reactor also increases axial leakage which removes neutrons that could otherwise be used to destroy minor actinides. Therefore, it is important to tailor the neutron spectrum in the core for optimized feedback and minor actinide destruction simultaneously by using minor actinide and fission product targets.

  4. Medium Power Lead Alloy Fast Reactor Balance of Plant Options

    SciTech Connect

    Vaclav Dosta; Pavel Hejzlar; Neil E. Todreas; Jacopo Buongiorno

    2004-09-01

    Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature (~550°C) compared to that of light water reactors (~300°C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO2 (S-CO2) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545°C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312°C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO2 cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be the S-CO2 recompression cycle, because it

  5. Shear punch testing of candidate reactor materials after irradiation in fast reactors and spallation environments

    NASA Astrophysics Data System (ADS)

    Maloy, S. A.; Romero, T. J.; Hosemann, P.; Toloczko, M. B.; Dai, Y.

    2011-10-01

    Ferritic/martensitic steels and nickel-base superalloys are potential materials for use in spallation targets and fusion and fast reactors. To investigate the effects of irradiation on these materials, tests were performed after irradiation in the high energy proton beam at the Paul Scherrer Institute (SINQ Target Irradiation Program (STIP), 570 MeV), as well on specimens obtained from a driver duct irradiated in the Fast Flux Test Facility (FFTF). Dose accumulations were up to 18 dpa for STIP irradiations (at 147-406 °C) and up to 155 dpa in FFTF (at 383-505 °C). The helium/dpa ratios ranged from 0.2 to 80 appm/dpa. Mechanical testing was performed at 25 °C. Increases in shear yield and shear maximum stress with increasing dose mirrored the results observed from companion tensile tests.

  6. Development of ceramic membrane reactors for high temperature gas cleanup

    SciTech Connect

    Roberts, D.L.; Way, J.D.

    1992-01-01

    The more difficult part of developing a membrane reactor was making a membrane with suitable permselectivity. Since both the ammonia and H[sub 2]S decomposition reactions produce H[sub 2], we want a membrane that selectively permeates hydrogen. We have taken the approach of coating a substrate that has fine pores (a microfilter or ultrafilter) with a completely dense layer and then (except for palladium coatings) malting super fine pores in the dense layer by leaching or pyrolysis. We used four different substrates (Norton 0.2 alumina monolith, Refractron alumina microfilters, Vycor glass, Alcoa ultrafilter) and five different coating materials (Poly N-methyl silazane, Aremco 617 alumina-based glaze, Polycyclohydridomethyl silazane, Aluminum phosphorus oxides, palladium). Only the palladium films on an alumina ultrafilter could be made regularly with a negligible defect population. The other approaches were plagued with cracks and poor reproducibility. We made palladium films on a tubular alumina ultrafilter (US Filter, Warrendale, PA) by electroless plating from a platinum amine complex solution in hydrazine (Rhoda, 1959). The best results were achieved with an ultrafilter pore size of 100 angstroms and with pretreatment of the surface in a stannic/stannous chloride solution. The films were about 5 microns thick and appeared defect free under electron microscopy. Because defect-free palladium films are infinitely selective to hydrogen with respect to the other components of gasifier product gas, we expect that the selectivity of our palladium films will exceed 1000 even with the inevitable defects that accompany real membranes. We have analyzed the performance and economics of membrane reactors for H[sub 2]S and NH[sub 3] decomposition using the reaction rates determined experimentally in our catalyst development work and using the permeation behavior of palladium films reported in the literature.

  7. Development of ceramic membrane reactors for high temperature gas cleanup

    SciTech Connect

    Roberts, D.L.; Way, J.D.

    1992-11-01

    The more difficult part of developing a membrane reactor was making a membrane with suitable permselectivity. Since both the ammonia and H{sub 2}S decomposition reactions produce H{sub 2}, we want a membrane that selectively permeates hydrogen. We have taken the approach of coating a substrate that has fine pores (a microfilter or ultrafilter) with a completely dense layer and then (except for palladium coatings) malting super fine pores in the dense layer by leaching or pyrolysis. We used four different substrates (Norton 0.2 alumina monolith, Refractron alumina microfilters, Vycor glass, Alcoa ultrafilter) and five different coating materials (Poly N-methyl silazane, Aremco 617 alumina-based glaze, Polycyclohydridomethyl silazane, Aluminum phosphorus oxides, palladium). Only the palladium films on an alumina ultrafilter could be made regularly with a negligible defect population. The other approaches were plagued with cracks and poor reproducibility. We made palladium films on a tubular alumina ultrafilter (US Filter, Warrendale, PA) by electroless plating from a platinum amine complex solution in hydrazine (Rhoda, 1959). The best results were achieved with an ultrafilter pore size of 100 angstroms and with pretreatment of the surface in a stannic/stannous chloride solution. The films were about 5 microns thick and appeared defect free under electron microscopy. Because defect-free palladium films are infinitely selective to hydrogen with respect to the other components of gasifier product gas, we expect that the selectivity of our palladium films will exceed 1000 even with the inevitable defects that accompany real membranes. We have analyzed the performance and economics of membrane reactors for H{sub 2}S and NH{sub 3} decomposition using the reaction rates determined experimentally in our catalyst development work and using the permeation behavior of palladium films reported in the literature.

  8. Simultaneous nuclear data target accuracy study for innovative fast reactors.

    SciTech Connect

    Aliberti, G.; Palmiotti, G.; Salvatores, M.; Nuclear Engineering Division; INL; CEA Cadarache

    2007-01-01

    The present paper summarizes the major outcomes of a study conducted within a Nuclear Energy Agency Working Party on Evaluation Cooperation (NEA WPEC) initiative aiming to investigate data needs for future innovative nuclear systems, to quantify them and to propose a strategy to meet them. Within the NEA WPEC Subgroup 26 an uncertainty assessment has been carried out using covariance data recently processed by joint efforts of several US and European Labs. In general, the uncertainty analysis shows that for the wide selection of fast reactor concepts considered, the present integral parameters uncertainties resulting from the assumed uncertainties on nuclear data are probably acceptable in the early phases of design feasibility studies. However, in the successive phase of preliminary conceptual designs and in later design phases of selected reactor and fuel cycle concepts, there will be the need for improved data and methods, in order to reduce margins, both for economic and safety reasons. It is then important to define as soon as possible priority issues, i.e. which are the nuclear data (isotope, reaction type, energy range) that need improvement, in order to quantify target accuracies and to select a strategy to meet the requirements needed (e.g. by some selected new differential measurements and by the use of integral experiments). In this context one should account for the wide range of high accuracy integral experiments already performed and available in national or, better, international data basis, in order to indicate new integral experiments that will be needed to account for new requirements due to innovative design features, and to provide the necessary full integral data base to be used for validation of the design simulation tools.

  9. Elevated-Temperature "Ultra" Fast Fracture Strength of Advanced Ceramics: An Approach to Elevated-Temperature "Inert" Strength

    NASA Technical Reports Server (NTRS)

    Choi, S. R.; Gyekenyesi, J. P.

    1999-01-01

    The determination of "ultra" fast fracture strengths of five silicon nitride ceramics at elevated temperatures has been made by using constant stress-rate ("dynamic fatigue") testing with a series of "ultra" fast test rates. The test material included four monolithic and one SiC whisker-reinforced composite silicon nitrides. Of the five test materials, four silicon nitrides exhibited the elevated -temperature strengths that approaches their respective room-temperature strengths at an "ultra" fast test rate of 3.3 x 10(exp 4) MPa/s. This implies that slow cracks growth responsible for elevated-temperature failure can be eliminated or minimized by using the "ultra" fast test rate. These ongoing experimental results have shed light on laying a theoretical and practical foundation on the concept and definition of elevated-temperature "inert" strength behavior of advanced ceramics.

  10. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    SciTech Connect

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  11. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    SciTech Connect

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma; Al Rashdan, Ahmad; Tsvetkov, Pavel Valeryevich; Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  12. Role of fast reactor and its cycle to reduce nuclear waste burden

    SciTech Connect

    Arie, Kazuo; Oomori, Takashi; Okita, Takeshi; Kawashima, Masatoshi; Kotake, Shoji; Fuji-ie, Yoichi

    2013-07-01

    The role of the metal fuel fast reactor with recycling of actinides and the five long-lived fission products based on the concept of the Self-Consistent Nuclear Energy System has been examined by evaluating the reduction of nuclear wastes during the transition period to this reactor system. The evaluation was done in comparison to an LWR once-through case and a conventional actinide recycling oxide fast reactor. As a result, it is quantitatively clarified that a metal fuel fast reactor with actinide and the five long-lived fission products (I{sup 129}, Tc{sup 99}, Zr{sup 93}, Cs{sup 135} and Sn{sup 126}) recycling could play a significant role in reducing the nuclear waste burden including the current LWR wastes. This can be achieved by using a fast neutron spectrum reactor enhanced with metal fuel that brings high capability as a 'waste burner'. (authors)

  13. A compact breed and burn fast reactor using spent nuclear fuel blanket

    SciTech Connect

    Hartanto, D.; Kim, Y.

    2012-07-01

    A long-life breed-and-burn (B and B) type fast reactor has been investigated from the neutronics points of view. The B and B reactor has the capability to breed the fissile fuels and use the bred fuel in situ in the same reactor. In this work, feasibility of a compact sodium-cooled B and B fast reactor using spent nuclear fuel as blanket material has been studied. In order to derive a compact B and B fast reactor, a tight fuel lattice and relatively large fuel pin are used to achieve high fuel volume fraction. The core is initially loaded with an LEU (Low Enriched Uranium) fuel and a metallic fuel is used in the core. The Monte Carlo depletion has been performed for the core to see the long-term behavior of the B and B reactor. Several important parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, fission power, and fast neutron fluence, are analyzed through Monte Carlo reactor analysis. Evolution of the core fuel composition is also analyzed as a function of burnup. Although the long-life small B and B fast reactor is found to be feasible from the neutronics point of view, it is characterized to have several challenging technical issues including a very high fast neutron fluence of the structural materials. (authors)

  14. Delayed neutron signal characterization in a fast reactor

    SciTech Connect

    Gross, K.C.; Strain, R.V.

    1980-01-01

    Experimental and analytical techniques have been developed for delayed neutron (DN) signal analysis and characterization that can provide diagnostic information to augment data from cover-gas analyses in the detection and identification of breached elements in an LMFBR. Eleven flow reduction tests have been run in EBR-II to provide base data support for predicting DN signal characteristics during exposed fuel operation. Results from the tests demonstrate the feasibility and practicability of response-analysis techniques for determining the transit time, T/sub tr/, for DN emitters traveling from the core to the detector, and the isotopic holdup time, T/sub h/, of DN precursors in the fuel element. T/sub tr/ has been found to vary with the relative grid location of the DN source, and T/sub h/ is affected by the form of fuel exposed to the coolant as well as the condition of the breach site. These parameters are incorporated into a mathematical formulism that enables one to compute for any exposed-fuel test an equivalent recoil area. This concept provides a basis for comparison of different run-beyond-cladding-breach tests in fast reactors.

  15. Alloys for a liquid metal fast breeder reactor

    DOEpatents

    Rowcliffe, Arthur F.; Bleiberg, Melvin L.; Diamond, Sidney; Bajaj, Ram

    1979-01-01

    An essentially gamma-prime precipitation-hardened iron-chromium-nickel alloy has been designed with emphasis on minimum nickel and chromium contents to reduce the swelling tendencies of these alloys when used in liquid metal fast breeder reactors. The precipitation-hardening components have been designed for phase stability and such residual elements as silicon and boron, also have been selected to minimize swelling. Using the properties of these alloys in one design would result in an increased breeding ratio over 20% cold worked stainless steel, a reference material, of 1.239 to 1.310 and a reduced doubling time from 15.8 to 11.4 years. The gross stoichiometry of the alloying composition comprises from about 0.04% to about 0.06% carbon, from about 0.05% to about 1.0% silicon, up to about 0.1% zirconium, up to about 0.5% vanadium, from about 24% to about 31% nickel, from 8% to about 11% chromium, from about 1.7% to about 3.5% titanium, from about 1.0% to about 1.8% aluminum, from about 0.9% to about 3.7% molybdenum, from about 0.04% to about 0.8% boron, and the balance iron with incidental impurities.

  16. Selection of materials for sodium fast reactor steam generators

    SciTech Connect

    Dubiez-Le Goff, S.; Garnier, S.; Gelineau, O.; Dalle, F.; Blat-Yrieix, M.; Augem, J. M.

    2012-07-01

    Sodium Fast Reactor (SFR) is considered in France as the most mature technology of the different Generation IV systems. In the short-term the designing work is focused on the identification of the potential tracks to demonstrate licensing capability, availability, in-service inspection capability and economical performance. In that frame materials selection for the major components, as the steam generator, is a particularly key point managed within a French Research and Development program launched by AREVA, CEA and EDF. The choice of the material for the steam generator is indeed complex because various aspects shall be considered like mechanical and thermal properties at high temperature, interaction with sodium on one side and water and steam on the other side, resistance to wastage, procurement, fabrication, weldability and ability for inspection and in-situ intervention. The following relevant options are evaluated: the modified 9Cr1Mo ferritic-martensitic grade and the Alloy 800 austenitic grade. The objective of this paper is to assess for both candidates their abilities to reach the current SFR needs regarding material design data, from AFCEN RCC-MRx Code in particular, compatibility with environments and manufacturability. (authors)

  17. A fuel for sub-critical fast reactor

    NASA Astrophysics Data System (ADS)

    Moiseenko, V. E.; Chernitskiy, S. V.; Ågren, O.; Noack, K.

    2012-06-01

    Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and 238U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

  18. Sodium fast reactor safety and licensing research plan. Volume I.

    SciTech Connect

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  19. One pass core design of a super fast reactor

    SciTech Connect

    Liu, Qingjie; Oka, Yoshiaki

    2013-07-01

    One pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern can significantly simplify the core concept. Upper core structure, coolant flow scheme as well as refueling procedure are as simple as in PWR. In one pass core design, supercritical-pressure water is at approximately 25.0 MPa and enters the core at 280 C. degrees and is heated up in one through flow pattern upwardly to the average outlet temperature of 500 C. degrees. Great density change in vertical direction can cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO{sub 2} fuel blanket assemblies also introduces great power increase during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at the beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of the blanket assembly. This can help to mitigate the power change in blanket due to Pu accumulation and to increase the outlet temperature of the blanket during cycle. Neutron transport and thermohydraulics coupled calculation shows that this design can satisfy the requirement in the Super FR principle for both 500 C. degrees outlet temperature and negative coolant void reactivity. (authors)

  20. A fuel for sub-critical fast reactor

    SciTech Connect

    Moiseenko, V. E.; Chernitskiy, S. V.; Agren, O.; Noack, K.

    2012-06-19

    Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and {sup 238}U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

  1. Ultrasonic decontamination of prototype fast breeder reactor fuel pins.

    PubMed

    Kumar, Aniruddha; Bhatt, R B; Behere, P G; Afzal, Mohd

    2014-04-01

    Fuel pin decontamination is the process of removing particulates of radioactive material from its exterior surface. It is an important process step in nuclear fuel fabrication. It assumes more significance with plutonium bearing fuel known to be highly radio-toxic owing to its relatively longer biological half life and shorter radiological half life. Release of even minute quantity of plutonium oxide powder in the atmosphere during its handling can cause alarming air borne activity and may pose a severe health hazard to personnel working in the vicinity. Decontamination of fuel pins post pellet loading operation is thus mandatory before they are removed from the glove box for further processing and assembly. This paper describes the setting up of ultrasonic decontamination process, installed inside a custom built fume-hood in the production line, comprising of a cleaning tank with transducers, heaters, pin handling device and water filtration system and its application in cleaning of fuel pins for prototype fast breeder reactor. The cleaning process yielded a typical decontamination efficiency of more than 99%. PMID:24405906

  2. Closing nuclear fuel cycle with fast reactors: problems and prospects

    SciTech Connect

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V.

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  3. Ceramics

    NASA Astrophysics Data System (ADS)

    Yao, Lichun; Yang, Jian; Qiu, Tai

    2014-09-01

    The effects of CuO addition on phase composition, microstructure, sintering behavior, and microwave dielectric properties of 0.80Sm(Mg0.5Ti0.5)O3-0.20 Ca0.8Sr0.2TiO3(8SMT-2CST) ceramics prepared by a conventional solid-state ceramic route have been studied. CuO addition shows no obvious influence on the phase of the 8SMT-2CST ceramics and all the samples exhibit pure perovskite structure. Appropriate CuO addition can effectively promote sintering and grain growth, and consequently improve the dielectric properties of the ceramics. The sintering temperature of the ceramics decreases by 50°C by adding 1.00 wt.%CuO. Superior microwave dielectric properties with a ɛ r of 29.8, Q × f of 85,500 GHz, and τ f of 2.4 ppm/°C are obtained for 1.00 wt.%CuO doped 8SMT-2CST ceramics sintered at 1500°C, which shows dense and uniform microstructure as well as well-developed grain growth.

  4. Design and Testing of D.C. Conduction Pump for Sodium Cooled Fast Reactor

    SciTech Connect

    Nashine, B.K.; Dash, S.K.; Gurumurthy, K.; Rajan, M.; Vaidyanathan, G.

    2006-07-01

    DC Conduction pump immersed in sodium forms a part of Failed Fuel Location Module (FFLM) of 500 MWe Fast Breeder Reactor (PFBR) currently under construction. FFLM housed in control plug of the reactor, is used to locate the failed fuel sub-assembly due to clad rupture in the fuel pin. The DC conduction pump sucks the sodium from the top of fuel sub-assemblies through the selector valve and pumps the sodium to hold up for detecting the presence of delayed neutrons. Presence of delayed neutron is the indication of failure in the sampled fuel sub-assembly. The DC Conduction Pump was chosen because of its low voltage operation (2 V) where argon/alumina ceramic can provide required electrical insulation even at operating temperature of 560 deg. C without much complication on the manufacturing front. Sampling of sodium from top of different sub-assemblies is achieved by operation of selector valve in-conjunction with the drive motor. FFLM requires the pump to be immersed in sodium pool at {approx} 560 deg. C located above the fuel sub-assemblies in the reactor. The Pump of 0.36 m{sup 3}/h capacity and developing 1.45 Kg/ cm{sup 2} pressure was designed, manufactured and tested. The DC Conduction Pump has a stainless steel duct filled with liquid sodium, which is to be pumped. The stainless steel duct is kept in magnetic field obtained by means of electromagnet. The electromagnet is made of soft iron and the coil made of copper conductor surrounds the yoke portion of electromagnet. The external DC source of 2000 Amps, 2 Volt is used to send current through sodium placed in the stainless steel duct and the same current is sent through copper coil of electromagnet for producing required magneto motive force, which in turn produces required magnetic field. The interaction of current in sodium (placed in stainless steel duct) and magnetic field produced by the electromagnet in the duct region produces pumping force in the sodium. Electromagnet, copper coil, stainless steel

  5. Fast Flux Test Facility Reactor Vessel Removal Study

    SciTech Connect

    BOWMAN, B.R.

    2002-10-23

    This study assesses the feasibility of removing the FFTF reactor vessel from its current location in the reactor cavity inside the Containment vessel to a transporter for relocation to a burial pit in the 200 Area.

  6. Rapid-L Operator-Free Fast Reactor Concept Without Any Control Rods

    SciTech Connect

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2003-07-15

    The 200-kW(electric) uranium-nitride-fueled lithium-cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for a lunar base power system. It is one of the variants of the RAPID (Refueling by All Pins Integrated Design) fast reactor concept, which enables quick and simplified refueling. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 2700 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 yr.Unique challenges in reactivity control systems design have been addressed in the RAPID-L concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt {sup 6}Li as a liquid poison instead of B{sub 4}C rods. In combination with LEMs, LIMs, and LRMs, RAPID-L can be operated without an operator. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, the RAPID-L reactor concept and its transient characteristics are presented.

  7. Method of detecting leakage of reactor core components of liquid metal cooled fast reactors

    DOEpatents

    Holt, Fred E.; Cash, Robert J.; Schenter, Robert E.

    1977-01-01

    A method of detecting the failure of a sealed non-fueled core component of a liquid-metal cooled fast reactor having an inert cover gas. A gas mixture is incorporated in the component which includes Xenon-124; under neutron irradiation, Xenon-124 is converted to radioactive Xenon-125. The cover gas is scanned by a radiation detector. The occurrence of 188 Kev gamma radiation and/or other identifying gamma radiation-energy level indicates the presence of Xenon-125 and therefore leakage of a component. Similarly, Xe-126, which transmutes to Xe-127 and Kr-84, which produces Kr-85.sup.m can be used for detection of leakage. Different components are charged with mixtures including different ratios of isotopes other than Xenon-124. On detection of the identifying radiation, the cover gas is subjected to mass spectroscopic analysis to locate the leaking component.

  8. Ceramics

    NASA Astrophysics Data System (ADS)

    Chen, Song; Zhu, De-Gui; Cai, Xu-Sheng

    2014-08-01

    The dense monoclinic-SrAl2Si2O8 ceramics have been prepared by a two-step sintering process at a sintering temperature of 1173 K (900 °C). Firstly, the pre-sintered monoclinic-SrAl2Si2O8 powders containing small SiO2·Al2O3 crystal phases were obtained by continuously sintering a powder mixture of SrCO3 and kaolin at 1223 K (950 °C) for 6 hours and 1673 K (1400 °C) for 4 hours, respectively. Subsequently, by the combination of the pre-sintered ceramic powders with the composite flux agents, which are composed of a SrO·3B2O3 flux agent and α-Al2O3, the low-temperature densification sintering of the monoclinic-SrAl2Si2O8 ceramics was accomplished at 1173 K (900 °C). The low-temperature sintering behavior and microstructure evolvement of the monoclinic-SrAl2Si2O8 ceramics have been investigated in terms of Al2O3 in addition to the composite flux agents. It shows that due to the low-meting characteristics, the SrO·3B2O3 flux agent can urge the dense microstructure formation of the monoclinic-SrAl2Si2O8 ceramics and the re-crystallization of the grains via a liquid-phase sintering. The introduction of α-Al2O3 to the SrO·3B2O3 flux agent can apparently lead to more dense microstructures for the monoclinic-SrAl2Si2O8 ceramics but also cause the re-precipitation of SiO2·Al2O3 compounds because of an excessive Al2O3 content in the SrO·3B2O3 flux agent.

  9. The Case Against the Fast Breeder Reactor: An Anti-Nuclear Establishment View.

    ERIC Educational Resources Information Center

    Lovins, Amory B.

    1973-01-01

    Environmentalists lobby points out that hazards which may result from mistakes in proposed fast breeder reactor for additional energy can be detrimental for mankind. Such projects must be carefully planned and cautiously executed. (PS)

  10. CORAL: a stepping stone for establishing the Indian fast reactor fuel reprocessing technology

    SciTech Connect

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

    2007-07-01

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

  11. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    SciTech Connect

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  12. A FAST BREEDER REACTOR SPENT FUEL MEASUREMENTS PROGRAM FOR BN-350 REACTOR

    SciTech Connect

    P. STAPLES; J. HALBIG; ET AL

    1999-04-01

    A project to verify the fissile content of fast breeder reactor spent nuclear fuel is underway in the Republic of Kasakhstan. There are a variety of assembly types with different irradiation histories and profiles in the reactor that require a variety of measurement and analysis procedures. These procedures will be discussed and compared as will the general process that has been designed to resolve any potential measurement discrepancies. The underwater counter is part of a system that is designed to assist the International Atomic Energy Agency (IAEA) in maintaining continuity of knowledge from the time of measurement until the measured item is placed in a welded container with a unique identification. In addition to satisfying IAEA requirements for the spent nuclear fuel, this measurement program is able to satisfy some of the measurement requirements for the Kasakhstan Atomic Energy Agency concerning the repackaging of the spent nuclear fuel into a standard canister. The project is currently operational in a mode requiring the IAEA's continuous presence.

  13. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  14. Method of locating a leaking fuel element in a fast breeder power reactor

    DOEpatents

    Honekamp, John R.; Fryer, Richard M.

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  15. Ceramics

    NASA Astrophysics Data System (ADS)

    Bin, Tang; Feng, Si; Ying-xiang, Li; He-tuo, Chen; Xiao, Zhang; Shu-ren, Zhang

    2014-11-01

    The effects of Ta2O5/Y2O3 codoping on the microstructure and microwave dielectric properties of Ba(Co0.56Zn0.40)1/3Nb2/3O3- xA- xB (A = 0.045 wt.% Ta2O5; B = 0.113 wt.% Y2O3) ceramics ( x = 0, 1, 2, 4, 8, 16, 32) prepared according to the conventional solid-state reaction technique were investigated. The x-ray diffraction (XRD) results showed that the main crystal phase in the sintered ceramics was BaZn0.33Nb0.67O3-Ba3CoNb2O9. The additional surface phase of Ba8CoNb6O24 and trace amounts of Ba5Nb4O15 second phase were present when Ta2O5/Y2O3 was added to the ceramics. The 1:2 B-site cation ordering was affected by the substitution of Ta5+ and Y3+ in the crystal lattice, especially for x = 4. Scanning electron microscopy (SEM) images of the optimally doped ceramics sintered at 1340°C for 20 h showed a compact microstructure with crystal grains in dense contact. Though the dielectric constant increased with the x value, appropriate addition would result in a tremendous modification of the Q × f and τ f values. Excellent microwave dielectric properties ( ɛ r = 35.4, Q × f = 62,993 GHz, and τ f = 2.6 ppm/°C) were obtained for the ceramic with x = 0.4 sintered in air at 1340°C for 20 h.

  16. Safety and core design of large liquid-metal cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  17. The Fast-Flow Discharge Reactor as an Undergraduate Instructional Tool.

    ERIC Educational Resources Information Center

    Provencher, G. M.

    1981-01-01

    A fast-flow discharge reactor has been used in an analytical chemistry demonstration of gas phase titration, in inorganic preparative chemistry, and in physical chemistry as a "practice" vacuum line, kinetic reactor, and spectroscopic source as well as an undergraduate research tool. (SK)

  18. SACRD: a data base for fast reactor safety computer codes, operational procedures

    SciTech Connect

    Forsberg, V.M.; Arwood, J.W.; Greene, N.M.; Raiford, G.B.

    1980-09-01

    SACRD (Safety Analysis Computerized Reactor Data) is a data base of nondesign-related information used in computer codes for fast reactor safety analyses. This document reports the procedures used in SACRD to help assure a reasonable level of integrity of the material contained in the data base. It also serves to document much of the computer software used with the data base.

  19. Influence of reactor radiation on {alpha}-Al{sub 2}O{sub 3} in electrically insulating ceramic

    SciTech Connect

    Astapova, E.S.; Kostyukov, N.S.

    1995-11-01

    Under the action of radiation, structural changes occur in ceramic materials; these changes influence its mechanical properties and corrosion resistance. Neutron resistance has the strongest effect, inducing a number of complex processes. Most ceramic materials consist of one or more crystalline phases cemented by glass phases. Neutron radiation produces opposite density changes in the crystalline and glass phase. The conflict between these processes in the ceramic leads to increase in radiation resistance, which is the essence of the compensation effect. Thus, on irradiation, the density of crystalline quartz in the free state decreases by 15%, while the density of quartz glass increase by 3%, with corresponding changes in the volume of the phases. In porcelain, such changes facilitate an increase in strength. The radiational strength of ceramic materials was investigated - in particular, the structural changes in the irradiation of the ceramic by fast neutrons in a flux of no more than 2{center_dot}10{sup 20} cm{sup -2}. The main effects noticed after irradiation of the ceramic by fast distance and decrease in intensity of the diffractional maxima in the crystalline phases of the ceramic, for example, in the {alpha}-Al{sub 2}O{sub 3} phase of the ceramics microlite, GB-7, ultraporcelain, and 22KhS. In the initial state, GB-7 ceramic has a homogeneous, analogous, but the {alpha}-Al{sub 2}O{sub 3} crystals are smaller. According to chemical analysis, the mass fraction of aluminum oxide in GB-7 and 22KhS is 97.09% and 95.14%, respectively.

  20. 70 Group Neutron Fast Reactor Cross Section Set Based on JENDL-2B.

    Energy Science and Technology Software Center (ESTSC)

    1984-02-06

    Version 00 These multigroup cross sections are used in fast reactor calculations. The benchmark calculations for the 23 fast critical assemblies used in the benchmark tests of JFS-2 were performed with one-dimensional diffusion theory by using the JFS-3-J2 set.

  1. Burnup concept for a long-life fast reactor core using MCNPX.

    SciTech Connect

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  2. Ceramics

    NASA Astrophysics Data System (ADS)

    Li, Enzhu; Zou, Mengying; Duan, Shuxin; Xu, Ning; Yuan, Ying; Zhou, Xiaohua

    2014-11-01

    The effects of excess Li content on the phase structure and microwave dielectric properties, especially on the temperature coefficient, of LiNb0.6 Ti0.5O3 (LNT) ceramics were studied. The results show that small amounts of Li effectively enhanced the sintering process due to the compensation of high volatility of Li, leading to a densification and homogenous microstructure, and therefore enhanced the dielectric properties. However, too much Li leads to a secondary phase and cause abnormal grain growth. The LNT + 5 wt.% Li ceramic sintered at 1075°C in the air shows the best properties of ɛ r = 69.73, Q × f = 5543 GHz, and τ f = -4.4 ppm/°C.

  3. Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

    SciTech Connect

    Takeda, T.; Shimazu, Y.; Hibi, K.; Fujimura, K.

    2012-07-01

    Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of this project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)

  4. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect

    Samuel Bays; Ayodeji Alajo

    2010-05-01

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  5. Emergency cooling down of fast-neutron reactors by natural convection (a review)

    NASA Astrophysics Data System (ADS)

    Zhukov, A. V.; Sorokin, A. P.; Kuzina, Yu. A.

    2013-05-01

    Various methods for emergency cooling down of fast-neutron reactors by natural convection are discussed. The effectiveness of using natural convection for these purposes is demonstrated. The operating principles of different passive decay heat removal systems intended for cooling down a reactor are explained. Experimental investigations carried out in Russia for substantiating the removal of heat in cooling down fast-neutron reactors are described. These investigations include experimental works on studying thermal hydraulics in small-scale simulation facilities containing the characteristic components of a reactor (reactor core elements, above-core structure, immersed and intermediate heat exchangers, pumps, etc.). It is pointed out that a system that uses leaks of coolant between fuel assemblies holds promise for fast-neutron reactor cooldown purposes. Foreign investigations on this problem area are considered with making special emphasis on the RAMONA and NEPTUN water models. A conclusion is drawn about the possibility of using natural convection as the main method for passively removing heat in cooling down fast-neutron reactors, which is confirmed experimentally both in Russia and abroad.

  6. Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

    NASA Astrophysics Data System (ADS)

    Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

    2001-02-01

    Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

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

    SciTech Connect

    Mann, A.; Herrick, R.; Gunn, J.; Husband, W.; Smith, M.; Fletcher, B.

    2007-07-01

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

  8. The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

    SciTech Connect

    John D. Bess; Michael A. Pope; Harold F. McFarlane

    2012-11-01

    The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactor (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.

  9. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    NASA Astrophysics Data System (ADS)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  10. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect

    Chad Pope

    2007-05-01

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day

  11. A Comparison of Fast-Spectrum and Moderated Space Fission Reactors

    SciTech Connect

    Poston, David I.

    2005-02-06

    The reactor neutron spectrum is one of the fundamental design choices for any fission reactor, but the implications of using a moderated spectrum are vastly different for space reactors as opposed to terrestrial reactors. In addition, the pros and cons of neutron spectra are significantly different among many of the envisioned space power applications. This paper begins with a discussion of the neutronic differences between fast-spectrum and moderated space reactors. This is followed by a discussion of the pros and cons of fast-spectrum and moderated space reactors separated into three areas--technical risk, performance, and safety/safeguards. A mix of quantitative and qualitative arguments is presented, and some conclusions generally can be made regarding neutron spectrum and space power application. In most cases, a fast-spectrum system appears to be the better alternative (mostly because of simplicity and higher potential operating temperatures); however, in some cases, such as a low-power (<100-kWt) surface reactor, a moderated spectrum could provide a better approach. In all cases, the determination of which spectrum is preferred is a strong function of the metrics provided by the 'customer' - i.e., if a certain level of performance is required, it could provide a different solution than if a certain level of safeguards is required (which in some cases could produce a null solution). The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

  12. Yttrium and rare earth stabilized fast reactor metal fuel

    DOEpatents

    Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

    1992-01-01

    To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

  13. A revaluation of helium/dpa ratios for fast reactor and thermal reactor data in fission-fusion correlations

    SciTech Connect

    Garner, F.A.; Greenwood, L.R.; Oliver, B.M.

    1996-10-01

    For many years it has been accepted that significant differences exist in the helium/dpa ratios produced in fast reactors and various proposed fusion energy devices. In general, the differences arise from the much larger rate of (n,{alpha}) threshold reactions occurring in fusion devices, reactions which occur for energies {ge} 6 MeV. It now appears, however, that for nickel-containing alloys in fast reactors the difference may not have been as large as was originally anticipated. In stainless steels that have a very long incubation period for swelling, for instance, the average helium concentration over the duration of the transient regime have been demonstrated in an earlier paper to be much larger in the FFTF out-of-core regions than first calculated. The helium/dpa ratios in some experiments conducted near the core edge or just outside of the FFTF core actually increase strongly throughout the irradiation, as {sup 59}Ni slowly forms by transmutation of {sup 58}Ni. This highly exothermic {sup 59}Ni(n,{alpha}) reaction occurs in all fast reactors, but is stronger in the softer spectra of oxide-fueled cores such as FFTF and weaker in the harder spectra of metal-fueled cores such as EBR-II. The formation of {sup 59}Ni also increases strongly in out-of-core unfueled regions where the reactor spectra softens with distance from the core.

  14. Historical Review of U.S. Transient Fast Reactor Fuel Testing

    SciTech Connect

    Carmack, William J.; MacLean, Heather J.; Crawford, Douglas C.

    2007-07-01

    Development of fast spectrum nuclear fuels in the United States has been pursued over the course of approximately 30 years including the EBR-I and FERMI reactors and continuing through the early 1990's culminating with the FFTF and the EBR-II Integral Fast Reactor programs. These programs primarily focused on oxide and metallic fuels and the development process provided sufficient evidence for licensing of the 20%Pu-MOX oxide fuel and the ternary U-Pu-Zr metallic alloy. The development of a transuranic, actinide burning fuel system will require significant development including the investigation and testing of the behavior of candidate fuel systems under transient conditions. This paper will review the historical status of both metallic and oxide fuel transient testing completed under previous U.S. fast reactor fuel development programs. (authors)

  15. Heat insulating system for a fast reactor shield slab

    DOEpatents

    Kotora Jr., James; Groh, Edward F.; Kann, William J.; Burelbach, James P.

    1986-04-01

    Improved thermal insulation for a nuclear reactor deck comprising many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

  16. Heat insulating system for a fast reactor shield slab

    DOEpatents

    Kotora, J. Jr.; Groh, E.F.; Kann, W.J.; Burelbach, J.P.

    1984-04-10

    Improved thermal insulation for a nuclear reactor deck comprises many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

  17. Yttrium and rare earth stabilized fast reactor metal fuel

    SciTech Connect

    Guon, J.; Grantham, L.F.; Specht, E.R.

    1992-05-12

    This patent describes an improved metal alloy reactor fuel consisting essentially of uranium, plutonium, and at least one element from the group consisting of yttrium, lanthanum, cerium, praseodymium, neodymium, promethium, samarium, europium, gadolinium, terbium, dysprosium, holmium, erbium, thulium, ytterbium and lutetium.

  18. Heat insulating system for a fast reactor shield slab

    DOEpatents

    Kotora, Jr., James; Groh, Edward F.; Kann, William J.; Burelbach, James P.

    1986-01-01

    Improved thermal insulation for a nuclear reactor deck comprising many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

  19. Fast wave current drive: Experimental status and reactor prospects

    SciTech Connect

    Ehst, D.A.

    1988-03-01

    The fast wave is one of the two possible wave polarizations which propagate according to the basic theory of cold plasmas. It is distinguished from the other (slow wave) branch by having an electric field vector which is mainly orthogonal to the confining magnetic field of the plasma. The plasma and fast wave qualitatively assume different behavior depending on the frequency range of the launched wave. The high frequency fast wave (HFFW), with a frequency (..omega..2..pi.. )approximately) GHz) much higher than the ion cyclotron frequency (..cap omega../sub i/), suffers electron Landau damping and drives current by supplying parallel momentum to superthermal electrons in a fashion similar to lower hybrid (slow wave) current drive. In the simple theory the HFFW should be superior to the slow wave and can propagate to very high density and temperature without impediment. Experiments, however, have not conclusively shown that HFFW current drive can be achieved at densities above the slow wave current drive limit, possibly due to conversion of the launched fast waves into slow waves by density fluctuations. Alternatively, the low frequency fast wave (LFFW), with frequencies ()approxreverse arrowlt) 100 MHz) only a few times the ion cyclotron frequency, is damped by electron Landau damping and, in a hot plasma ()approxreverse arrowgt) 10 keV), by electron transit time magnetic pumping; current drive is achieved by pushing superthermal electrons, and efficiency is prediocted to be slightly better than for lower hybrid current drive. Most significantly, the slow wave does not propagate in high density plasma when ..omega.. )approximately) ..cap omega../sub i/, so parasitic coupling to the slow wave can be avoided, and no density and temperture limitations are foreseen. Experiments with fast wve current drive invariably find current drive efficiency as good as obtained in lower hybrid experiments at comparable, low temperatures. 45 refs., 4 figs., 1 tab

  20. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    NASA Astrophysics Data System (ADS)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  1. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    SciTech Connect

    Haihua Zhao; Hongbin Zhang

    2007-11-01

    The existing sodium cooled fast reactors (SFR) have two types of designs – loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphénix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL’s Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed.

  2. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    SciTech Connect

    Widiawati, Nina Su’ud, Zaki

    2015-09-30

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  3. Modeling the behavior of metallic fast reactor fuels during extended transients

    NASA Astrophysics Data System (ADS)

    Kramer, J. M.; Liu, Y. Y.; Billone, M. C.; Tsai, H. C.

    1993-09-01

    Passive safety features in metal-fueled reactors utilizing the Integral Fast Reactor (IFR) fuel system make it possible to avoid core damage for extended time periods even when automatic scram systems fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this intermediate time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements.

  4. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    SciTech Connect

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant.

  5. Under-sodium viewing technology for improvement of fast-reactor safeguards

    SciTech Connect

    Beddingfield, David H; Gerhart, Jeremy J; Kawakubo, Yoko

    2009-01-01

    The current safeguards approach for fast reactors relies exclusively on maintenance of continuity of knowledge to track the movement of fuel assemblies through these facilities. The remote handling of fuel assemblies, the visual opacity of the liquid metal coolant. and the chemical reactivity of sodium all combine and result in significant limitations on the available options to verify fuel assembly identification numbers or the integrity of these assemblies. These limitations also serve to frustrate attempts to restore the continuity-of-knowledge in instances where the information is under a variety of scenarios. The technology of ultrasonic under-sodium viewing offers new options to the safeguards community for recovering continuity-of-knowledge and applying more traditional item accountancy to fast reactor facilities. We have performed a literature review to investigate the development of under-sodium viewing technologies. In this paper we will summarize our findings and report the state of development of this technology and we will present possible applications to the fast reactor system to improve the existing safeguards approach at these reactors and in future fast reactors.

  6. Behavior of 241Am in fast reactor systems - a safeguards perspective

    SciTech Connect

    Beddingfield, David H; Lafleur, Adrienne M

    2009-01-01

    Advanced fuel-cycle developments around the world currently under development are exploring the possibility of disposing of {sup 241}Am from spent fuel recycle processes by burning this material in fast reactors. For safeguards practitioners, this approach could potentially complicate both fresh- and spent-fuel safeguards measurements. The increased ({alpha},n) production in oxide fuels from the {sup 241}Am increases the uncertainty in coincidence assay of Pu in MOX assemblies and will require additional information to make use of totals-based neutron assay of these assemblies. We have studied the behavior of {sup 241}Am-bearing MOX fuel in the fast reactor system and the effect on neutron and gamma-ray source-terms for safeguards measurements. In this paper, we will present the results of simulations of the behavior of {sup 241}Am in a fast breeder reactor system. Because of the increased use of MOX fuel in thermal reactors and advances in fuel-cycle designs aimed at americium disposal in fast reactors, we have undertaken a brief study of the behavior of americium in these systems to better understand the safeguards impacts of these new approaches. In this paper we will examine the behavior of {sup 241}Am in a variety of nuclear systems to provide insight into the safeguards implications of proposed Am disposition schemes.

  7. Design Study of Small Lead-Cooled Fast Reactors Using SiC Cladding and Structure

    SciTech Connect

    Abu Khalid Rivai; Minoru Takahashi

    2006-07-01

    Effects of SiC cladding and structure on neutronics of reactor core for small lead-cooled fast reactors have been investigated analytically. The fuel of this reactor was uranium nitride with {sup 235}U enrichment of 11% in inner core and 13% in outer core. The reactors were designed by optimizing the use of natural uranium blanket and nitride fuel to prolong the fuel cycle. The fuels can be used without re-shuffling for 15 years. The coolant of this reactor was lead. A calculation was also conducted for steel cladding and structure type as comparison with SiC cladding and structure type. The results of calculation indicated that the neutron energy spectrum of the core using SiC was slightly softer than that using steel. The SiC type reactor was designed to have criticality at the beginning of cycle (BOC), although the steel type reactor could not have critical condition with the same size and geometry. In other words, the SiC type core can be designed smaller than the steel type core. The result of the design analysis showed that neutron flux distributions and power distribution was made flatter because the outer core enrichment was higher than inner core. The peak power densities could remain constant over the reactor operation. The consumption capability of uranium was quite high, i.e. 13% for 125 MWt reactor and 25% for 375 MWt reactor at EOC. (authors)

  8. The performance of ENDF/B-V. 2 nuclear data for fast reactor calculations

    SciTech Connect

    Atkinson, C.A.; Collins, P.J.

    1987-01-01

    Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations.

  9. 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.

  10. High Temperature Ultrasonic Transducers for In-Service Inspection of Liquid Metal Fast Reactors

    SciTech Connect

    Griffin, Jeffrey W.; Posakony, Gerald J.; Harris, Robert V.; Baldwin, David L.; Jones, Anthony M.; Bond, Leonard J.

    2011-12-31

    In-service inspection of liquid metal (sodium) fast reactors requires the use of ultrasonic transducers capable of operating at high temperatures (>200°C), high gamma radiation fields, and the chemically reactive liquid sodium environment. In the early- to mid-1970s, the U.S. Atomic Energy Commission supported development of high-temperature, submersible single-element transducers, used for scanning and under-sodium imaging in the Fast Flux Test Facility and the Clinch River Breeder Reactor. Current work is building on this technology to develop the next generation of high-temperature linear ultrasonic transducer arrays for under-sodium viewing and in-service inspections.

  11. Fuel damage during off-normal transients in metal-fueled fast reactors

    SciTech Connect

    Kramer, J.M.; Bauer, T.H.

    1990-01-01

    Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs.

  12. Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report

    SciTech Connect

    Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

    2006-09-01

    Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature

  13. Example Work Domain Analysis for a Reference Sodium Fast Reactor

    SciTech Connect

    Hugo, Jacques; Oxstrand, Johanna

    2015-01-01

    The nuclear industry is currently designing and building a new generation of reactors that will include different structural, functional, and environmental aspects, all of which are likely to have a significant impact on the way these plants are operated. In order to meet economic and safety objectives, these new reactors will all use advanced technologies to some extent, including new materials and advanced digital instrumentation and control systems. New technologies will affect not only operational strategies, but will also require a new approach to how functions are allocated to humans or machines to ensure optimal performance. Uncertainty about the effect of large scale changes in plant design will remain until sound technical bases are developed for new operational concepts and strategies. Up-to-date models and guidance are required for the development of operational concepts for complex socio-technical systems. This report describes how the classical Work Domain Analysis method was adapted to develop operational concept frameworks for new plants. This adaptation of the method is better able to deal with the uncertainty and incomplete information typical of first-of-a-kind designs. Practical examples are provided of the systematic application of the method in the operational analysis of sodium-cooled reactors. Insights from this application and its utility are reviewed and arguments for the formal adoption of Work Domain Analysis as a value-added part of the Systems Engineering process are presented.

  14. Fast reactor power plant design having heat pipe heat exchanger

    DOEpatents

    Huebotter, P.R.; McLennan, G.A.

    1984-08-30

    The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

  15. Fast reactor power plant design having heat pipe heat exchanger

    DOEpatents

    Huebotter, Paul R.; McLennan, George A.

    1985-01-01

    The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

  16. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    SciTech Connect

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  17. Lead-cooled fast reactor use in future equilibrium energy production

    SciTech Connect

    Sekimoto, Hiroshi; Kuznetsov, V.V.

    1994-12-31

    The design of a lead cooled fast reactor is discussed. In previous works, general characteristics of future nuclear equilibrium energy utilization have been investigated where the toxic radioactive materials are confined in a nuclear center. Natural uranium and/or thorium, are supplied to the center as a fuel fed to fission reactors. All of the actinides are recycled in the reactor. The end products of the heavy-isotope decay series (lead and bismuth) and the stable fission products are taken out of the center. The discharged short- and middle-life fission products are cooled until they die out in the center, then the previously described process is used.

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

    SciTech Connect

    Gregg, R.; Hesketh, K.

    2013-07-01

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

  19. Use of Multi-Purpose Modular Fast Reactors SVBR-75/100 in Market Conditions

    SciTech Connect

    Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G.; Dragunov, Yu.G.; Stepanov, V.S.; Klimov, N.N.; Kopytov, I.I.; Krushelnitsky, V.N.

    2006-07-01

    This report presents an innovative nuclear power technology (NPT), based on the use of modular type fast-neutron reactors (FR) (SVBR-75/100) having heavy liquid-metal coolant (HLMC) i.e. eutectic lead-bismuth alloy, which was mastered in Russia for the nuclear submarine (NS) reactors. Reactor SVBR-75/100 possesses inherent self-protection and passive safety properties that allows elimination of many safety systems necessary for traditional type reactors. Use of this NPT makes it possible to eliminate conflicting requirements among safety needs and economic factors, which is particularly found in traditional reactors, to increase considerably the investment attractiveness of nuclear power (NP) based on the use of fast-neutron reactors for the near future, when the cost of natural uranium is low and to assure NP development in the market conditions. On the basis of the factory-fabricated 'standard' reactor modules, it is possible to construct the nuclear power plants (NPP) with different power and purposes. (authors)

  20. High temperature ceramic membrane reactors for coal liquid upgrading. Quarterly report No. 1, September 21, 1989--December 20, 1989

    SciTech Connect

    Tsotsis, T.T.

    1992-06-19

    In this project we well evaluate the performance of Sel-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated. (VC)

  1. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    SciTech Connect

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  2. High temperature ceramic membrane reactors for coal liquid upgrading. Quarter report No. 9, September 21, 1991--December 20, 1991

    SciTech Connect

    Tsotsis, T.T.

    1992-07-01

    In this project we intend to study a novel process concept, i.e, the use of ceramic membranes reactors in upgrading of coal derived liquids. Membrane reactors have been used in a number of catalytic reaction processes in order to overcome the limitations on conversion imposed by thermodynamic equilibrium. They have, furthermore, the inherent capability for combining reaction and separation in a single step. Thus they offer promise for improving and optimizing yield, selectivity and performance of processes involving complex liquids, as those typically found in coal liquid upgrading. Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we will evaluate the performance of Sol-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

  3. High temperature ceramic membrane reactors for coal liquid upgrading. Quarterly report No. 12, June 21, 1992--September 20, 1992

    SciTech Connect

    Tsotsis, T.T.

    1992-12-31

    In this project we intend to study a novel process concept, i.e.,the use of ceramic membranes reactors in upgrading of coal derived liquids. Membrane reactors have been used in a number of catalytic reaction processes in order to overcome the limitations on conversion imposed by thermodynamic equilibrium. They have, furthermore, the inherent capability for combining reaction and separation in a single step. Thus they offer promise for improving and optimizing yield, selectivity and performance of processes involving complex liquids, as those typically found in coal liquid upgrading. Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we wig evaluate the performance of Sel-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

  4. Uncertainty Assessment for Fast Reactors Based on Nuclear Data Adjustment

    NASA Astrophysics Data System (ADS)

    Ivanova, T.; Ivanov, E.; Ecrabet, F.

    2014-04-01

    The paper presents IRSN's results of the OECD/NEA WPEC Subgroup 33 benchmark exercise which is focused upon combined use of differential and integral data using adjustment technique. The results are generated by BERING code using different sets of input data: integral parameters and sensitivity coefficients for fast benchmark experiments and applications computed by deterministic ERANOS code and Monte Carlo SCALE sequences, COMMARA-2.0 and JENDL-4.0 cross-section-covariance data and integral correlations provided by JAEA. The paper demonstrates results of the adjustment when using different input data and two adjustment algorithms implemented in BERING.

  5. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

    ERIC Educational Resources Information Center

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

  6. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion.

    PubMed

    Marques, J G; Sousa, M; Santos, J P; Fernandes, A C

    2011-08-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard. PMID:21071234

  7. Catalytic fast pyrolysis of white oak wood in-situ using a bubbling fluidized bed reactor

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Catalytic fast pyrolysis was performed on white oak wood using two zeolite-type catalysts as bed material in a bubbling fluidized bed reactor. The two catalysts chosen, based on a previous screening study, were Ca2+ exchanged Y54 (Ca-Y54) and a proprietary ß-zeolite type catalyst (catalyst M) both ...

  8. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    SciTech Connect

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  9. Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

    SciTech Connect

    Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.; Johnson, Gerald D.; Brown, W. F.; Paxton, Michael M.; Puigh, Raymond J.; Eiholzer, Cheryl R.; Martinez, C.; Blotter, M. A.

    2000-02-28

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

  10. Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

    DOEpatents

    Brehm, Jr., William F.; Colburn, Richard P.

    1982-01-01

    An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

  11. A Numerical Model for Coupling of Neutron Diffusion and Thermomechanics in Fast Burst Reactors

    SciTech Connect

    Samet Y. Kadioglu; Dana A. Knoll; Cassiano De Oliveira

    2008-11-01

    We develop a numerical model for coupling of neutron diffusion adn termomechanics in order to stimulate transient behavior of a fast burst reactor. The problem involves solving a set of non-linear different equations which approximate neutron diffusion, temperature change, and material behavior. With this equation set we will model the transition from a supercritical to subcritical state and possible mechanical vibration.

  12. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    NASA Astrophysics Data System (ADS)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  13. Fuel supply of nuclear power industry with the introduction of fast reactors

    NASA Astrophysics Data System (ADS)

    Muraviev, E. V.

    2014-12-01

    The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

  14. Neutronic Assessment of Transmutation Target Compositions in Heterogeneous Sodium Fast Reactor Geometries

    SciTech Connect

    Samuel E. Bays; Rodolfo M. Ferrer; Michael A. Pope; Benoit Forget; Mehdi Asgari

    2008-02-01

    The sodium fast reactor is under consideration for consuming the transuranic waste in the spent nuclear fuel generated by light water reactors. This work is concerned with specialized target assemblies for an oxide-fueled sodium fast reactor that are designed exclusively for burning the americium and higher mass actinide component of light water reactor spent nuclear fuel (SNF). The associated gamma and neutron radioactivity, as well as thermal heat, associated with decay of these actinides may significantly complicate fuel handling and fabrication of recycled fast reactor fuel. The objective of using targets is to isolate in a smaller number of assemblies these concentrations of higher actinides, thus reducing the volume of fuel having more rigorous handling requirements or a more complicated fabrication process. This is in contrast to homogeneous recycle where all recycled actinides are distributed among all fuel assemblies. Several heterogeneous core geometries were evaluated to determine the fewest target assemblies required to burn these actinides without violating a set of established fuel performance criteria. The DIF3D/REBUS code from Argonne National Laboratory was used to perform the core physics and accompanying fuel cycle calculations in support of this work. Using the REBUS code, each core design was evaluated at the equilibrium cycle condition.

  15. Performance Comparison of Metallic, Actinide Burning Fuel in Lead-Bismuth and Sodium Cooled Fast Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-04-01

    Various methods have been proposed to “incinerate” or “transmutate” the current inventory of trans-uranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non-fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years.

  16. Computational fluid dynamics modelling of biomass fast pyrolysis in fluidised bed reactors, focusing different kinetic schemes.

    PubMed

    Ranganathan, Panneerselvam; Gu, Sai

    2016-08-01

    The present work concerns with CFD modelling of biomass fast pyrolysis in a fluidised bed reactor. Initially, a study was conducted to understand the hydrodynamics of the fluidised bed reactor by investigating the particle density and size, and gas velocity effect. With the basic understanding of hydrodynamics, the study was further extended to investigate the different kinetic schemes for biomass fast pyrolysis process. The Eulerian-Eulerian approach was used to model the complex multiphase flows in the reactor. The yield of the products from the simulation was compared with the experimental data. A good comparison was obtained between the literature results and CFD simulation. It is also found that CFD prediction with the advanced kinetic scheme is better when compared to other schemes. With the confidence obtained from the CFD models, a parametric study was carried out to study the effect of biomass particle type and size and temperature on the yield of the products. PMID:26927234

  17. Prediction of the thermophysical properties of molten salt fast reactor fuel from first-principles

    NASA Astrophysics Data System (ADS)

    Gheribi, A. E.; Corradini, D.; Dewan, L.; Chartrand, P.; Simon, C.; Madden, P. A.; Salanne, M.

    2014-05-01

    Molten fluorides are known to show favourable thermophysical properties which make them good candidate coolants for nuclear fission reactors. Here we investigate the special case of mixtures of lithium fluoride and thorium fluoride, which act both as coolant and as fuel in the molten salt fast reactor concept. By using ab initio parameterised polarisable force fields, we show that it is possible to calculate the whole set of properties (density, thermal expansion, heat capacity, viscosity and thermal conductivity) which are necessary for assessing the heat transfer performance of the melt over the whole range of compositions and temperatures. We then deduce from our calculations several figures of merit which are important in helping the optimisation of the design of molten salt fast reactors.

  18. GIF sodium fast reactor project R and D on safety and operation

    SciTech Connect

    Vasile, A.; Sofu, T.; Jeong, H. Y.; Sakai, T.

    2012-07-01

    The 'Safety and Operation' project is started in 2009 within the framework of Generation-IV International Forum (GIF) Sodium Fast Reactor (SFR) research and development program. In the safety area, the project involves R and D activities on phenomenological model development and experimental programs, conceptual studies in support of the design of safety provisions, preliminary assessment of safety systems, framework and methods for analysis of safety architecture. In the operation area, the project involves R and D activities on fast reactors safety tests and analysis of reactor operations, feedback from decommissioning, in-service inspection technique development, under-sodium viewing and sodium chemistry. This paper presents a summary of such activities and the main achievements. (authors)

  19. CABRI Reactor: The fast neutron Hodoscope Calibration curves calculation with MORET

    NASA Astrophysics Data System (ADS)

    Bernard, Franck; Chevalier, Vincent; Venanzi, Damiano

    2014-06-01

    This poster presents the Hodoscope calibration curves calculation with 3D Monte Carlo code MORET. The fast neutron hodoscope is a facility of the CABRI research reactor at Cadarache (FRANCE). This hodoscope is designed to measure the fuel motion during a RIA in a pressurized water reactor. The fuel motion is measured by counting fast fission neutrons emerging from the test fuel placed in an experimental loop functioning like a Pressurized Water Reactor (T=300°C and P=155 bar), at the center of the CABRI core. The detection system of the hodoscope measures a signal which is a function of the fuel motion. The calibration curves allow then to convert the signal in a fuel mass. In order to calculate these curves, we have developed a method based on a Monte Carlo calculation code.

  20. Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

    SciTech Connect

    Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.; Attenberger, S.E.

    1987-11-01

    The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor (ETR) plasma (Tokamak Ignition/Burn Experimental Reactor (TIBER-II)) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-D transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alpha concentration significantly influence the ignition and steady-state burn capability. 23 refs., 9 figs., 4 tabs.

  1. Time-resolved and time-integrated radiography of fast reactor fuel elements

    SciTech Connect

    De Volpi, A.

    1981-01-01

    The fast-reactor safety program has some unusual requirements in radiography. Applications may be divided into two areas: time-resolved or time-integrated radiography. The fast-neutron hodoscope has supplied all recent time-resolved cineradiographic in-pile fuel-motion data, and various x-ray and photographic techniques have been used for out-of-pile experiments. Thick containers and the large number of radioactive fuel pins involved in safety research have been responsible for some nonconventional applications of time-integrated radiography of stationary objects. Hodoscopes record fuel-motion during transient experiments at the TREAT reactor in the United States and CABRI in France. Other special techniques have been under development for out-of-pile nondestructive radiography of fuel element subassemblies, including fast-neutron and gamma-ray tomographic methods.

  2. Ceramics for fusion reactors: The role of the lithium orthosilicate as breeder

    NASA Astrophysics Data System (ADS)

    Carella, Elisabetta; Hernández, Teresa

    2012-11-01

    Lithium-based oxide ceramics are studied as breeder blanket materials for the controlled thermonuclear reactors (CTR). Lithium orthosilicate (Li4SiO4) is one of the most promising candidates because of its lithium concentration (0.54 g/cm3), its high melting temperature (1523 K) and its excellent tritium release behavior. It is reported that the diffusion of tritium is closely related to that of lithium, so it is possible to find an indirect measure of the trend of tritium studying the diffusivity of Li+. In the present work, the synthesis of the Li4SiO4 is carried out by Spray drying followed by pyrolysis. The study of the Li+ ion diffusion on the sintered bodies, is investigated by means of electrical conductivity measurements. The effect of the γ-ray irradiation is evaluated by the impedance spectroscopy method (EIS) from room temperature to 1173 K. The results indicate that the síntesis process employed can produce Li4SiO4 in the form of pebbles, finally the best ion species for the electrical conduction is the Li+ and is shown that the g-irradiation to a dose of 5MGy, facilitate its mobility through the creation of defects, without change in its conduction process.

  3. U. S. fast reactor materials and structures program

    SciTech Connect

    Harms, W.O.; Purdy, C.M.

    1984-01-01

    The U.S. DOE has sponsored a vigorous breeder reactor materials and structures program for 15 years. Important contributions have resulted from this effort in the areas of design (inelastic rules, verified methods, seismic criteria, mechanical properties data); resolution of licensing issues (technical witnessing, confirmatory testing); construction (fabrication/welding procedures, nondestructive testing techniques); and operation (sodium purification, instrumentation and chemical analysis, radioactivity control, and in-service inspection. The national LMFBR program currently is being restructured. The Materials and Structures Program will focus its efforts in the following areas: (1) removal of anticipated licensing impediments through confirmation of the adequacy of structural design methods and criteria for components containing welds and geometric discontinuities, the generation of mechanical properties for stainless steel castings and weldments, and the evaluation of irradiation effects; (2) qualification of modified 9 Cr-1 Mo steel and tribological coatings for design flexibility; (3) development of improved inelastic design guidelines and procedures; (4) reform of design codes and standards and engineering practices, leading to simpler, less conservative rules and to simplified design analysis methods; and (5) incorporation of information from foreign program.

  4. Fast neutron fluence of yonggwang nuclear unit 1 reactor pressure vessel

    SciTech Connect

    Yoo, C.; Km, B.; Chang, K.; Leeand, S.; Park, J.

    2006-07-01

    The Code of Federal Regulations, Title 10, Part 50, Appendix H, requires that the neutron dosimetry be present to monitor the reactor vessel throughout plant life. The Ex-Vessel Neutron Dosimetry System has been installed for Yonggwang Nuclear Unit 1 after complete withdrawal of all six in-vessel surveillance capsules. This system has been installed in the reactor cavity annulus in order to measure the fast neutron spectrum coming out through the reactor pressure vessel. Cycle specific neutron transport calculations were performed to obtain the energy dependent neutron flux throughout the reactor geometry including dosimetry positions. Comparisons between calculations and measurements were performed for the reaction rates of each dosimetry sensors and results show good agreements. (authors)

  5. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    SciTech Connect

    Benedict, R.W.; Goff, K.M.

    1993-01-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  6. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    SciTech Connect

    Benedict, R.W.; Goff, K.M.

    1993-03-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  7. A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Travel

    NASA Technical Reports Server (NTRS)

    Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

    1998-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a greater than 5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all ma or systems including payload, central truss, nuclear reactor (including divertor and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, power utilization, and component design.

  8. A Subcritical, Gas-Cooled Fast Transmutation Reactor with a Fusion Neutron Source

    SciTech Connect

    Stacey, W.M.; Beavers, V.L.; Casino, W.A.; Cheatham, J.R.; Friis, Z.W.; Green, R.D.; Hamilton, W.R.; Haufler, K.W.; Hutchinson, J.D.; Lackey, W.J.; Lorio, R.A.; Maddox, J.W.; Mandrekas, J.; Manzoor, A.A.; Noelke, C.A.; Oliveira, C. de; Park, M.; Tedder, D.W.; Terry, M.R.; Hoffman, E.A.

    2005-05-15

    A design is presented for a subcritical, He-cooled fast reactor, driven by a tokamak D-T fusion neutron source, for the transmutation of spent nuclear fuel (SNF). The reactor is fueled with coated transuranic (TRU) particles and is intended for the deep-burn (>90%) transmutation of the TRUs in SNF without reprocessing of the coated fuel particles. The reactor design is based on the materials, fuel, and separations technologies under near-term development in the U.S. Department of Energy (DOE) Nuclear Energy Program and on the plasma physics and fusion technologies under near-term development in the DOE Fusion Energy Sciences Program, with the objective of intermediate-term ({approx}2040) deployment. The physical and performance characteristics and research and development requirements of such a reactor are described.

  9. Advanced Fast Reactor - 100 (AFR-100) Report for the Technical Review Panel

    SciTech Connect

    Grandy, Christopher; Sienicki, James J.; Moisseytsev, Anton; Krajtl, Lubomir; Farmer, Mitchell T.; Kim, Taek K.; Middleton, B.

    2014-06-04

    This report is written to provide an overview of the Advanced Fast Reactor-100 in the requested format for a DOE technical review panel. This report was prepared with information that is responsive to the DOE Request for Information, DE-SOL-0003674 Advanced Reactor Concepts, dated February 27, 2012 from DOE’s Office of Nuclear Energy, Office of Nuclear Reactor Technologies. The document consists of two main sections. The first section is a summary of the AFR-100 design including the innovations that are incorporated into the design. The second section contains a series of tables that respond to the various questions requested of the reactor design team from the subject DOE RFI.

  10. A spherical torus nuclear fusion reactor space propulsion vehicle concept for fast interplanetary travel

    NASA Astrophysics Data System (ADS)

    Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

    1999-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a>5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all major systems including payload, central truss, nuclear reactor (including diverter and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, and component design.

  11. Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor

    SciTech Connect

    Guidez, J.; Martin, L.; Dupraz, R.

    2006-07-01

    The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960's, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 deg. C, Inspection of the conical shell supporting the core dia-grid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remote inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960's, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant. (authors)

  12. Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Rao, K. Bhanu Sankara

    2009-04-01

    The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

  13. Progress reports for Gen IV sodium fast reactor activities FY 2007.

    SciTech Connect

    Cahalan, J. E.; Tentner, A. M.; Nuclear Engineering Division

    2007-10-04

    An important goal of the US DOE Sodium Fast Reactor (SFR) program is to develop the technology necessary to increase safety margins in future fast reactor systems. Although no decision has been made yet about who will build the next demonstration fast reactor, it seems likely that the construction team will include a combination of international companies, and the safety design philosophy for the reactor will reflect a consensus of the participating countries. A significant amount of experience in the design and safety analysis of Sodium Fast Reactors (SFR) using oxide fuel has been developed in both Japan and France during last few decades. In the US, the traditional approach to reactor safety is based on the principle of defense-in-depth, which is usually expressed in physical terms as multiple barriers to release of radioactive material (e.g. cladding, reactor vessel, containment building), but it is understood that the 'barriers' may consist of active systems or even procedures. As implemented in a reactor design, defense-in-depth is classed in levels of safety. Level 1 includes measures to specify and build a reliable design with significant safety margins that will perform according to the intentions of the designers. Level 2 consists of additional design measures, usually active systems, to protect against unlikely accidental events that may occur during the life of the plant. Level 3 design measures are intended to protect the public in the event of an extremely unlikely accident not foreseen to occur during the plant's life. All of the design measures that make up the first three levels of safety are within the design basis of the plant. Beyond Level 3, and beyond the normal design basis, there are accidents that are not expected to occur in a whole generation of plants, and it is in this class that severe accidents, i.e. accidents involving core melting, are included. Beyond design basis measures to address severe accidents are usually identified as being

  14. Development of Fast Reactor Structural Integrity Monitoring Technology Using Optical Fiber Sensors

    NASA Astrophysics Data System (ADS)

    Matsuba, Ken-Ichi; Ito, Chikara; Kawahara, Hirotaka; Aoyama, Takafumi

    Significant thermal stresses are loaded onto the structures of sodium-cooled fast reactor (SFR) due to high temperature and large temperature gradients associated with employing sodium coolant with its high thermal conductivity and low heat capacity. Therefore, it is important to monitor the temperature variation, related stress and displacement, and vibration in the cooling system piping and components in order to assure structural integrity while the reactor plant is in-service. SFR structural integrity monitoring can be enhanced by an optical fiber sensor, which is capable of continuous or dispersed distribution measurements of various properties such as radiation dose, temperature, strain, displacement and acceleration. In the experimental fast reactor Joyo, displacement and vibration measurements of the primary cooling system have been carried out using Fiber Bragg Grating (FBG) sensors to evaluate the durability and measurement accuracy of FBG sensors in a high gamma-ray environment. The data were successfully obtained with no significant signal loss up to an accumulated gamma-ray dose of approximately 4×104 Gy corresponding to 120 EFPDs (effective full power days) operation. Measured displacement of the piping support was nearly equal to the calculated thermal displacement. Measured vibration power spectra of the piping support were similar to those measured with a reference acceleration sensor. The measured results indicate that the FBG sensor is suitable for monitoring the displacement and vibration aspects of fast reactor cooling system integrity in a high gamma-ray environment.

  15. Ferritic steels for sodium-cooled fast reactors: Design principles and challenges

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Vijayalakshmi, M.

    2010-09-01

    An overview of the current status of development of ferritic steels for emerging fast reactor technologies is presented in this paper. The creep-resistant 9-12Cr ferritic/martensitic steels are classically known for steam generator applications. The excellent void swelling resistance of ferritic steels enabled the identification of their potential for core component applications of fast reactors. Since then, an extensive knowledge base has been generated by identifying the empirical correlations between chemistry of the steels, heat treatment, structure, and properties, in addition to their in-reactor behavior. A few concerns have also been identified which pertain to high-temperature irradiation creep, embrittlement, Type IV cracking in creep-loaded weldments, and hard zone formation in dissimilar joints. The origin of these problems and the methodologies to overcome the limitations are highlighted. Finally, the suitability of the ferritic steels is re-evaluated in the emerging scenario of the fast reactor technology, with a target of achieving better breeding ratio and improved thermal efficiency.

  16. A Heterogeneous Sodium Fast Reactor Designed to Transmute Minor Actinide Actinide Waste Isotopes into Plutonium Fuel

    SciTech Connect

    Samuel E. Bays

    2011-02-01

    An axial heterogeneous sodium fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting Am-241 into Pu-238. This Pu-238 is then co-recycled with the spent driver fuel to make new driver fuel. Because Pu-238 is significantly more fissile than Am-241 in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because, the Am-241 neutron capture worth is significantly stronger in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap which recovers the axial leakage lost by the active core. The sodium fast reactor proposed by this work is designed as an overall transuranic burner. Therefore, a low transuranic conversion ratio is achieved by a degree of core flattening which increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Because minor actinides are irradiated only once in the axial target region; elemental partitioning is not required. This fact enables the use of metal targets with electrochemical reprocessing. Therefore, the irradiation environment of both drivers and targets was constrained to ensure applicability of the established experience database for metal alloy sodium fast reactor fuels.

  17. Potentiality of a ceramic membrane reactor for the laccase-catalyzed removal of bisphenol A from secondary effluents.

    PubMed

    Arca-Ramos, A; Eibes, G; Feijoo, G; Lema, J M; Moreira, M T

    2015-11-01

    In this study, the removal of bisphenol A (BPA) by laccase in a continuous enzymatic membrane reactor (EMR) was investigated. The effects of key parameters, namely, type of laccase, pH, and enzyme activity, were initially evaluated. Once optimal conditions were determined, the continuous removal of the pollutant in an EMR was assessed in synthetic and real biologically treated wastewaters. The reactor configuration consisted of a stirred tank reactor coupled to a ceramic membrane, which prevented the sorption of the pollutant and allowed the recovery and recycling of laccase. Nearly complete removal of BPA was attained under both operation regimes with removal yields above 94.5 %. In experiments with real wastewater, the removal of BPA remained high while the presence of colloids and certain ions and the formation of precipitates on the membrane potentially affected enzyme stability and made necessary the periodic addition of laccase. Polymerization and degradation were observed as probable mechanisms of BPA transformation by laccase. PMID:26209248

  18. Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors

    SciTech Connect

    Roald A. Wigeland; James E. Cahalan

    2011-04-01

    Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.

  19. The Integral Fast Reactor concept: Today's hope for tomorrow's electrical energy needs

    SciTech Connect

    Dwight, C.C.; Phipps, R.D.

    1989-01-01

    Acid rain and the greenhouse effect are getting more attention as their impacts on the environment become evident around the world. Substantial evidence indicates that fossil fuel combustion for electrical energy production activities is a key cause of those problems. A change in electrical energy production policy is essential to a stable, healthy environment. That change is inevitable, it's just a matter of when and at what cost. Vision now, instead of reaction later, both in technological development and public perception, will help to limit the costs of change. The Integral Fast Reactor (IFR) is a visionary concept developed by Argonne National Laboratory that involves electrical energy production through fissioning of heavy metals by fast neutrons in a reactor cooled by liquid sodium. Physical characteristics of the coolant and fuel give the reactor impressive characteristics of inherent and passive safety. Spent fuel is pyrochemically reprocessed and returned to the reactor in the IFR's closed fuel cycle. Advantages in waste management are realized, and the reactor has the potential for breeding, i.e., producing as much or more fuel than it uses. This paper describes the IFR concept and shows how it is today's hope for tomorrow's electrical energy needs. 14 refs., 1 fig., 1 tab.

  20. SSTAR: The U.S. Lead-Cooled Fast Reactor (LFR)

    SciTech Connect

    Smith, C F; Halsey, W G; Brown, N W; Sienicki, J J; Moisseytsev, A; Wade, D C

    2007-09-25

    It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the Global Nuclear Energy Partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the Small Secure Transportable Autonomous Reactor (SSTAR) reactor has been under ongoing development under the U.S. Generation IV Nuclear Energy Systems Initiative. It a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation aims, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the U.S. Generation IV Lead-cooled Fast Reactor system.

  1. Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

    NASA Astrophysics Data System (ADS)

    Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

    2015-12-01

    U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

  2. Porous ceramic/agarose composite adsorbents for fast protein liquid chromatography.

    PubMed

    Xia, Haifeng; Jin, Xionghua; Wu, Puqiang; Zheng, Zhiyong

    2012-02-01

    Porous ceramic/agarose composite adsorbents were designed and prepared with silica ceramic beads and 4% agarose gel, and then functionalized with a special ligand carboxymethyl. A novel method was introduced to fabricating of the porous silica ceramic beads. The morphology of SEM shows a spherical shape and a porous structure of the ceramic beads. Nitrogen adsorption-desorption analysis gives an average pore size of 287.5 Å, a BET surface area of 29.33 m²/g and a porosity of 41.8%, respectively. Additionally, X-ray diffraction pattern indicates that the amorphous silica has been transformed into two crystal phases of quartz and cristobalite, leading to a porous and rigid skeleton and ensuring the application of the composite beads at high flow velocities. Lysozyme of hen egg-white with the activity of 12,700 U/mg was purified by the composite ion-exchanger in one step and the recovery and purification factor reaches 95.2% and 7.9, respectively. PMID:22226554

  3. Under-Sodium Viewing: A Review of Ultrasonic Imaging Technology for Liquid Metal Fast Reactors

    SciTech Connect

    Griffin, Jeffrey W.; Peters, Timothy J.; Posakony, Gerald J.; Chien, Hual-Te; Bond, Leonard J.; Denslow, Kayte M.; Sheen, Shuh-Haw; Raptis, Paul

    2009-03-27

    This current report is a summary of information obtained in the "Information Capture" task of the U.S. DOE-funded "Under Sodium Viewing (USV) Project." The goal of the multi-year USV project is to design, build, and demonstrate a state-of-the-art prototype ultrasonic viewing system tailored for periodic reactor core in-service monitoring and maintenance inspections. The study seeks to optimize system parameters, improve performance, and re-establish this key technology area which will be required to support any new U.S. liquid-metal cooled fast reactors.

  4. 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.

  5. Comparative analysis of using natural and radiogenic lead as heat-transfer agent in fast reactors

    NASA Astrophysics Data System (ADS)

    Laas, R. A.; Gizbrekht, R. V.; Komarov, P. A.; Nesterov, V. N.

    2016-06-01

    Fast reactors with lead coolant have several advantages over analogues. Performance can be further improved by replacement of natural composition lead with radiogenic one. Thus, two main issues need to be addressed: induced radioactivity in coolant and efficient neutron multiplication factor in the core will be changed and need to be estimated. To address these issues analysis of the scheme of the nuclear transformations in the lead heat-transfer agent in the process of radiation was carried out. Induced radioactivity of radiogenic and natural lead has been studied. It is shown that replacement of lead affects multiplication factor in a certain way. Application of radiogenic lead can significantly affect reactor operation.

  6. Development and analysis of a compact low-conversion ratio fast burner reactor.

    SciTech Connect

    Smith, M. A.; Hill, R. N.; Nuclear Engineering Division

    2006-05-12

    This report explores design options for compact fast burner reactors that can achieve low conversion ratios. Operational characteristics and whole-core reactivity coefficients are generated and contrasted with low conversion ratio designs of previous studies. A compact core point design is then selected and detailed reactivity coefficients are displayed and discussed. The effectiveness of fast spectrum systems for actinide transmutation has been well documented. The key advantage of the fast spectrum resides in the severely reduced capture/fission ratios. this inhibits the production of the higher actinides that dominate the long-term radiotoxicity of nuclear waste. In conventional fast burner studies, the transmutation rate was limited by constraints placed on the fuel composition. In an earlier phase of this study the entire range of fuel compositions (including non-uranium fuel) was explored to assess the performance and safety limits of fast burner reactor systems. In this report, similar fuel compositions are utilized for application in compact configurations to achieve conversion ratios below 0.5.

  7. Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

    SciTech Connect

    Higinbotham, W.A.

    1994-11-07

    The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 {+-} 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF.

  8. The effects of spectral shift absorbers on the design and safety of fast spectrum space reactors

    NASA Astrophysics Data System (ADS)

    King, Jeffrey Charles

    Spectral Shift Absorbers (SSAs) are incorporated into space reactors to maintain them sufficiently subcritical when submerged in seawater or wet sand and subsequently flooded, following a launch abort accident. The effect of four SSAs (samarium-149, europium-151, gadolinium-155, and gadolinium-157) on the submersion criticality, operation, and temperature reactivity feedback of the thermal spectrum reactors developed in the Systems for Nuclear Auxilary Power (SNAP) program is extensively documented. Recent work on SSAs in fast spectrum space reactors, preferred for compactness and higher powers, has focused on rhenium as the primary SSA. In addition to identifying additional SSAs, the present work investigates the effects of SSAs on the overall size and mass, temperature reactivity feedback, and operational lifetime of fast spectrum space reactors. The fast spectrum S4 reactor has a sectored Mo-14%Re solid-core, loadedwith UN fuel, cooled by He-30%Xe, and designed to avoid single point failures at a steady thermal power of 550 kWth. The addition of SSAs to the reactor core increases the fuel enrichment and decreases the size and mass of the reactor and the radiation shadow shield. SSA additions of boron-10, europium-151, gadolinium-155 and iridium result in the smallest and lightest S4 reactors. The effects of SSA additions on the operational lifetime and the temperature and burnup reactivity coefficients of the S^4 reactor are studied. An increasein fuel enrichment with SSAs markedly increases the operational lifetime by decreasing the burnup reactivity coefficient with only a slight decrease in the temperature reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt%), the temperature and burnup reactivity coefficients are the highest (-0.2709 ¢/K and -1.3470 /atom%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to -0.2649 ¢/K and -1.0230 /atom%, and

  9. High temperature ceramic membrane reactors for coal liquid upgrading. Quarterly report No. 2, December 21, 1989--March 20, 1990

    SciTech Connect

    Tsotsis, T.T.

    1992-06-19

    Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we will evaluate the performance of Sel-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

  10. High temperature ceramic membrane reactors for coal liquid upgrading. Quarterly report No. 8, June 21, 1991--September 20, 1991

    SciTech Connect

    Tsotsis, T.T.

    1992-06-19

    Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we will evaluate the performance of Sel-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will also be investigated.

  11. High temperature ceramic membrane reactors for coal liquid upgrading. Quarterly report No. 7, March 21, 1991--June 20, 1991

    SciTech Connect

    Tsotsis, T.T.

    1992-06-19

    Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we will evaluate the performance of Sel-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

  12. Overview of past and current activities on fuels for fast reactors at the Institute for Transuranium Elements

    NASA Astrophysics Data System (ADS)

    Fernandez, A.; McGinley, J.; Somers, J.; Walter, M.

    2009-07-01

    Nuclear energy has the potential to provide a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gas emissions. The renewal of interest in fast neutron spectra reactors to meet more ambitious sustainable development criteria (i.e., resource maximisation and waste minimisation), opens a favourable framework for R&D activities in this area. The Institute for Transuranium Elements has extensive experience in the fabrication, characterization and irradiation testing (Phénix, Dounreay, Rapsodie) of fast reactor fuels, in oxide, nitride and carbide forms. An overview of these past and current activities on fast reactor fuels is presented.

  13. Development of material irradiation rig with precision temperature control in experimental fast reactor JOYO

    NASA Astrophysics Data System (ADS)

    Kataoka, H.; Yasu, T.; Takatsudo, H.; Miyakawa, S.

    1998-10-01

    In the experimental fast reactor JOYO, an on-line instrumented material irradiation rig has been developed to acquire various irradiation data. Material Testing Rig with Temperature Control (MARICO) is capable of collecting creep rupture strength data for fast reactor fuel cladding materials. The specimen temperature in the MARICO-1 rig is controlled with an accuracy of ±4°C by the use of a gas gap method similar to that of FFTF/MOTA. Presently, irradiation testing using MARICO-1 has been so successful that the development of MARICO-2 is under way for completion by the early part of the next century. MARICO-2 will contain an electrical heater for temperature control, a larger irradiation space and other data collecting improvements.

  14. Fast reactor safety: proceedings of the international topical meeting. Volume 1

    SciTech Connect

    Not Available

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences.

  15. An improved resonance self-shielding method for heterogeneous fast reactor assembly and core calculations

    SciTech Connect

    Lee, C.; Yang, W. S.

    2013-07-01

    An improved resonance self-shielding method has been developed to accurately estimate the effective multigroup cross sections for heterogeneous fast reactor assembly and core calculations. In the method, the heterogeneity effect is considered by the use of isotopic escape cross sections while the resonance interference effect is accounted for through the narrow resonance approximation or slowing-down calculations for specific compositions. The isotopic escape cross sections are calculated by solving fixed-source transport equations with the method of characteristics for the whole problem domain. This method requires no pre-calculated resonance integral tables or parameters that are typically necessary in the subgroup method. Preliminary results for multi pin-cell fast reactor problems show that the escape cross sections estimated from the explicit-geometry fixed source calculations produce more accurate eigenvalue and self-shielded effective cross sections than those from conventional one-dimensional geometry models. (authors)

  16. Fast pyrolysis of eucalyptus waste in a conical spouted bed reactor.

    PubMed

    Amutio, Maider; Lopez, Gartzen; Alvarez, Jon; Olazar, Martin; Bilbao, Javier

    2015-10-01

    The fast pyrolysis of a forestry sector waste composed of Eucalyptus globulus wood, bark and leaves has been studied in a continuous bench-scale conical spouted bed reactor plant at 500°C. A high bio-oil yield of 75.4 wt.% has been obtained, which is explained by the suitable features of this reactor for biomass fast pyrolysis. Gas and bio-oil compositions have been determined by chromatographic techniques, and the char has also been characterized. The bio-oil has a water content of 35 wt.%, and phenols and ketones are the main organic compounds, with a concentration of 26 and 10 wt.%, respectively. In addition, a kinetic study has been carried out in thermobalance using a model of three independent and parallel reactions that allows quantifying this forestry waste's content of hemicellulose, cellulose and lignin. PMID:26203554

  17. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    SciTech Connect

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  18. High temperature ceramic membrane reactors for coal liquid upgrading. Quarterly report No. 10, December 21, 1991--March 20, 1992

    SciTech Connect

    Tsotsis, T.T.

    1992-07-01

    In this project we will study a novel process concept, i.e., the use of ceramic membrane reactors in upgrading of coal model compounds and coal derived liquids. In general terms, the USC research team is responsible for constructing and operating the membrane reactor apparatus and for testing various inorganic membranes for the upgrading of coal derived asphaltenes and coal model compounds. The USC effort will involve the principal investigator of this project and two graduate research assistants. The ALCOA team is responsible for the preparation of the inorganic membranes, for construction and testing of the ceramic membrane modules, and for measurement of their transport properties. The ALCOA research effort will involve Dr. Paul K. T. Liu, who is the project manager of the ALCOA research team, an engineer and a technician. UNOCAL`s contribution will be limited to overall technical assistance in catalyst preparation and the operation of the laboratory upgrading membrane reactor and for analytical back-up and expertise in oil analysis and materials characterization. UNOCAL is a no-cost contractor but will be involved in all aspects of the project, as deemed appropriate.

  19. Measurements of thermal-hydraulic parameters in liquid-metal-cooled fast-breeder reactors

    SciTech Connect

    Sackett, J.I.

    1983-01-01

    This paper discusses instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, sodium purity, and leakage. The paper identifies the overall instrumentation requirements for LMFBR's and those aspects of instrumentation which are unique or of special concern to LMFBR systems. It also gives an overview of the status of instrument design and performance.

  20. Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors

    SciTech Connect

    Griffin, C W; Rasmussen, D E; Lloyd, M H

    1983-01-01

    The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities.

  1. Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

    NASA Technical Reports Server (NTRS)

    Mayo, W.; Lantz, E.

    1973-01-01

    A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

  2. Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

    NASA Astrophysics Data System (ADS)

    Herrero, J. J.; Ochoa, R.; Martínez, J. S.; Díez, C. J.; García-Herranz, N.; Cabellos, O.

    2014-04-01

    The assessment of the uncertainty levels on the design and safety parameters for the innovative European Sodium Fast Reactor (ESFR) is mandatory. Some of these relevant safety quantities are the Doppler and void reactivity coefficients, whose uncertainties are quantified. Besides, the nuclear reaction data where an improvement will certainly benefit the design accuracy are identified. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

  3. Testing and Research Capabilities at the Sandia Fast Pulsed Reactor Facility

    NASA Astrophysics Data System (ADS)

    Berry, Donald T.

    1994-07-01

    A wide variety of space-based system components have been qualified for use through neutron irradiation testing performed at the Sandia Pulsed Reactor (SPR) Facility. The SPR Facility is the operating location for two fast burst reactors, SPR II and SPR III, which have been used to induce neutron and gamma damage in electronic components and other materials for customers in the Department of Energy, Department of Defense, NASA, and the private sector. In addition to the pulse mode of operation, during which peak fluxes of up to 1023 n/m2-s are achieved, the steady state mode allows for the long term irradiation of components and systems in a fast neutron environment at a flux of up to 5×1015 n/m2-s. The SPR reactors are operated in a 9.2 meter diameter exposure cell, or Kiva, suitable for the irradiation of large test articles external to the reactors. Currently, a new upgraded version of SPR III (SPR HIM) is in fabrication; a unique feature of SPR HIM is its 190 mm (usable diameter) central irradiation cavity, the largest of any U.S. fast burst reactor. An improved cooling system permits continuous operation at power levels in excess of 20 kWt. The SPR Facility is also the operating site for a critical assembly which was used to characterize prototypic fuels in arrays appropriate for the Space Nuclear Thermal Propulsion Program. Work continues on use of the facility to design, build, and operate critical assemblies for a diverse customer base.

  4. R and D program for core instrumentation improvements devoted for French sodium fast reactors

    SciTech Connect

    Jeannot, J. P.; Rodriguez, G.; Jammes, C.; Bernardin, B.; Portier, J. L.; Jadot, F.; Maire, S.; Verrier, D.; Loisy, F.; Prele, G.

    2011-07-01

    Under the framework of French R and D studies for Generation IV reactors and more specifically for sodium-cooled fast reactors (SFR); the CEA, EDF and AREVA have launched a joint coordinated research programme. This paper deals with the R and D sets out to achieve better inspection, maintenance, availability and decommissioning. In particular the instrumentation requirements for core monitoring and detection in the case of accidental events. Requirements mainly involve diversifying the means of protection and improving instrumentation performance in terms of responsiveness and sensitivity. Operation feedback from the Phenix and Superphenix prototype reactors and studies, carried out within the scope of the EFR projects, has been used to define the needs for instrumentation enhancement. (authors)

  5. Development and application of modeling tools for sodium fast reactor inspection

    SciTech Connect

    Le Bourdais, Florian; Marchand, Benoît; Baronian, Vahan

    2014-02-18

    To support the development of in-service inspection methods for the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID) project led by the French Atomic Energy Commission (CEA), several tools that allow situations specific to Sodium cooled Fast Reactors (SFR) to be modeled have been implemented in the CIVA software and exploited. This paper details specific applications and results obtained. For instance, a new specular reflection model allows the calculation of complex echoes from scattering structures inside the reactor vessel. EMAT transducer simulation models have been implemented to develop new transducers for sodium visualization and imaging. Guided wave analysis tools have been developed to permit defect detection in the vessel shell. Application examples and comparisons with experimental data are presented.

  6. Development and application of modeling tools for sodium fast reactor inspection

    NASA Astrophysics Data System (ADS)

    Le Bourdais, Florian; Marchand, Benoît; Baronian, Vahan

    2014-02-01

    To support the development of in-service inspection methods for the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID) project led by the French Atomic Energy Commission (CEA), several tools that allow situations specific to Sodium cooled Fast Reactors (SFR) to be modeled have been implemented in the CIVA software and exploited. This paper details specific applications and results obtained. For instance, a new specular reflection model allows the calculation of complex echoes from scattering structures inside the reactor vessel. EMAT transducer simulation models have been implemented to develop new transducers for sodium visualization and imaging. Guided wave analysis tools have been developed to permit defect detection in the vessel shell. Application examples and comparisons with experimental data are presented.

  7. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    SciTech Connect

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  8. Conjugate heat transfer analysis of multiple enclosures in prototype fast breeder reactor

    SciTech Connect

    Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C.

    1995-09-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled reactor under design. The main vessel of the reactor serves as the primary boundary. It is surrounded by a safety vessel which in turn is surrounded by biological shield. The gaps between them are filled with nitrogen. Knowledge of temperature distribution prevailing under various operating conditions is essential for the assessment of structural integrity. Due to the presence of cover gas over sodium free level within the main vessel, there are sharp gradients in temperatures. Also cover gas height reduces during station blackout conditions due to sodium level rise in main vessel caused by temperature rise. This paper describes the model used to analyse the natural convection in nitrogen, conduction in structures and radiation interaction among them. Results obtained from parametric studies for PFBR are also presented.

  9. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    SciTech Connect

    Samuel Bays; Pavel Medvedev; Michael Pope; Rodolfo Ferrer; Benoit Forget; Mehdi Asgari

    2009-04-01

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  10. Analysis of fission-product effects in a Fast Mixed-Spectrum Reactor concept

    SciTech Connect

    White, J.R.; Burns, T.J.

    1980-02-01

    The Fast Mixed-Spectrum Reactor (FMSR) concept has been proposed by BNL as a means of alleviating certain nonproliferation concerns relating to civilian nuclear power. This breeder reactor concept has been tailored to operate on natural uranium feed (after initial startup), thus eliminating the need for fuel reprocessing. The fissile material required for criticality is produced, in situ, from the fertile feed material. This process requires that large burnup and fluence levels be achievable, which, in turn, necessarily implies that large fission-product inventories will exist in the reactor. It was the purpose of this study to investigate the effects of large fission-product inventories and to analyze the effect of burnup on fission-product nuclide distributions and effective cross sections. In addition, BNL requested that a representative 50-group fission-product library be generated for use in FMSR design calculations.

  11. Design, optimization and evaluation of a free-fall biomass fast pyrolysis reactor and its products

    NASA Astrophysics Data System (ADS)

    Ellens, Cody James

    The focus of this work is a radiatively heated, free-fall, fast pyrolysis reactor. The reactor was designed and constructed for the production of bio-oil from the fast pyrolysis of biomass. A central composite design of experiments was performed to evaluate the novel reactor by varying four operating conditions: reactor temperature, biomass particle size, carrier gas flow rate and biomass feed rate. Maximum bio-oil yields of 72 wt % were achieved at a heater set point temperature of 600 °C, using particle sizes of 300 micron, carrier gas flow rates of 4 sL/min and Red oak biomass feed rates of 1.75 kg/hr. Optimal operating conditions were identified for maximum bio-oil yields at a heater set point temperature of 572 °C, feeding 240 micron sized Red oak biomass particles at 2 kg/hr. Carrier gas flow rates were not found to be significant over the 1 -- 5 sL/min range tested.

  12. A survey of alternative once-through fast reactor core designs

    SciTech Connect

    Fei, T.; Richard, J. G.; Kersting, A. R.; Don, S. M.; Oi, C.; Driscoll, M. J.; Shwageraus, E.

    2012-07-01

    Reprocessing of Light Water Reactor (LWR) spent fuel to recover plutonium or transuranics for use in Sodium cooled Fast Reactors (SFRs) is a distant prospect in the U.S.A. This has motivated our evaluation of potentially cost-effective operation of uranium startup fast reactors (USFRs) in a once-through mode. This review goes beyond findings reported earlier based on a UC fueled MgO reflected SFR to describe a broader parametric study of options. Cores were evaluated for a variety of fuel/coolant/reflector combinations: UC/UZr/UO{sub 2}/UN;Na/Pb; MgO/SS/Zr. The challenge is achieving high burnup while minimizing enrichment and respecting both cladding fluence/dpa and reactivity lifetime limits. These parametric studies show that while UC fuel is still the leading contender, UO{sub 2} fuel and ZrH 1.7 moderated metallic fuel are also attractive if UC proves to be otherwise inadequate. Overall, these findings support the conclusion that a competitive fuel cycle cost and uranium utilization compared to LWRs is possible for SFRs operated on a once-through uranium fueled fuel cycle. In addition, eventual transition to TRU recycle mode is studied, as is a small test reactor to demonstrate key features. (authors)

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

    NASA Astrophysics Data System (ADS)

    Bays, Samuel Eugene

    2008-10-01

    In the past several years there has been a renewed interest in sodium fast reactor (SFR) technology for the purpose of destroying transuranic waste (TRU) produced by light water reactors (LWR). The utility of SFRs as waste burners is due to the fact that higher neutron energies allow all of the actinides, including the minor actinides (MA), to contribute to fission. It is well understood that many of the design issues of LWR spent nuclear fuel (SNF) disposal in a geologic repository are linked to MAs. Because the probability of fission for essentially all the "non-fissile" MAs is nearly zero at low neutron energies, these isotopes act as a neutron capture sink in most thermal reactor systems. Furthermore, because most of the isotopes produced by these capture reactions are also non-fissile, they too are neutron sinks in most thermal reactor systems. Conversely, with high neutron energies, the MAs can produce neutrons by fast fission. Additionally, capture reactions transmute the MAs into mostly plutonium isotopes, which can fission more readily at any energy. The transmutation of non-fissile into fissile atoms is the premise of the plutonium breeder reactor. In a breeder reactor, not only does the non-fissile "fertile" U-238 atom contribute fast fission neutrons, but also transmutes into fissile Pu-239. The fissile value of the plutonium produced by MA transmutation can only be realized in fast neutron spectra. This is due to the fact that the predominate isotope produced by MA transmutation, Pu-238, is itself not fissile. However, the Pu-238 fission cross section is significantly larger than the original transmutation parent, predominately: Np-237 and Am-241, in the fast energy range. Also, Pu-238's fission cross section and fission-to-capture ratio is almost as high as that of fissile Pu-239 in the fast neutron spectrum. It is also important to note that a neutron absorption in Pu-238, that does not cause fission, will instead produce fissile Pu-239. Given this

  14. Engineering development of ceramic membrane reactor system for converting natural gas to hydrogen and synthesis gas for liquid transportation fuels

    SciTech Connect

    1998-05-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through April 1998.

  15. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect

    1999-12-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through November 1999.

  16. Impact of transmutation issues on interpretation of data obtained from fast reactor irradiation experiments

    NASA Astrophysics Data System (ADS)

    Greenwood, L. R.; Garner, F. A.

    2004-08-01

    The subject of fission-fusion correlation is usually cast in terms of reactor-to-reactor differences, but recently the fusion community has become aware of the impact of differences within a given surrogate facility, especially in constant time experiments when different dose levels are attained in different positions of one reactor. For some materials, it is not safe to assume that in-reactor spectral variations are small and of no consequence. This point is illustrated using calculations for fusion-relevant materials that were irradiated in the Fast Flux Test Facility-Materials Open Test Assembly (FFTF-MOTA) over a wide range of in-core and out-of-core positions spanning more than two orders of magnitude in dpa rate. It is shown that although both the neutron spectrum and flux changes, the spectral effectiveness factor, dpa/10 22 n/cm 2 ( E>0.1 MeV), remains remarkably constant over this range. The transmutation rate per dpa varies strongly with reactor position, however.

  17. Impact of Transmutation Issues on Interpretation of Data Obtained From Fast Reactor Irradiation Experiments

    SciTech Connect

    Greenwood, Lawrence R.; Garner, Francis A.

    2004-04-15

    The subject of fission-fusion correlation is usually cast in terms of reactor-to-reactor differences, but recently the fusion community has become aware of the impact of differences within a given surrogate facility, especially in constant time experiments when different dose levels are attained in different positions of one reactor. For some materials, it is not safe to assume that in-reactor spectral variations are small and of no consequence. This point is illustrated using calculations for fusion-relevant materials that were irradiated in the Fast Flux Test Facility – Materials Open Test Assembly (FFTF-MOTA) over a wide range of in-core and out of core positions spanning more than two orders of magnitude in dpa rate. It is shown that although both the neutron spectrum and flux changes, the spectral effectiveness factor, dpa/10(22) superscript n/cm(2) superscript (E > 0.1 MeV), remains remarkably constant over this range. The transmutation rate per dpa varies strongly with reactor position, however.

  18. SVBR-100 module-type fast reactor of the IV generation for regional power industry

    NASA Astrophysics Data System (ADS)

    Zrodnikov, A. V.; Toshinsky, G. I.; Komlev, O. G.; Stepanov, V. S.; Klimov, N. N.

    2011-08-01

    In the report the following is presented: basic conceptual provisions of the innovative nuclear power technology (NPT) based on modular fast reactors (FR) SVBR-100, summarized results of calculations of the reactor, analysis of the opportunities of multi-purpose application of such reactor facilities (RF) including export potentials with due account of nonproliferation requirements. The most important features of the proposed NPT analyzed in the report are as follows: (1) integral (monoblock) arrangement of the primary circuit equipment with entire elimination of the primary circuit pipelines and valves that considerably reduces the construction and assembly works period and coupling with high boiling point of lead-bismuth coolant (LBC) deterministically eliminates accidents of the LOCA type, (2) option for 100 MWe power and dimensions of the reactor provide: on the one hand, an opportunity to transport the reactor monoblock in factory-readiness by railway as well as other kinds of transport, on the other hand, core breeding ratio (CBR) exceeds 1 while MOX-fuel is used. The preferable area of application of RF SVBR-100 is regional and small power requiring power-units of electric power in a range of (100-600) MW, which could be used for cogeneration-based district heating while locating them nearby cities as well as for generation of electric power in a mode of load tracking in the regions with low network systems.

  19. Preliminary safety calculations to improve the design of Molten Salt Fast Reactor

    SciTech Connect

    Brovchenko, M.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Capellan, N.; Ghetta, V.; Laureau, A.

    2012-07-01

    Molten salt reactors are liquid fuel reactors so that they are flexible in operation but very different in the safety approach from solid fuel reactors. This study bears on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the R and D studies. This paper presents the first step of the safety approach: the systematic description of the MSFR, limited here to the main systems surrounding the core. This systematic description is the basis on which we will be able to devise accidental scenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarios lead to reactor shut down. Because of the decay heat generated in the fuel salt, it must be cooled. After the description of the tools developed to calculate the residual heat, the different contributions are discussed in this study. The decay heat of fission products in the MSFR is evaluated to be low (3% of nominal power), mainly due to the reprocessing that transfers the fission products to the gas reprocessing unit. As a result, the contribution of the actinides is significant (0.5% of nominal power). The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable (> 1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system. (authors)

  20. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers 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 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  1. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    SciTech Connect

    Oktamuliani, Sri Su’ud, Zaki

    2015-09-30

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.

  2. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    NASA Astrophysics Data System (ADS)

    Oktamuliani, Sri; Su'ud, Zaki

    2015-09-01

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.

  3. Thermal criteria to compare fast reactors coolants for the intermediate loop

    SciTech Connect

    Saez, Manuel; Rodriguez, Gilles

    2007-07-01

    Fast Breeder Reactors (FBR) are typically using a liquid metal as the primary coolant. Up to now, sodium is the referenced coolant for all large-scale FBR, but lead and sodium-potassium alloy have both also been used successfully for smaller rigs. The French Atomic Energy Commission (CEA) has an extensive experience and significant expertise in Sodium cooled Fast Reactors (SFR) over the past 40 years of R and D and feedback experiments. Some improvements are needed on the SFR to meet the Generation IV goals, and in particular the safety and the reliability through the intermediate loop coolant. As sodium reacts exo-thermically with air and water and to eliminate the drawback of the water-sodium interaction when a steam generator tube is ruptured, CEA is involved in a substantial effort in order to investigate the interest to use an alternative coolant than sodium in the intermediate loop. This paper presents the main thermal criteria to compare Fast Reactors coolants for the intermediate loop under natural and forced convection. Neutronics considerations are not taken into account for the intermediate loop coolant. Transport, transfer and energetic criteria are analysed in the field of turbulent flows. Criteria are applied to the following potential coolant candidates: sodium, lithium, tin, bismuth, lead, lead-bismuth alloy, lead-lithium alloy, gallium, indium, potassium and sodium-potassium alloy. According to this thermal analysis, the gallium as heat transfer agent for the intermediate loop is considered as a promising candidate. For the discussion of the applicability of the gallium as heat transfer agent for the intermediate loop, a limited thermal hydraulic pre-sizing of a steam generator is undertaken using simple engineering methods implemented in COPERNIC code, a CEA tool dedicated to reactor systems pre-sizing. (authors)

  4. An assessment of liquid-metal centrifugal pumps at three fast reactors

    SciTech Connect

    Smith, M.S.; Wood, D.H.; Drischler, J.D. )

    1993-10-01

    The results of an analysis using data reports submitted to the Centralized Reliability Data Organization (CREDO) to predict the onset of the wearout life period for large sodium centrifugal pumps is described. For CREDO data collection and analysis purposes, a mechanical pump'' includes the pumping unit, its driver, and the coupling between the two. Statistical data were compiled from event reports received from three fast reactors: the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF) in the US and the JOYO Experimental Fast Reactor operated by the Power Reactor and Nuclear Fuel Development Corporation of Japan. Cumulative event rates were calculated for the investigated pumps at each facility and for the entire population. For all pumps, the event rate was computed as 34.4 event/million operating hours with 5 and 95% one-sided confidence limits of 26.3 and 44.4 event/million operating hours, respectively. The cumulative event rates for EBR-II, FFTF, and JOYO were computed as 30.0, 32.4, and 40.6 event/million pump operating hours, respectively. Results from EBR-II indicate that there is a definite time-dependent relationship between event rates and pump age; the common event mode at EBR-II is pump binding or seizing due to the buildup of sodium deposits in the vicinity of the lower labyrinth seal. There is no indication from FFTF that the six centrifugal pumps have reached the end of their useful life; these pumps have been event free for their last 40,000 operating hours. Following a 50,000-h even-free operating period at JOYO, bearings in the secondary pumps required additional unscheduled maintenance. However, there is no indication that these pumps have entered into the wearout life period; more data are required to draw any such conclusion.

  5. ODS Ferritic/martensitic alloys for Sodium Fast Reactor fuel pin cladding

    NASA Astrophysics Data System (ADS)

    Dubuisson, Philippe; Carlan, Yann de; Garat, Véronique; Blat, Martine

    2012-09-01

    The development of ODS materials for the cladding for Sodium Fast Reactors is a key issue to achieve the objectives required for the GEN IV reactors. CEA, AREVA and EDF have launched in 2007 an important program to determine the optimal fabrication parameters, and to measure and understand the microstructure and properties before, under and after irradiation of such cladding materials. The aim of this paper is to present the French program and the major results obtained recently at CEA on Fe-9/14/18Cr1WTiY2O3 ferritic/martensitic ODS materials. The first step of the program was to consolidate Fe-9/14/18Cr ODS materials as plates and bars to study the microstructure and the mechanical properties of the new alloys. The second step consists in producing tubes at a geometry representative of the cladding of new Sodium Fast Reactors. The optimization of the fabrication route at the laboratory scale is conducted and different tubes were produced. Their microstructure depends on the martensitic (Fe-9Cr) or ferritic (Fe-14Cr) structure. To join the plug to the tube, the reference process is the welding resistance. A specific approach is developed to model the process and support the development of the welds performed within the "SOPRANO" facility. The development at CEA of Fe-9/14/18Cr new ODS materials for the cladding for GENIV Sodium Fast Reactors is in progress. The first microstructural and mechanical characterizations are very encouraging and the full assessment and qualification of this new alloys and products will pass through the irradiation of specimens, tubes, fuel pins and subassemblies up to high doses.

  6. Simulation of reactor pulses in fast burst and externally driven nuclear assemblies

    NASA Astrophysics Data System (ADS)

    Green, Taylor Caldwell, IV

    The following research contributes original concepts to the fields of deterministic neutron transport modeling and reactor power excursion simulation. A deterministic neutron transport code was created to assess the value of new methods of determining neutron current, fluence, and flux values through the use of view factor and average path length calculations. The neutron transport code is also capable of modeling the highly anisotropic neutron transport of deuterium-tritium fusion external source neutrons using diffusion theory with the aid of a modified first collision source term. The neutron transport code was benchmarked with MCNP, an industry standard stochastic neutron transport code. Deterministic neutron transport methods allow users to model large quantities of neutrons without simulating their interactions individually. Subsequently, deterministic methods allow users to more easily couple neutron transport simulations with other physics simulations. Heat transfer and thermoelastic mechanics physics simulation modules were each developed and benchmarked using COMSOL, a commercial heat transfer and mechanics simulation software. The physics simulation modules were then coupled and used to simulate reactor pulses in fast burst and externally driven nuclear assemblies. The coupled system of equations represents a new method of simulating reactor pulses that allows users to more fully characterize pulsed assemblies. Unlike older methods of reactor pulse simulation, the method presented in this research does not require data from the operational reactor in order to simulate its behavior. The ability to simulate the coupled neutron transport and thermo-mechanical feedback present in pulsed reactors prior their construction would significantly enhance the quality of pulsed reactor pre-construction safety analysis. Additionally, a graphical user interface is created to allow users to run simulations and visualize the results using the coupled physics simulation

  7. Proceedings of the Office of Fusion Energy/DOE workshop on ceramic matrix composites for structural applications in fusion reactors

    SciTech Connect

    Jones, R.H. ); Lucas, G.E. )

    1990-11-01

    A workshop to assess the potential application of ceramic matrix composites (CMCs) for structural applications in fusion reactors was held on May 21--22, 1990, at University of California, Santa Barbara. Participants included individuals familiar with materials and design requirements in fusion reactors, ceramic composite processing and properties and radiation effects. The primary focus was to list the feasibility issues that might limit the application of these materials in fusion reactors. Clear advantages for the use of CMCs are high-temperature operation, which would allow a high-efficiency Rankine cycle, and low activation. Limitations to their use are material costs, fabrication complexity and costs, lack of familiarity with these materials in design, and the lack of data on radiation stability at relevant temperatures and fluences. Fusion-relevant feasibility issues identified at this workshop include: hermetic and vacuum properties related to effects of matrix porosity and matrix microcracking; chemical compatibility with coolant, tritium, and breeder and multiplier materials, radiation effects on compatibility; radiation stability and integrity; and ability to join CMCs in the shop and at the reactor site, radiation stability and integrity of joints. A summary of ongoing CMC radiation programs is also given. It was suggested that a true feasibility assessment of CMCs for fusion structural applications could not be completed without evaluation of a material tailored'' to fusion conditions or at least to radiation stability. It was suggested that a follow-up workshop be held to design a tailored composite after the results of CMC radiation studies are available and the critical feasibility issues are addressed.

  8. An autonomous long-term fast reactor system and the principal design limitations of the concept

    NASA Astrophysics Data System (ADS)

    Tsvetkova, Galina Valeryevna

    The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National

  9. A Study of Fast Reactor Fuel Transmutation in a Candidate Dispersion Fuel Design

    SciTech Connect

    Mark DeHart; Hongbin Zhang; Eric Shaber; Matthew Jesse

    2010-11-01

    Dispersion fuels represent a significant departure from typical ceramic fuels to address swelling and radiation damage in high burnup fuel. Such fuels use a manufacturing process in which fuel particles are encapsulated within a non-fuel matrix. Dispersion fuels have been studied since 1997 as part of an international effort to develop and test very high density fuel types for the Reduced Enrichment for Research and Test Reactors (RERTR) program.[1] The Idaho National Laboratory is performing research in the development of an innovative dispersion fuel concept that will meet the challenges of transuranic (TRU) transmutation by providing an integral fission gas plenum within the fuel itself, to eliminate the swelling that accompanies the irradiation of TRU. In this process, a metal TRU vector produced in a separations process is atomized into solid microspheres. The dispersion fuel process overcoats the microspheres with a mixture of resin and hollow carbon microspheres to create a TRUC. The foam may then be heated and mixed with a metal power (e.g., Zr, Ti, or Si) and resin to form a matrix metal carbide, that may be compacted and extruded into fuel elements. In this paper, we perform reactor physics calculations for a core loaded with the conceptual fuel design. We will assume a “typical” TRU vector and a reference matrix density. We will employ a fuel and core design based on the Advanced Burner Test Reactor (ABTR) design.[2] Using the CSAS6 and TRITON modules of the SCALE system [3] for preliminary scoping studies, we will demonstrate the feasibility of reactor operations. This paper will describe the results of these analyses.

  10. High temperature ceramic membrane reactors for coal liquid upgrading. Quarterly report No. 11, March 21, 1992--June 20, 1992

    SciTech Connect

    Tsotsis, T.T.

    1992-12-31

    Membrane reactors have been used in a number of catalytic reaction processes in order to overcome the limitations on conversion imposed by thermodynamic equilibrium. Having the inherent capability for combining reaction and separation in a single step, they offer promise for improving and optimizing yield, selectivity and performance of processes involving complex liquids, such as these typically found in coal liquid upgrading. Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. This project will evaluate the performance of Sol-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. Development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

  11. The cyclone: A multifunctional reactor for the fast pyrolysis of biomass

    SciTech Connect

    Lede, J.

    2000-04-01

    This paper reports the experimental results of the fast pyrolysis of wood sawdust performed in two different cyclone reactors. The mass balances are close to 100% and the char fractions always smaller than 3%. The flexibility of the cyclone reactor is such that, according to the operating conditions, it can be used either for the fast gasification or for the gas liquefaction of biomass. Side experiments reveal that a fraction of the gaseous products can be used as the carrier gas (recycling process) without noticeable changes of the gas composition and with vast gasification yields close to 100%. It is shown that the vapor-phase cracking reactions mainly occur inside a very thin and hot boundary layer close to the heated surface of the cyclone. The results of the modeling of these phenomena are used to derive kinetic constants that provide to be in very good agreement with those of the literature. The conclusion is that the cyclone appears as a very efficient multifunctional reactor making it possible to perform in less than a second heating and pyrolysis of the reactants as well as the quenching and separation of the products.

  12. Run - Beyond - Cladding - Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

    SciTech Connect

    Batte, G.L. ); Hoffman, G.L. )

    1990-01-01

    In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab.

  13. A CFD M&S PROCESS FOR FAST REACTOR FUEL ASSEMBLIES

    SciTech Connect

    Kurt D. Hamman; Ray A. Berry

    2008-09-01

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-e and SST (Menter) k-? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  14. A 100 MWe advanced sodium-cooled fast reactor core concept

    SciTech Connect

    Kim, T. K.; Grandy, C.; Hill, R. N.

    2012-07-01

    An Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developed targeting a small electrical grid to be transportable to the plant site and operable for a long time without frequent refueling. The reactor power rating was strategically decided to be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability. The design parameters were determined by relaxing the peak fast fluence limit and bulk coolant outlet temperature to beyond irradiation experience assuming that advanced cladding and structural materials developed under US-DOE programs would be available when the AFR-100 is deployed. With a de-rated power density and U-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling. The average discharge burnup of 101 MWd/kg is comparable to conventional design values, but the peak discharge fast fluence of {approx}6x10{sup 23} neutrons/cm{sup 2} is beyond the current irradiation experiences with HT-9 cladding. The evaluated reactivity coefficients provide sufficient negative feedbacks and the reactivity control systems provide sufficient shutdown margins. The integral reactivity parameters obtained from quasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficient conditions for acceptable asymptotic core outlet temperature following postulated unprotected accidents. Additionally, the AFR-100 has sufficient thermal margins by grouping the fuel assemblies into eight orifice zones. (authors)

  15. Destruction of hazardous air pollutants using a fast rise time pulsed corona reactor

    SciTech Connect

    Korzekwa, R.A.; Grothaus, M.G.; Hutcherson, R.K.; Roush, R.A.; Brown, R.

    1998-04-01

    Increasingly stringent environmental regulation imposed on both the military and civilian sectors has created a growing demand for alternative abatement methods for a variety of hazardous compounds. One alternative, the nonthermal plasma, shows promise of providing an efficient means for the destruction of dilute concentrations of hazardous air pollutants. The Dahlgren Laboratory of the Naval Surface Warfare Center has extensively investigated one type of nonthermal plasma discharge, the pulsed corona reactor, for the destruction of volatile organic compounds and chemical warfare agents. In this reactor, a fast rise time ({approximately}10ns), short duration ({lt}100ns), high-voltage pulse is repetitively delivered to a wire-cylinder electrode geometry, thereby producing a multitude of streamer discharges along its length. The resulting nonthermal plasma contains highly reactive chemical radicals which can interact with and destroy the hazardous molecules entrained in the ambient atmosphere flowing through the reactor volume. Increased electrical efficiency was obtained using a combination of high efficiency constant-current capacitor-charging, high repetition-rate spark gap switching, and resonant energy transfer to the reactor. Promising results have been obtained for toluene, methylene chloride, and dichlorodifluoromethane in air at concentrations of a few hundred parts per million. The device has been operated at voltages up to 30 kV, pulse repetition rates up to 1.4 kHz, and flow rates up to 60 scr(l)/min. Detailed electrical measurements have been made to properly characterize the electrical properties of the pulsed corona reactor and to validate subsequent improvements in the reactor energy efficiency. {copyright} {ital 1998 American Institute of Physics.}

  16. Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

    SciTech Connect

    Cheng, Lap Y.; Ludewig, Hans; Jo, Jae

    2006-07-01

    A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400 MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645 m{sup 2}) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800 kPa. (authors)

  17. Performance of low smeared density sodium-cooled fast reactor metal fuel

    SciTech Connect

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-06-17

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  18. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    SciTech Connect

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Asiah, Nur; Shafii, M. Ali; Khairurrijal

    2010-12-23

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (k{sub eff}) is in almost linear relations with the change of the fuel volume to coolant ratio.

  19. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

  20. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    SciTech Connect

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

  1. Performance of low smeared density sodium-cooled fast reactor metal fuel

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  2. Performance of Low Smeared Density Sodium-cooled Fast Reactor Metal Fuel

    SciTech Connect

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  3. Development of reactivity feedback effect measurement techniques under sub-critical condition in fast reactors

    SciTech Connect

    Kitano, A.; Nishi, H.; Suzuki, T.; Okajima, S.; Kanemoto, S.

    2012-07-01

    The first-of-a-kind reactor has been licensed by a safety examination of the plant design based on the measured data in precedent mock-up experiments. The validity of the safety design can be confirmed without a mock-up experiment, if the reactor feed-back characteristics can be measured before operation, with the constructed reactor itself. The 'Synthesis Method', a systematic and sophisticated method of sub-criticality measurement, is proposed in this work to ensure the safety margin before operation. The 'Synthesis Method' is based on the modified source multiplication method (MSM) combined with the noise analysis method to measure the reference sub-criticality level for MSM. A numerical simulation for the control-rod reactivity worth and the isothermal feed-back reactivity was conducted for typical fast reactors of 100 MWe-size, 300 MWe-size, 750 MWe-size, and 1500 MWe-size to investigate the applicability of Synthesis Method. The number of neutron detectors and their positions necessary for the measurement were investigated for both methods of MSM and the noise analysis by a series of parametric survey calculations. As a result, it was suggested that a neutron detector located above the core center and three or more neutron detectors located above the radial blanket region enable the measurement of sub-criticality within 10% uncertainty from -$0.5 to -$2 and within 15% uncertainty for the deeper sub-criticality. (authors)

  4. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    SciTech Connect

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  5. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-01

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  6. GRS Method for Uncertainties Evaluation of Parameters in a Prospective Fast Reactor

    NASA Astrophysics Data System (ADS)

    Peregudov, A.; Andrianova, O.; Raskach, K.; Tsibulya, A.

    2014-04-01

    A number of recent studies have been devoted to the uncertainty estimation of reactor calculation parameters by the GRS (Generation Random Sampled) method. This method is based on direct sampling input data resulting in formation of random sets of input parameters which are used for multiple calculations. Once these calculations are performed, statistical processing of the calculation results is carried out to determine the mean value and the variance of each calculation parameter of interest. In our study this method is used to estimate the uncertainty of calculation parameters (keff, power density, dose rate) of a prospective sodium-cooled fast reactor. Neutron transport calculations were performed by the nodal diffusion code TRIGEX and Monte Carlo code MMK.

  7. Impacts of Heterogeneous Recycle in Fast Reactors on Overall Fuel Cycle

    SciTech Connect

    Temitope A. Taiwo; Samuel E. Bays; Abdullatif M. Yacout; Edward M. Hoffman; Michael Todosow; Taek K. Kim; Massimo Salvatores

    2011-03-01

    A study in the United States has evaluated the attributes of the heterogeneous recycle approach for plutonium and minor actinide transmutation in fast reactor fuel cycles, with comparison to the homogeneous recycle approach, where pertinent. The work investigated the characteristics, advantages, and disadvantages of the approach in the overall fuel cycle, including reactor transmutation, systems and safety impacts, fuel separation and fabrication issues, and proliferation risk and transportation impacts. For this evaluation, data from previous and ongoing national studies on heterogeneous recycle were reviewed and synthesized. Where useful, information from international sources was included in the findings. The intent of the work was to provide a comprehensive assessment of the heterogeneous recycle approach at the current time.

  8. 2400MWt GAS-COOLED FAST REACTOR DHR STUDIES STATUS UPDATE.

    SciTech Connect

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01

    A topical report on demonstrating the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor was published in March 2006. The analysis was performed with the system code RELAP5-3D (version 2.4.1.1a) and the model included the full complement of the power conversion unit (PCU): heat exchange components (recuperator, precooler, intercooler) and rotating machines (turbine, compressor). A re-analysis of the success case in Ref is presented in this report. The case was redone to correct unexpected changes in core heat structure temperatures when the PCU model was first integrated with the reactor model as documented in Ref [1]. Additional information on the modeling of the power conversion unit and the layout of the heat exchange components is provided in Appendix A.

  9. Fabrication of UO/sub 2/-stainless steel fuel for a fast-reactor safety test

    SciTech Connect

    Rhude, H.V.; O'Keefe, G.B.; Noland, R.A.

    1983-01-01

    As part of its fast breeder reactor (LMFBR) safety research program, Argonne National Laboratory is conducting a series of tests in the TREAT Reactor to clarify the boilup and freezing behavior of UO/sub 2/ fuel and stainless steel in the transition phase of an accident sequence. Although much work has been done on UO/sub 2/ dispersed in a stainless steel matrix, no references could be found on work relating to the dispersal of stainless steel in a UO/sub 2/ matrix. Consequently a development program was started to produce the necessary fuel. The major difficulty encountered is attributable to the melting range of 316 stainless steel (1391 to 1399/sup 0/C) which is well below the normal sintering temperature of UO/sub 2/ (1750/sup 0/C). Attempts to produce the required fuel pellets by standard sintering techniques did not work. Alternative approaches to sintering were investigated to determine the density achievable without sintering.

  10. Thermal-hydraulic investigations of the European Fast Reactor DHR (decay heat removal) system

    SciTech Connect

    Dueweke, M.; Friedel, G.; Friedrich, H.J. ); Azarian, G. ); Thomasson, R.K. )

    1989-11-01

    With the framework of the European Fast Reactor (EFR) program, a 2-yr conceptual design study was launched in spring 1988. One major area of investigation is the decay heat removal (DHR) following a reactor trip, when the steam plant heat sink is unavailable. Decay heat will be removed from the primary sodium by a safety-grade direct rector cooling system (DRCS), which should be as independent from the emergency power supply as possible. At present, the DRCS for EFR features three identical sodium loops, each with a 30-MW rating, operating in natural circulation under all circumstances. Each loop consists of a sodium/sodium heat-exchanging U-tube dip cooler and a sodium/air heat exchanger (AHX). The thermohydraulic behavior was studied with the one-dimensional system code DYANA and with the two-dimensional thermohydraulic code ATTICA.

  11. Generation IV nuclear energy system initiative. Pin core subassembly designfor the Gas-Cooled Fast Reactor.

    SciTech Connect

    Farmer, M. T.; Hoffman, E. A.; Pfeiffer, P. F.; Therios, I. U.

    2006-07-31

    The Gas-Cooled Fast Reactor (GFR) is one of six systems selected for viability assessment in the Generation IV program. It features a closed nuclear fuel cycle, consisting of a high-temperature helium-cooled fast spectrum reactor, coupled to a direct-cycle helium turbine for electricity production. The GFR combines the advances of fast spectrum systems with those of high-temperature systems. It was clear from the very beginning that GFR design should be driven by the objective to offer a complementary approach to liquid metal cooling. On this basis, CEA and the US DOE decided to collaborate on the pre-conceptual design of a GFR. This reactor design will provide a high level of safety and full recycling of the actinides, and will also be highly proliferation resistant and economically attractive. The status of this collaborative project is that two unit sizes, 600 MWt and 2400 MWt were selected as the focus of the design and safety studies. Researchers studied fuel forms, fuel assembly/element designs, core configurations, primary and balance-of-plant layouts, and safety approaches for both of these unit sizes. Results regarding the feasibility of this GFR design are encouraging. For example, sustainability and non-proliferation goals can be met and the proposed concept has attractive safety features. These features take advantage of the helium in terms of its neutronic quasi-transparency as well as the enhanced Doppler effect in connection with candidate fuel and structural materials. The current design trend is to consider high unit power for the GFR (2400 MWt), an attractive level for the power density (100 MW/m{sup 3}), and the implementation of an innovative plate type fuel or pin type sub-assembly with carbide-based actinide compounds and SiC-based structural materials. Work is still needed to refine the safety approach, to select the main system options, and to more definitively establish economic parameters.

  12. Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems: Annual and Final Report

    SciTech Connect

    Kevan D. Weaver; Theron Marshall; James Parry

    2005-10-01

    The Idaho National Laboratory (INL) contribution to the Nuclear Energy Research Initiative (NERI) project number 2002-005 was divided into reactor physics, and thermal-hydraulics and plant design. The research targeted credible physics and thermal-hydraulics models for a gas-cooled fast reactor, analyzing various fuel and in-core fuel cycle options to achieve a true breed and burn core, and performing a design basis Loss of Coolant Accident (LOCA) analysis on that design. For the physics analysis, a 1/8 core model was created using different enrichments and simulated equilibrium fuel loadings. The model was used to locate the hot spot of the reactor, and the peak to average energy deposition at that location. The model was also used to create contour plots of the flux and energy deposition over the volume of the reactor. The eigenvalue over time was evaluated using three different fuel configurations with the same core geometry. The breeding capabilities of this configuration were excellent for a 7% U-235 model and good in both a plutonium model and a 14% U-235 model. Changing the fuel composition from the Pu fuel which provided about 78% U-238 for breeding to the 14% U-235 fuel with about 86% U-238 slowed the rate of decrease in the eigenvalue a noticeable amount. Switching to the 7% U-235 fuel with about 93% U-238 showed an increase in the eigenvalue over time. For the thermal-hydraulic analysis, the reactor design used was the one forwarded by the MIT team. This reactor design uses helium coolant, a Brayton cycle, and has a thermal power of 600 MW. The core design parameters were supplied by MIT; however, the other key reactor components that were necessary for a plausible simulation of a LOCA were not defined. The thermal-hydraulic and plant design research concentrated on determining reasonable values for those undefined components. The LOCA simulation was intended to provide insights on the influence of the Reactor Cavity Cooling System (RCCS), the

  13. U.S. Plans for the Next Fast Reactor Transmutation Fuels Irradiation Test

    SciTech Connect

    B. A. Hilton

    2007-09-01

    The U.S. Advanced Fuel Cycle Initiative (AFCI) seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposal and the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository. One important component of the technology development is actinide-bearing transmutation fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. Metallic alloy and oxide fuel forms are being developed as the near term options for fast reactor implementation.

  14. Microstructure analysis for chemical interaction between cesium and SUS 316 steel in fast breeder reactor application

    SciTech Connect

    Sasaki, K.; Fukumoto, K. I.; Oshima, T.; Tanigaki, T.; Masayoshi, U.

    2012-07-01

    In this study the corrosion products on a surface after cesium corrosion examination at 650 deg. C for 100 hrs were characterized by TEM observation around the corroded area on the surface in order to understand the corrosion mechanism of cesium fission product for cladding materials in fast reactor. The experimental results suggest the main corrosion mechanism occurred in the process of the separation of cesium chromate and metal (Fe, Ni). The main reaction of corrosion process was considered to be equation, 2Cs + 7/2 O{sub 2} + 2Cr {yields} Cs{sub 2}Cr{sub 2}O{sub 7}(L). (authors)

  15. Nuclear Data for Innovative Fast Reactors: Impact of Uncertainties and New Requirements

    SciTech Connect

    G. Palmiotti; M. Salvatores; M. Assawaroongruengchot

    2009-12-01

    It is widely accepted that the current status of neutronics calculations for fast reactor design is such that present uncertainties on nuclear data should still be significantly reduced, in order to get full benefit from advances in modeling and simulation. Only a parallel effort in advanced simulation, in high accuracy validation experiments and in nuclear data improvement will provide designers with more general and well validated calculation tools to meet tight design target accuracies to further improve safety and economics. The present paper presents very recent results related to nuclear data uncertainty impact assessment, as a new step in the frame of an international activity, sponsored by OECD-NEA.

  16. 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.

  17. Cold Trap Dismantling and Sodium Removal at a Fast Breeder Reactor - 12327

    SciTech Connect

    Graf, A.; Petrick, H.; Stutz, U.; Hosking, P.

    2012-07-01

    The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. As this reactor type used sodium as a coolant in its primary and secondary circuit, seven cold traps containing various amounts of partially activated sodium needed to be disposed of as part of the dismantling. The resulting combined difficulties of radioactive contamination and high chemical reactivity were handled by treating the cold traps differently depending on their size and the amount of sodium contained inside. Six small cold traps were processed onsite by cutting them up into small parts using a band saw under a protective atmosphere. The sodium was then converted to sodium hydroxide by using water. The remaining large cold trap could not be handled in the same way due to its dimensions (2.9 m x 1.1 m) and the declared amount of sodium inside (1,700 kg). It was therefore manually dismantled inside a large box filled with a protective atmosphere, while the resulting pieces were packaged for later burning in a special facility. The experiences gained by KNK during this process may be advantageous for future dismantling projects in similar sodium-cooled reactors worldwide. The dismantling of a prototype fast breeder reactor provides the challenge not only to dismantle radioactive materials but also to handle sodium-contaminated or sodium-containing components. The treatment of sodium requires additional equipment and installations to ensure a safe handling. Since it is not permitted to bring sodium into a repository, all sodium has to be neutralized either through a controlled reaction with water or by incinerating. The resulting components can be disposed of as normal radioactive waste with no further conditions. The handling of sodium needs skilled and experienced workers to minimize the inherent risks. And the example of the disposal of the large KNK cold trap shows the interaction with others and

  18. Decay heat of sodium fast reactor: Comparison of experimental measurements on the PHENIX reactor with calculations performed with the French DARWIN package

    SciTech Connect

    Benoit, J. C.; Bourdot, P.; Eschbach, R.; Boucher, L.; Pascal, V.; Fontaine, B.; Martin, L.; Serot, O.

    2012-07-01

    A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)

  19. Ultra-fast and energy-efficient sintering of ceramics by electric current concentration

    NASA Astrophysics Data System (ADS)

    Zapata-Solvas, E.; Gómez-García, D.; Domínguez-Rodríguez, A.; Todd, R. I.

    2015-02-01

    Electric current activated/assisted sintering (ECAS) techniques, such as electrical discharge sintering (EDS) or resistive sintering (RS), have been intensively investigated for longer than 50 years. In this work, a novel system including an electrically insulated graphite die for Spark Plasma Sintering (SPS) is described, which allows the sintering of any refractory ceramic material in less than 1 minute starting from room temperature with heating rates higher than 2000°C/min and an energy consumption up to 100 times lower than with SPS. The system alternates or combines direct resistive sintering (DRS) and indirect resistive sintering (IRS). Electrical insulation of the die has been achieved through the insertion of a film made of alumina fibers between the graphite die and the graphite punches, which are protected from the alumina fiber film by a graphite foil. This system localized the electric current directly through the sample (conductive materials) as in DRS and EDS, or through the thin graphite foil (non-conductive materials) as in IRS, and is the first system capable of being used under EDS or RS conditions independently combining current concentration/localization phenomena.

  20. Ultra-fast and energy-efficient sintering of ceramics by electric current concentration

    PubMed Central

    Zapata-Solvas, E.; Gómez-García, D.; Domínguez-Rodríguez, A.; Todd, R. I.

    2015-01-01

    Electric current activated/assisted sintering (ECAS) techniques, such as electrical discharge sintering (EDS) or resistive sintering (RS), have been intensively investigated for longer than 50 years. In this work, a novel system including an electrically insulated graphite die for Spark Plasma Sintering (SPS) is described, which allows the sintering of any refractory ceramic material in less than 1 minute starting from room temperature with heating rates higher than 2000°C/min and an energy consumption up to 100 times lower than with SPS. The system alternates or combines direct resistive sintering (DRS) and indirect resistive sintering (IRS). Electrical insulation of the die has been achieved through the insertion of a film made of alumina fibers between the graphite die and the graphite punches, which are protected from the alumina fiber film by a graphite foil. This system localized the electric current directly through the sample (conductive materials) as in DRS and EDS, or through the thin graphite foil (non-conductive materials) as in IRS, and is the first system capable of being used under EDS or RS conditions independently combining current concentration/localization phenomena. PMID:25686537

  1. An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors

    SciTech Connect

    Gentry, Cole A; Godfrey, Andrew T; Terrani, Kurt A; Gehin, Jess C; Powers, Jeffrey J; Maldonado, G Ivan

    2014-01-01

    An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

  2. Feasibility of a Monte Carlo-deterministic hybrid method for fast reactor analysis

    SciTech Connect

    Heo, W.; Kim, W.; Kim, Y.; Yun, S.

    2013-07-01

    A Monte Carlo and deterministic hybrid method is investigated for the analysis of fast reactors in this paper. Effective multi-group cross sections data are generated using a collision estimator in the MCNP5. A high order Legendre scattering cross section data generation module was added into the MCNP5 code. Both cross section data generated from MCNP5 and TRANSX/TWODANT using the homogeneous core model were compared, and were applied to DIF3D code for fast reactor core analysis of a 300 MWe SFR TRU burner core. For this analysis, 9 groups macroscopic-wise data was used. In this paper, a hybrid calculation MCNP5/DIF3D was used to analyze the core model. The cross section data was generated using MCNP5. The k{sub eff} and core power distribution were calculated using the 54 triangle FDM code DIF3D. A whole core calculation of the heterogeneous core model using the MCNP5 was selected as a reference. In terms of the k{sub eff}, 9-group MCNP5/DIF3D has a discrepancy of -154 pcm from the reference solution, 9-group TRANSX/TWODANT/DIF3D analysis gives -1070 pcm discrepancy. (authors)

  3. Feasibility study of fuel cladding performance for application in ultra-long cycle fast reactor

    NASA Astrophysics Data System (ADS)

    Jung, Ju Ang; Kim, Seung Hyun; Shin, Sang Hun; Bang, In Cheol; Kim, Ji Hyun

    2013-09-01

    As a part of the research and development activities for long-life core sodium-cooled fast reactors, the cladding performance of the ultra-long cycle fast reactor (UCFR) is evaluated with two design power levels (1000 MWe and 100 MWe) and cladding peak temperatures (873 K and 923 K). The key design concept of the UCFR is that it is non-refueling during its 30-60 years of operation. This concept may require a maximum peak cladding temperature of 923 K and a cladding radiation damage of over 200 dpa (displacements per atom). Therefore, for the design of the UCFR, deformation due to thermal creep, irradiation creep, and swelling must be taken into consideration through quantitative evaluations. As candidate cladding materials for use in UCFRs, ferritic-martensitic (FM) steels, oxide dispersion strengthened (ODS) steels, and SiC-based composite materials are studied using deformation behavior modeling for a feasibility evaluation. The results of this study indicate that SiC is a potential UCFR cladding material, with the exception of irradiation creep due to high neutron fluence stemming from its long operating time of about 30-60 years.

  4. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    NASA Astrophysics Data System (ADS)

    Anuar, Nuraslinda; Muhamad Pauzi, Anas

    2016-01-01

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the βmin is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the βmin, resulting in a list of candidate designs that possess the β value that is larger than the βmin. The proposed methodology can also be applied to purposes other than technological foresight.

  5. Adaptive polynomial chaos techniques for uncertainty quantification of a gas cooled fast reactor transient

    SciTech Connect

    Perko, Z.; Gilli, L.; Lathouwers, D.; Kloosterman, J. L.

    2013-07-01

    Uncertainty quantification plays an increasingly important role in the nuclear community, especially with the rise of Best Estimate Plus Uncertainty methodologies. Sensitivity analysis, surrogate models, Monte Carlo sampling and several other techniques can be used to propagate input uncertainties. In recent years however polynomial chaos expansion has become a popular alternative providing high accuracy at affordable computational cost. This paper presents such polynomial chaos (PC) methods using adaptive sparse grids and adaptive basis set construction, together with an application to a Gas Cooled Fast Reactor transient. Comparison is made between a new sparse grid algorithm and the traditionally used technique proposed by Gerstner. An adaptive basis construction method is also introduced and is proved to be advantageous both from an accuracy and a computational point of view. As a demonstration the uncertainty quantification of a 50% loss of flow transient in the GFR2400 Gas Cooled Fast Reactor design was performed using the CATHARE code system. The results are compared to direct Monte Carlo sampling and show the superior convergence and high accuracy of the polynomial chaos expansion. Since PC techniques are easy to implement, they can offer an attractive alternative to traditional techniques for the uncertainty quantification of large scale problems. (authors)

  6. Plant System Design of Supercritical CO{sub 2} Direct Cycle Gas Turbine Fast Reactor

    SciTech Connect

    Katsuhiro, Tozawa; Nobumasa, Tsuji; Yasushi, Muto; Yasuyoshi, Kato

    2006-07-01

    The conceptual plant design and preliminary safety analysis of SCDFR, Supercritical CO{sub 2} Direct Cycle Gas Turbine Fast Reactor, were performed. Plant thermal power is 600 MW. Core outlet/inlet pressure and temperature are 12.5/12.8 MPa and 527/388 deg C respectively. The core height and equivalent diameter are about 1.2 m and about 3.146 m respectively. The core can be burning for 10 years without refueling by adding 6.5% content of {sup 237}Np into the fuel as a burnable poison. Reactor pressure vessel height and inner diameter are about 19.3 m and about 6.55 m respectively. Steel containment vessel contains the reactor system and the gas turbine system. Preliminary analysis of core temperature behavior during the depressurization accident in SCDFR was performed. In the result of the analysis, core temperature is limited under 900 deg C, assumed limit temperature of the fuel clad, at the condition of minimum gas circulation flow rate of 2.0 m{sup 3}/s. On the other hand, gas circulator designed flow rate of the auxiliary core cooling system is over 11.6 m{sup 3}/s. These show that the integrity of the fuel clad during depressurization accident is maintained. We conclude that the plant concept of SCDFR is developed and the plant safety under depressurization accident conditions is confirmed by preliminary analysis. (authors)

  7. Optimization of a heterogeneous fast breeder reactor core with improved behavior during unprotected transients

    SciTech Connect

    Poumerouly, S.; Schmitt, D.; Massara, S.; Maliverney, B.

    2012-07-01

    Innovative Sodium-cooled Fast Reactors (SFRs) are currently being investigated by CEA, AREVA and EDF in the framework of a joint French collaboration, and the construction of a GEN IV prototype, ASTRID (Advanced Sodium Technical Reactor for Industrial Demonstration), is scheduled in the years 2020. Significant improvements are expected so as to improve the reactor safety: the goal is to achieve a robust safety demonstration of the mastering of the consequences of a Core Disruptive Accident (CDA), whether by means of prevention or mitigation features. In this framework, an innovative design was proposed by CEA in 2010. It aims at strongly reducing the sodium void effect, thereby improving the core behavior during unprotected loss of coolant transients. This design is strongly heterogeneous and includes, amongst others, a fertile plate, a sodium plenum associated with a B{sub 4}C upper blanket and a stepwise modulation of the fissile height of the core (onwards referred to as the 'diabolo shape'). In this paper, studies which were entirely carried out at EDF are presented: the full potential of this heterogeneous concept is thoroughly investigated using the SDDS methodology. (authors)

  8. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    SciTech Connect

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  9. Performance of low smeared density sodium-cooled fast reactor metal fuel

    DOE PAGESBeta

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-06-17

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactormore » designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.« less

  10. A CFD model for biomass fast pyrolysis in fluidized-bed reactors

    NASA Astrophysics Data System (ADS)

    Xue, Qingluan; Heindel, T. J.; Fox, R. O.

    2010-11-01

    A numerical study is conducted to evaluate the performance and optimal operating conditions of fluidized-bed reactors for fast pyrolysis of biomass to bio-oil. A comprehensive CFD model, coupling a pyrolysis kinetic model with a detailed hydrodynamics model, is developed. A lumped kinetic model is applied to describe the pyrolysis of biomass particles. Variable particle porosity is used to account for the evolution of particle physical properties. The kinetic scheme includes primary decomposition and secondary cracking of tar. Biomass is composed of reference components: cellulose, hemicellulose, and lignin. Products are categorized into groups: gaseous, tar vapor, and solid char. The particle kinetic processes and their interaction with the reactive gas phase are modeled with a multi-fluid model derived from the kinetic theory of granular flow. The gas, sand and biomass constitute three continuum phases coupled by the interphase source terms. The model is applied to investigate the effect of operating conditions on the tar yield in a fluidized-bed reactor. The influence of various parameters on tar yield, including operating temperature and others are investigated. Predicted optimal conditions for tar yield and scale-up of the reactor are discussed.

  11. Passive shutdown device for gas cooled fast reactor: Lithium injection module

    SciTech Connect

    Van Rooijen, W. F. G.; Kloosterman, J. L.; Van Der Hagen, T. H. J. J.; Van Dam, H.

    2006-07-01

    In this paper a passive reactivity control system for a Gas Cooled Fast Reactor is proposed. The Generation IV GCFR features a core with a relatively high power density, and control of transients and adequate shutdown under accidental situations must be assured for safe operation. Using a passive shutdown device rules out the possibility of unprotected transients. The proposed devices work by the passive introduction of {sup 6}Li into the core (Lithium Injection Module). Control is by the outlet temperature of the coolant gas in the fuel assemblies, employing a freeze seal. The proposed devices can be integrated into the regular control assemblies. A total of four LIMs is proposed in the core. Thermohydraulic calculations were done using the CATHARE code for a 600 MWth GCFR, for 2 types of transients: a loss of flow, and a control rod withdrawal. The calculations show that activation of one LIM is sufficient to keep the reactor power bounded, while activation of all LIMs in the core will shut down the reactor. The passive LIM devices are able to exclude unprotected transients in the GCFR core. (authors)

  12. Plutonium disposition in the BN-600 fast-neutron reactor at the Beloyarsk nuclear power plant

    NASA Astrophysics Data System (ADS)

    Moses, D. L.; Chebeskov, A. N.; Matveev, V. I.; Vasiliev, B. A.; Maltsev, V. V.

    In 1996, the United States and the Russian Federation completed an initial joint study that evaluated the candidate options for the disposition of surplus weapons-derived plutonium in both countries. While Russia advocates building new reactors for converting weapons-derived plutonium to spent fuel, the cost is high, and the continuing joint study of the Russian options is considering only the use of the existing VVER-1000 LWRs in Russia (and possibly in Ukraine) and the existing BN-600 fast-neutron reactor at the Beloyarsk Nuclear Power Plant in Russia. The BN-600 reactor, which currently uses enriched uranium fuel, is capable with certain design modifications of converting up to 1.3 metric tons (MT) of surplus weapons-derived plutonium to spent fuel each year. The steps needed to convert BN-600 to a plutonium-burner core will be discussed. The step involving the hybrid core allows an early and timely start that takes advantage of the limited capacity for fabricating uranium-plutonium mixed-oxide fuel early in the disposition program. The design lifetime of BN-600 must safely and reliably be extended by 10 yr to at least 2020 so that a sufficient amount of plutonium (˜20 MT) can be converted to spent fuel.

  13. Development of fast breeder reactor fuel reprocessing technology at the Power Reactor and Nuclear Fuel Development Corporation

    SciTech Connect

    Kawata, T.; Takeda, H.; Togashi, A.; Hayashi, S. . Tokai Works); Stradley, J.G. )

    1991-01-01

    For the past two decades, a broad range of research development (R D) programs to establish fast breeder reactor (FBR) system and its associated fuel cycle technology have been pursued by the Power Reactor and Nuclear Fuel Development Corporation (PNC). Developmental activities for FBR fuel reprocessing technology have been primarily conducted at PNC Tokai Works where many important R D facilities for nuclear fuel cycle are located. These include cold and uranium tests for process equipment development in the Engineering Demonstration Facilities (EDF)-I and II, and laboratory-scale hot tests in the Chemical Processing Facility (CPF) where fuel dissolution and solvent extraction characteristics are being investigated with irradiated FBR fuel pins whose burn-up ranges up to 100,000 MWd/t. An extensive effort has also been made at EDF-III to develop advanced remote technology which enables to increase plant availability and to decrease radiation exposures to the workers in future reprocessing plants. The PNC and the United States Department of Energy (USDOE) entered into the joint collaboration in which the US shares the R Ds to support FBR fuel reprocessing program at the PNC. Several important R Ds on advanced process equipment such as a rotary dissolver and a centrifugal contactor system are in progress in a joint effort with the Oak Ridge National Laboratory (ORNL) Consolidated Fuel Reprocessing Program (CFRP). In order to facilitate hot testing on advanced processes and equipment, the design of a new engineering-scale hot test facility is now in progress aiming at the start of hot operation in late 90's. 31 refs., 2 tabs.

  14. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor.

    PubMed

    Leipold, F; Furtula, V; Salewski, M; Bindslev, H; Korsholm, S B; Meo, F; Michelsen, P K; Moseev, D; Nielsen, S K; Stejner, M

    2009-09-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic using gyrotrons operated at 60 GHz will meet the requirements for spatially and temporally resolved measurements of the velocity distributions of confined fast alphas in ITER by evaluating the scattered radiation (CTS signal). While a receiver antenna on the low field side of the tokamak, resolving near perpendicular (to the magnetic field) velocity components, has been enabled, an additional antenna on the high field side (HFS) would enable measurements of near parallel (to the magnetic field) velocity components. A compact design solution for the proposed mirror system on the HFS is presented. The HFS CTS antenna is located behind the blankets and views the plasma through the gap between two blanket modules. The viewing gap has been modified to dimensions 30x500 mm(2) to optimize the CTS signal. A 1:1 mock-up of the HFS mirror system was built. Measurements of the beam characteristics for millimeter-waves at 60 GHz used in the mock-up agree well with the modeling. PMID:19791936

  15. A U. S. Perspective on Fast Reactor Fuel Fabrication Technology and Experience Part I: Metal Fuels and Assembly Design

    SciTech Connect

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter; Douglas C. Crawford; Mitchell K. Meyer

    2009-06-01

    This paper is Part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II and the Fast Flux Test Facility, and it also refers to the impact of development in other nations. Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated into a foundation of research and resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  16. A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Fielding, Randall S.; Porter, Douglas L.; Crawford, Douglas C.; Meyer, Mitchell K.

    2009-06-01

    This paper is part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF). Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated in a considerable amount of research that resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  17. Investigations of Alternative Steam Generator Location and Flatter Core Geometry for Lead-Cooled Fast Reactors

    SciTech Connect

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

    2006-07-01

    This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MW{sub th} power. This was compared to a similar design, but with the steam generators located in the downcomers. The transients investigated were Total-Loss-of-Power and unprotected Loss-Of-Flow. It is shown that this reactor peaks at 1041 K after 29 hours during a Total-Loss-Of-Power accident. The difference between locating the steam generators in the risers and the downcomers is insignificant for this accident type. During an unprotected Loss-Of-Flow accident at full power, the core outlet temperature stabilizes at 1010 K, which is 337 K above nominal outlet temperature. The second investigation concerns a 1426 MW{sub th} critical reactor where the influence of the core height versus the core outlet temperature is studied during an unprotected Loss-Of-Flow and Total-Loss-Of-Power accident. A pancake type core geometry of 1.0 m height and 5.8 m diameter, is compared to a compact core of 2 m height and 4.5 m diameter. Moderators, like BeO and hydrides, and their influence on safety coefficients and burnup swings are also presented. Both cores incinerate transuranics from spent LWR fuel with minor actinide fraction of 5%. We show that LFRs can be designed both to breed and burn transuranics from LWRs. It is shown that the hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. The computational fluid dynamics code STAR-CD was used for all thermal hydraulic calculations, and the MCNP and MCB for neutronics, and burn-up calculations. (authors)

  18. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    SciTech Connect

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-22

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  19. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    NASA Astrophysics Data System (ADS)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  20. RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao; Cliff Davis

    2008-06-01

    An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heat from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet

  1. Overview of recent studies related to lead-alloy-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Takahashi, Minoru; Sa, Rongyuan; Pramutadi, Asril; Yamaki-Irisawa, Eriko

    2012-06-01

    The recent progress of the studies related to lead alloy-cooled reactors (LFR) and the accelerator driven system (ADS) is summarized. The compatibility of materials with lead alloys has been clarified under steady and transient temperature conditions. Higher Cr content, Si and Al addition and Al-Fe alloy-coating improved the corrosion resistance of steels. The Al-Fe alloy-coated steel was not corroded even under high temperature transient conditions. The ceramics of SiC and Si3N4 are expected to be used as cladding material for high temperature LFR. For the analytical consideration of mass transport in lead alloy circuit, the diffusion coefficient of Ni was measured using the capillary methods. A new bubble visualization method in LBE using gamma-ray radiography was developed. The thermal interaction of lead-bismuth eutectic (LBE) and lead droplets with sub-cooled water, and the behaviors of droplet fragmentation were investigated, and the visualization of volatile liquids in high temperature liquids was achieved.

  2. Interim status report on lead-cooled fast reactor (LFR) research and development.

    SciTech Connect

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G.; Li, N.; Hosemann, P.; Zhang, J.; Bolind, A.; LLNL; LANL; Univ. of Illinois

    2008-03-31

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup

  3. Measuring Fracture Times Of Ceramics

    NASA Technical Reports Server (NTRS)

    Shlichta, Paul J.; Bister, Leo; Bickler, Donald G.

    1989-01-01

    Electrical measurements complement or replace fast cinematography. Electronic system measures microsecond time intervals between impacts of projectiles on ceramic tiles and fracture tiles. Used in research on ceramics and ceramic-based composite materials such as armor. Hardness and low density of ceramics enable them to disintegrate projectiles more efficiently than metals. Projectile approaches ceramic tile specimen. Penetrating foil squares of triggering device activate display and recording instruments. As ceramic and resistive film break oscilloscope plots increase in electrical resistance of film.

  4. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    SciTech Connect

    Not Available

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  5. Summary of experimental data for critical arrays of water moderated Fast Test Reactor fuel

    SciTech Connect

    Durst, B.M.; Bierman, S.R.; Clayton, E.D.; Mincey, J.F.; Primm, R.T. III

    1981-05-01

    A research program, funded by the Consolidated Fuel Reprocessing Program (CFRP) of Oak Ridge National Laboratory (ORNL), was initiated at Battelle Pacific Northwest Laboratory (PNL) to acquire experimental data on heterogeneous water moderated arrays of Fast Test Reactor (FTR) fuel pins. The objective of this program is to provide critical experiment data for validating calculational techniques used in criticality assessments of reprocessing equipment containing FTR-type fuels. Consequently, the experiments were designed to permit accurate definition in Monte Carlo computer codes currently used in these assessments. Square and triangular pitched lattices of fuel have been constructed under a variety of conditions covering the range from undermoderated to overmoderated arrays. Experiments were conducted composed of arrays which were water reflected, partially concrete reflected, and arrays with interspersed solid neutron absorbers. The absorbers utilized were Boral, and cadmium plates and gadolinium cylindrical rods. Data from non-CFRP sponsored subcritical experiments (previously performed at Hanford) also are included.

  6. Recovery of tritium dissolved in sodium at the steam generator of fast breeder reactor

    SciTech Connect

    Oya, Y.; Oda, T.; Tanaka, S.; Okuno, K.

    2008-07-15

    The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D{sub 2} gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 1/200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment. (authors)

  7. Impact of nuclear data on sodium-cooled fast reactor calculations

    NASA Astrophysics Data System (ADS)

    Aures, Alexander; Bostelmann, Friederike; Zwermann, Winfried; Velkov, Kiril

    2016-03-01

    Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

  8. Fast reactor safety: proceedings of the international topical meeting. Volume 2. [R

    SciTech Connect

    Not Available

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments.

  9. Model for collisional fast ion diffusion into Tokamak Fusion Test Reactor loss cone

    SciTech Connect

    Chang, C.S. |; Zweben, S.J.; Schivell, J.; Budny, R.; Scott, S.

    1994-08-01

    An analytic model is developed to estimate the classical pitch angle scattering loss of energetic fusion product ions into prompt loss orbits in a tokamak geometry. The result is applied to alpha particles produced by deutrium-tritium fusion reactions in a plasma condition relevant to Tokamak Fusion Test Reactor (TFTR). A poloidal angular distribution of collisional fast ion loss at the first wall is obtained and the numerical result from the TRANSP code is discussed. The present model includes the effect that the prompt loss boundary moves away from the slowing-down path due to reduction in banana thickness, which enables us to understand, for the first time. the dependence of the collisional loss rate on Z{sub eff}.

  10. ATWS Transients for the 2400 MWt Gas-Cooled Fast Reactor

    SciTech Connect

    Cheng,L.Y.; Ludewig, H.

    2007-08-05

    Reactivity transients have been analyzed with an updated RELAPS-3D (ver. 2.4.2) system model of the pin core design for the 2400MWt gas-cooled fast reactor (GCFR). Additional reactivity parameters were incorporated in the RELAP5 point-kinetics model to account for reactivity feedbacks due to axial and radial expansion of the core, fuel temperature changes (Doppler effect), and pressure changes (helium density changes). Three reactivity transients without scram were analyzed and the incidents were initiated respectively by reactivity ramp, loss of load, and depressurization. During the course of the analysis the turbine bypass model for the power conversion unit (PCU) was revised to enable a better utilization of forced flow cooling after the PCU is tripped. The analysis of the reactivity transients demonstrates the significant impact of the PCU on system pressure and core flow. Results from the modified turbine bypass model suggest a success path for the GCFR to mitigate reactivity transients without scram.

  11. A description of the demonstration Integral Fast Reactor fuel cycle facility.

    PubMed

    Courtney, J C; Carnes, M D; Dwight, C C; Forrester, R J

    1991-10-01

    A fuel examination facility at the Idaho National Engineering Laboratory is being converted into a facility that will electrochemically process spent fuel. This is an important step in the demonstration of the Integral Fast Reactor concept being developed by Argonne National Laboratory. Renovations are designed to bring the facility up to current health and safety and environmental standards and to support its new mission. Improvements include the addition of high-reliability earthquake hardened off-gas and electrical power systems, the upgrading of radiological instrumentation, and the incorporation of advances in contamination control. A major task is the construction of a new equipment repair and decontamination facility in the basement of the building to support operations. PMID:1917494

  12. Hybrid parallel strategy for the simulation of fast transient accidental situations at reactor scale

    NASA Astrophysics Data System (ADS)

    Faucher, V.; Galon, P.; Beccantini, A.; Crouzet, F.; Debaud, F.; Gautier, T.

    2014-06-01

    This contribution is dedicated to the latest methodological developments implemented in the fast transient dynamics software EUROPLEXUS (EPX) to simulate the mechanical response of fully coupled fluid-structure systems to accidental situations to be considered at reactor scale, among which the Loss of Coolant Accident, the Core Disruptive Accident and the Hydrogen Explosion. Time integration is explicit and the search for reference solutions within the safety framework prevents any simplification and approximations in the coupled algorithm: for instance, all kinematic constraints are dealt with using Lagrange Multipliers, yielding a complex flow chart when non-permanent constraints such as unilateral contact or immersed fluid-structure boundaries are considered. The parallel acceleration of the solution process is then achieved through a hybrid approach, based on a weighted domain decomposition for distributed memory computing and the use of the KAAPI library for self-balanced shared memory processing inside subdomains.

  13. Proliferation resistance of the fuel cycle for the Integral Fast Reactor

    SciTech Connect

    Burris, L.

    1993-09-01

    Argonne National Laboratory has developed an electrorefining pyrochemical process for recovery and recycle of metal fuel discharged from the Integral Fast Reactor (FR). This inherently low decontamination process has an overall decontamination factor of only about 100 for the plutonium metal product. As a result, all of the fuel cycle operations must be conducted in heavily shielded cells containing a high-purity argon atmosphere. The FR fuel cycle possesses high resistance to clandestine diversion or overt, state- supported removal of plutonium for nuclear weapons production because of two main factors: the highly radioactive product, which is also contaminated with heat- and neutron-producing isotopes of plutonium and other actinide elements, and the difficulty of removing material from the FR facility through the limited number of cell transfer locks without detection.

  14. Preliminary design of ultra-long cycle fast reactor employing breed-and-burn strategy

    SciTech Connect

    Tak, T. W.; Yu, H.; Kim, J. H.; Lee, D.; Kim, T. K.

    2012-07-01

    A new design of ultra-long cycle fast reactor with power rate of 1000 MWe (UCFR) has been developed based on the strategy of breed-and burn. The bottom region of the core with low enriched uranium (LEU) plays a role of igniter of the core burning and the upper natural uranium (NU) region acts as blanket for breeding. Fissile materials are bred in the blanket and the active core moves upward at a speed of 5.4 cm/year. Through the core depletion calculation using Monte Carlo code, McCARD, it is confirmed that a full power operation of 60 years without refueling is feasible. Core performance characteristics have been evaluated in terms of axial/radial power shapes, reactivity feedback coefficients, etc. This design will serve as a base model for further design study of UCFRs using LWR spent fuels in the blanket region. (authors)

  15. Safety design approach for external events in Japan sodium-cooled fast reactor

    SciTech Connect

    Yamano, H.; Kubo, S.; Tani, A.; Nishino, H.; Sakai, T.

    2012-07-01

    This paper describes a safety design approach for external events in the design study of Japan sodium-cooled fast reactor. An emphasis is introduction of a design extension external condition (DEEC). In addition to seismic design, other external events such as tsunami, strong wind, abnormal temperature, etc. were addressed in this study. From a wide variety of external events consisting of natural hazards and human-induced ones, a screening method was developed in terms of siting, consequence, frequency to select representative events. Design approaches for these events were categorized on the probabilistic, statistical and deterministic basis. External hazard conditions were considered mainly for DEECs. In the probabilistic approach, the DEECs of earthquake, tsunami and strong wind were defined as 1/10 of exceedance probability of the external design bases. The other representative DEECs were also defined based on statistical or deterministic approaches. (authors)

  16. Design and physical studies of fast reactor for bimodal space thermionic system with single-cell TFEs

    NASA Astrophysics Data System (ADS)

    Kirillov, E. Ya.; Klimov, A. V.; Ogloblin, B. G.; Radchenko, I. S.; Shumov, D. P.

    1997-01-01

    The paper presents the design studies and results of neutron-physical calculations of a fast nuclear reactor of a bimodal space thermionic system with single-cell thermionic fuel elements (TFE) designed for operation in two modes. These modes are (a) the propulsion mode making possible the system movement in outer space by the use of a reactive thrust generated by hydrogen heated in the reactor and (b) the electric power mode providing power supply to space vehicle-mounted systems with energy consumption level of 40kW(e) for a long time. The paper also discusses the problems of nuclear reactor safeguarding in an emergency.

  17. Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

    NASA Astrophysics Data System (ADS)

    Cherny, V. A.; Kochetkov, L. A.; Nevinitsa, A. I.

    2013-12-01

    The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.

  18. Thermal Hydraulic Challenges of Gas Cooled Fast Reactors with Passive Safety Features

    SciTech Connect

    Michael Pope; Jeong-Ik Lee; Pavel Hejzlar; Michael J. Driscoll

    2009-05-01

    Transient response of a Gas cooled Fast Reactor (GFR) coupled to a recompression supercritical CO2 (S-CO2) power conversion system (PCS) in a direct cycle to Loss of Coolant and Loss of Generator Load Accidents is analyzed using RELAP5-3D. A number of thermal hydraulic challenges for GFR design are pointed out as the designers strive to accommodate cooling of the high power density core of a fast reactor by a gas with its inherently low heat transfer capability, in particular under post LOCA events when system pressure is lost and when reliance on passive decay heat removal is emphasized. Although it is possible to design a S-CO2 cooled GFR that can survive LOCA by cooling the core through natural circulating loops between the core and elevated emergency cooling heat exchangers, it is not an attractive approach because of various bypass paths that can, depending on break location, degrade core cooling. Moreover, natural circulation gas loops can operate in deteriorated heat transfer regimes with substantial reduction of heat transfer coefficient: as low as 30% of forced convection values, and data and correlations in these regimes carry large uncertainties. Therefore, reliable battery powered blowers for post-LOCA decay heat removal (DHR) that provide flow in well defined regimes with low uncertainty, and can be easily over-designed to accommodate bypass flows were selected. The results confirm that a GFR with such a DHR system and negative coolant void worth can withstand LOCA with and without scram as well as loss of electrical load without exceeding core temperature and turbomachinery overspeed limits.

  19. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    NASA Astrophysics Data System (ADS)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-10-01

    The study evaluates the possible use of graphite foam as the bonding material between U-Pu-Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U-15Pu-6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600-660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

  20. The technical and economic impact of minor actinide transmutation in a sodium fast reactor

    SciTech Connect

    Gautier, G. M.; Morin, F.; Dechelette, F.; Sanseigne, E.; Chabert, C.

    2012-07-01

    Within the frame work of the French National Act of June 28, 2006 pertaining to the management of high activity, long-lived radioactive waste, one of the proposed processes consists in transmuting the Minor Actinides (MA) in the radial blankets of a Sodium Fast Reactor (SFR). With this option, we may assess the additional cost of the reactor by comparing two SFR designs, one with no Minor Actinides, and the other involving their transmutation. To perform this exercise, we define a reference design called SFRref, of 1500 MWe that is considered to be representative of the Reactor System. The SFRref mainly features a pool architecture with three pumps, six loops with one steam generator per loop. The reference core is the V2B core that was defined by the CEA a few years ago for the Reactor System. This architecture is designed to meet current safety requirements. In the case of transmutation, for this exercise we consider that the fertile blanket is replaced by two rows of assemblies having either 20% of Minor Actinides or 20% of Americium. The assessment work is performed in two phases. - The first consists in identifying and quantifying the technical differences between the two designs: the reference design without Minor Actinides and the design with Minor Actinides. The main differences are located in the reactor vessel, in the fuel handling system and in the intermediate storage area for spent fuel. An assessment of the availability is also performed so that the impact of the transmutation can be known. - The second consists in making an economic appraisal of the two designs. This work is performed using the CEA's SEMER code. The economic results are shown in relative values. For a transmutation of 20% of MA in the assemblies (S/As) and a hypothesis of 4 kW allowable for the washing device, there is a large external storage demanding a very long cooling time of the S/As. In this case, the economic impact may reach 5% on the capital part of the Levelized Unit

  1. Feasibility study on ultralong-cycle operation and material performance for compact liquid metal-cooled fast reactors: a review work

    SciTech Connect

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung; Kim, T. K.; Hong, Ser Gi

    2015-11-01

    This paper reviews the feasibility of ultralong-cycle operation on a compact liquid metal-cooled fast reactor (LMR) firstly by assessing the operation of a long-life fast reactor core and secondly by evaluating material performance in respect to both long-cycle operation and compact-size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen-IV) reactor system development. The breed-and-burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super-Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long-life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long-cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research for a complete reactor model with the other reactor components.

  2. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    DOE PAGESBeta

    Cheng, Lap-Yan; Wei, Thomas Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

  3. Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

    SciTech Connect

    Ponciroli, Roberto; Passerini, Stefano; Vilim, Richard B.

    2016-01-01

    In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based on the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.

  4. Measurement control design and performance assessment in the Integral Fast Reactor fuel cycle

    SciTech Connect

    Orechwa, Y.; Bucher, R.G.

    1994-08-01

    The Integral Fast Reactor (IFR)--consisting of a metal fueled and liquid metal cooled reactor together with an attendant fuel cycle facility (FCF)--is currently undergoing a phased demonstration of the closed fuel cycle at Argonne National Laboratory. The recycle technology is pyrometalurgical based with incomplete fission product separation and all transuranics following plutonium for recycle. The equipment operates in batch mode at 500 to 1,300 C. The materials are highly radioactive and pyrophoric, thus the FCF requires remote operation. Central to the material control and accounting system for the FCF are the balances for mass measurements. The remote operation of the balances limits direct adjustment. The radiation environment requires that removal and replacement of the balances be minimized. The uniqueness of the facility precludes historical data for design and performance assessment. To assure efficient operation of the facility, the design of the measurement control system has called for procedures which assess the performance of the balances in great detail and will support capabilities for the correction of systematic changes in the performance of the balances through software.

  5. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    SciTech Connect

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  6. Initial Requirements for Gas-Cooled Fast Reactor (GFR) System Design, Performance, and Safety Analysis Models

    SciTech Connect

    Kevan D. Weaver; Thomas Y. C. Wei

    2004-08-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  7. Consolidated fuel reprocessing program: Criticality experiments with fast test reactor fuel pins in an organic moderator

    SciTech Connect

    Bierman, S.R.

    1986-12-01

    The results obtained in a series of criticality experiments performed as part of a joint program on criticality data development between the United States Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan are presented in this report along with a complete description of the experiments. The experiments involved lattices of Fast Test Reactor (FTR) fuel pins in an organic moderator mixture similar to that used in the solvent extraction stage of fuel reprocessing. The experiments are designed to provide data for direct comparison with previously performed experimental measurements with water moderated lattices of FTR fuel pins. The same lattice arrangements and FTR fuel pin types are used in these organic moderated experimental assemblies as were used in the water moderated experiments. The organic moderator is a mixture of 38 wt % tributylphosphate in a normal paraffin hydrocarbon mixture of C{sub 11}H{sub 24} to C{sub 15}H{sub 32} molecules. Critical sizes of 1054.8, 599.2, 301.8, 199.5 and 165.3 fuel pins were obtained respectively for organic moderated lattices having 0.761 cm, 0.968 cm, 1.242 cm, 1.537 cm and 1.935 cm square lattice pitches as compared to 1046.9, 571.9, 293.9, 199.7 and 165.1 fuel pins for the same lattices water moderated.

  8. Conceptual design features of the Kalimer-600 sodium cooled fast reactor

    SciTech Connect

    Hahn, Dohee; Kim, Yeong-Il; Kim, Seong-O; Lee, Jae-Han; Lee, Yong-Bum; Jeong, Hae-Yong

    2007-07-01

    An advanced sodium cooled fast reactor concept, KALIMER-600, has been developed by the Korea Atomic Energy Research Institute to satisfy the Gen-IV technology goals of sustainability, safety and reliability, economics and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on a proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design were verified through a safety analysis of its bounding events. The results for various unprotected events imply that the KALIMER-600 design can accommodate all the analyzed ATWS events. This self-regulation capability of the power without a scram is mainly attributed to the inherent reactivity feedback mechanisms implemented in the metal fuel core design and completely passive decay heat removal system. (authors)

  9. Advanced control system for the Integral Fast Reactor fuel pin processor

    SciTech Connect

    Lau, L.D.; Randall, P.F.; Benedict, R.W.; Levinskas, D.

    1993-03-01

    A computerized control system has been developed for the remotely-operated fuel pin processor used in the Integral Fast Reactor Program, Fuel Cycle Facility (FCF). The pin processor remotely shears cast EBR- reactor fuel pins to length, inspects them for diameter, straightness, length, and weight, and then inserts acceptable pins into new sodium-loaded stainless-steel fuel element jackets. Two main components comprise the control system: (1) a programmable logic controller (PLC), together with various input/output modules and associated relay ladder-logic associated computer software. The PLC system controls the remote operation of the machine as directed by the OCS, and also monitors the machine operation to make operational data available to the OCS. The OCS allows operator control of the machine, provides nearly real-time viewing of the operational data, allows on-line changes of machine operational parameters, and records the collected data for each acceptable pin on a central data archiving computer. The two main components of the control system provide the operator with various levels of control ranging from manual operation to completely automatic operation by means of a graphic touch screen interface.

  10. Advanced control system for the Integral Fast Reactor fuel pin processor

    SciTech Connect

    Lau, L.D.; Randall, P.F.; Benedict, R.W.; Levinskas, D.

    1993-01-01

    A computerized control system has been developed for the remotely-operated fuel pin processor used in the Integral Fast Reactor Program, Fuel Cycle Facility (FCF). The pin processor remotely shears cast EBR- reactor fuel pins to length, inspects them for diameter, straightness, length, and weight, and then inserts acceptable pins into new sodium-loaded stainless-steel fuel element jackets. Two main components comprise the control system: (1) a programmable logic controller (PLC), together with various input/output modules and associated relay ladder-logic associated computer software. The PLC system controls the remote operation of the machine as directed by the OCS, and also monitors the machine operation to make operational data available to the OCS. The OCS allows operator control of the machine, provides nearly real-time viewing of the operational data, allows on-line changes of machine operational parameters, and records the collected data for each acceptable pin on a central data archiving computer. The two main components of the control system provide the operator with various levels of control ranging from manual operation to completely automatic operation by means of a graphic touch screen interface.

  11. High temperature ceramic membrane reactors for coal liquid upgrading. Final report, September 21, 1989--November 20, 1992

    SciTech Connect

    Tsotsis, T.T.; Liu, P.K.T.; Webster, I.A.

    1992-12-31

    Membrane reactors are today finding extensive applications for gas and vapor phase catalytic reactions (see discussion in the introduction and recent reviews by Armor [92], Hsieh [93] and Tsotsis et al. [941]). There have not been any published reports, however, of their use in high pressure and temperature liquid-phase applications. The idea to apply membrane reactor technology to coal liquid upgrading has resulted from a series of experimental investigations by our group of petroleum and coal asphaltene transport through model membranes. Coal liquids contain polycyclic aromatic compounds, which not only present potential difficulties in upgrading, storage and coprocessing, but are also bioactive. Direct coal liquefaction is perceived today as a two-stage process, which involves a first stage of thermal (or catalytic) dissolution of coal, followed by a second stage, in which the resulting products of the first stage are catalytically upgraded. Even in the presence of hydrogen, the oil products of the second stage are thought to equilibrate with the heavier (asphaltenic and preasphaltenic) components found in the feedstream. The possibility exists for this smaller molecular fraction to recondense with the unreacted heavy components and form even heavier undesirable components like char and coke. One way to diminish these regressive reactions is to selectively remove these smaller molecular weight fractions once they are formed and prior to recondensation. This can, at least in principle, be accomplished through the use of high temperature membrane reactors, using ceramic membranes which are permselective for the desired products of the coal liquid upgrading process. An additional incentive to do so is in order to eliminate the further hydrogenation and hydrocracking of liquid products to undesirable light gases.

  12. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    SciTech Connect

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  13. Thermal-hydraulics of internally heated molten salts and application to the Molten Salt Fast Reactor

    NASA Astrophysics Data System (ADS)

    Fiorina, Carlo; Cammi, Antonio; Luzzi, Lelio; Mikityuk, Konstantin; Ninokata, Hisashi; Ricotti, Marco E.

    2014-04-01

    The Molten Salt Reactors (MSR) are an innovative kind of nuclear reactors and are presently considered in the framework of the Generation IV International Forum (GIF-IV) for their promising performances in terms of low resource utilization, waste minimization and enhanced safety. A unique feature of MSRs is that molten fluoride salts play the distinctive role of both fuel (heat source) and coolant. The presence of an internal heat generation perturbs the temperature field and consequences are to be expected on the heat transfer characteristics of the molten salts. In this paper, the problem of heat transfer for internally heated fluids in a straight circular channel is first faced on a theoretical ground. The effect of internal heat generation is demonstrated to be described by a corrective factor applied to traditional correlations for the Nusselt number. It is shown that the corrective factor can be fully characterized by making explicit the dependency on Reynolds and Prandtl numbers. On this basis, a preliminary correlation is proposed for the case of molten fluoride salts by interpolating the results provided by an analytic approach previously developed at the Politecnico di Milano. The experimental facility and the related measuring procedure for testing the proposed correlation are then presented. Finally, the developed correlation is used to carry out a parametric investigation on the effect of internal heat generation on the main out-of-core components of the Molten Salt Fast Reactor (MSFR), the reference circulating-fuel MSR design in the GIF-IV. The volumetric power determines higher temperatures at the channel wall, but the effect is significant only in case of large diameters and/or low velocities.

  14. Polonium Issue in Fast Reactor Lead Coolants and One of the Ways of Its Solution

    SciTech Connect

    Khorasanov, G.L.; Ivanov, A.P.; Blokhin, A.I.

    2002-07-01

    One of the main issues in using materials for nuclear facilities is to minimize the production of the most hazardous radionuclides. In the ideal case, all nuclear reactor materials, except a fuel, should be low-activation. The term 'low-activation material' means that this one loses its induced activity in a short time after removal from irradiation. Proposals for building a fusion reactor using low-activation materials are given in Ref.1, 2. For this purpose, low-activation structural materials based on V-Ti-Cr alloys are in the stage of R and D in several countries [3,4]. Another technique to avoid the hazardous activity is in using isotopically enriched materials [5-7]. Although isotopic tailoring option requires tremendous technical efforts and it is too expensive, its application can be first of all assumed for those structural and functional materials which generate very hazardous radionuclides under irradiation. In modern projects of next generation NPPs the preference is given to fast reactors (FRs) with a lead coolant [8]. As it known, the coolant circulating through a FR core is activated, and in the future we should have problems with handling a completed coolant after FR decommissioning or at realization of repair or emergency activities. There, it is desirable to have a low-activation coolant with the low contents of hazardous radionuclides. In papers [9,10] presented at the previous ICONE conferences it was proposed to use lead isotope, Pb-206, as a coolant instead of lead natural, Pb-nat. This paper is devoted to more detailed calculations of accumulating stable bismuth, Bi-209, and polonium radioisotopes, Po-209 (T{sub 1/2}=102 y) and Po-210 (T{sub 1/2}=138 d), in 1 kg of Pb-nat or Pb-206 placed in the core of the BOR-60 type FR. (authors)

  15. Filtered fast neutron irradiation system using Texas A&M University Nuclear Science Center Reactor

    NASA Astrophysics Data System (ADS)

    Jang, S. Y.; Kim, C. H.; Reece, W. D.; Braby, L. A.

    2004-09-01

    A heavily filtered fast neutron irradiation system (FNIS) was developed for a variety of applications, including the study of long-term health effects of fast neutrons by evaluating the biological mechanisms of damage in cultured cells and living animals such as rats or mice. This irradiation system includes an exposure cave made with a lead-bismuth alloy, a cave positioning system, a gamma and neutron monitoring system, a sample transfer system, and interchangeable filters. This system was installed in the irradiation cell of the Texas A&M University Nuclear Science Center Reactor (NSCR). For a realistic modeling of the NSCR, the irradiation cell, and the FNIS, this study used the Monte Carlo N-Particle (MCNP) code and a set of high-temperature ENDF/B-VI continuous neutron cross-section data. Sensitivity analysis was performed to find the characteristics of the FNIS as a function of the thickness of the lead-bismuth alloy. A paired ion chamber system was constructed with a tissue-equivalent plastic (A-150) and propane gas for total dose monitoring and with graphite and argon for gamma dose monitoring. This study, in addition, tested the Monte Carlo modeling of the FNIS system, as well as the performance of the system by comparing the calculated results with experimental measurements using activation foils and paired ion chambers.

  16. Anomalous fast ion losses at high β on the tokamak fusion test reactor

    SciTech Connect

    Fredrickson, E. D.; Bell, M. G.; Budny, R. V.; Darrow, D. S.; White, R.

    2015-03-15

    This paper describes experiments carried out on the Tokamak Fusion Test Reactor (TFTR) [R. J. Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] to investigate the dependence of β-limiting disruption characteristics on toroidal field strength. The hard disruptions found at the β-limit in high field plasmas were not found at low field, even for β's 50% higher than the empirical β-limit of β{sub n} ≈ 2 at high field. Comparisons of experimentally measured β's to TRANSP simulations suggest anomalous loss of up to half of the beam fast ions in the highest β, low field shots. The anomalous transport responsible for the fast ion losses may at the same time broaden the pressure profile. Toroidal Alfvén eigenmodes, fishbone instabilities, and Geodesic Acoustic Modes are investigated as possible causes of the enhanced losses. Here, we present the first observations of high frequency fishbones [F. Zonca et al., Nucl. Fusion 49, 085009 (2009)] on TFTR. The interpretation of Axi-symmetric Beam-driven Modes as Geodesic Acoustic Modes and their possible correlation with transport barrier formation are also presented.

  17. 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

  18. CHARACTERISTICS OF A FAST RISE TIME POWER SUPPLY FOR A PULSED PLASMA REACTOR FOR CHEMICAL VAPOR DESTRUCTION

    EPA Science Inventory

    Rotating spark gap devices for switching high-voltage direct current (dc) into a corona plasma reactor can achieve pulse rise times in the range of tens of nanoseconds. The fast rise times lead to vigorous plasma generation without sparking at instantaneous applied voltages highe...

  19. An electric heating control system for a nuclear power unit equipped with a fast-neutron reactor

    NASA Astrophysics Data System (ADS)

    Shmuel'Zon, M. B.; Barskii, L. A.

    2007-10-01

    An electric heating control system for a nuclear power unit equipped with a fast-neutron reactor is considered, which allows the required temperatures in the heat zones to be maintained when they are heated up and stabilized. The specific features of the controlled plant and the control equipment employed are taken into account.

  20. Laser resonance ionization mass spectrometry for failed fuel detection and location in the experimental fast reactor JOYO

    NASA Astrophysics Data System (ADS)

    Harano, Hideki; Nose, Shoichi; Ito, Kazuhiro; Watanabe, Kenichi; Iguchi, Tetsuo

    2001-08-01

    For the improvement of the failed fuel detection and location technique base on a tagging gas method, we are developing a RIMS prototype system and plan to demonstrate its performance at the experimental fast reactor JOYO. We report the basic design of the system and preliminary experimental results for trace Xe isotopic analysis.

  1. A physics study for negative void reactivity in compact supercritical CO{sub 2}-cooled fast reactor

    SciTech Connect

    Kim, Y.; Hartanto, D.; Lee, J. I.

    2013-07-01

    A compact S-CO{sub 2}-cooled fast reactor which has negative Coolant Void Reactivity (CVR) has been investigated. A negative CVR is important for the gas cooled fast reactor as an inherent safety mechanism to prevent the sudden positive reactivity insertion when the loss of coolant accident happens. An alternative solution to reduce the CVR is investigated in this study by using O-17 instead of O-16 in UO{sub 2} fuel. By using O-17 in the fuel, it is found that the CVR can even be negative. Impacts of the radial reflector on the CVR are also evaluated for the small SCO{sub 2} cooled fast reactor in this study. We have considered a pure lead (Pb) reflector and a lead magnesium eutectic (LME) reflector as alternative radial reflectors of the S-CO 2-cooled fast reactor. It has been shown that, with the LME radial reflector, the CVR can be negative, while the pure lead reflector provides a slightly positive CVR. (authors)

  2. Importance of the (n,gamma) Cm-247 Evaluation on Neutron Emission in Fast Reactor Fuel Cycle Analysis

    SciTech Connect

    Benoit Forget; Mehdi Asgari; Rodolfo M. Ferrer

    2007-11-01

    As part of the GNEP program, it is envisioned to build a fast reactor for the transmutation of minor actinides. The spent nuclear fuel from the current fleet of light water reactors would be recycled, the current baseline is the UREX+1a process, and would act as a feed for the fast reactor. As the fuel is irradiated in a fast reactor a certain quantity of minor actinides would thus build up in the fuel stream creating possible concerns with the neutron emission of these minor actinides for fuel transportation, handling and fabrication. Past neutronic analyses had not tracked minor actinides above Cm-246 in the transmutation chain, because of the small influence on the overall reactor performance and cycle parameters. However, when trying to quantify the neutron emission from the recycled fuel with high minor actinide content, these higher isotopes play an essential role and should be included in the analysis. In this paper, the influence of tracking these minor actinides on the calculated neutron emission is presented. Also presented is the particular influence of choosing a different evaluated cross section data set to represent the minor actinides above Cm-246. The first representation uses the cross-sections provided by MC2-2 for all isotopes, while the second representation uses infinitely diluted ENDF/BVII.0 cross-sections for Cm-247 to Cf-252 and MC2-2 for all other isotopes.

  3. A low-temperature co-fired ceramic micro-reactor system for high-efficiency on-site hydrogen production

    NASA Astrophysics Data System (ADS)

    Jiang, Bo; Maeder, Thomas; Santis-Alvarez, Alejandro J.; Poulikakos, Dimos; Muralt, Paul

    2015-01-01

    A ceramic-based, meso-scale fuel processor for on-board production of syngas fuel was demonstrated for applications in micro-scale solid-oxide fuel cells (μ-SOFCs). The processor had a total dimension of 12 mm × 40 mm × 2 mm, the gas reforming micro reactor occupying the hot end of a cantilever had outer dimensions of 12 × 18 mm. The device was fabricated through a novel progressive lamination process in low-temperature co-fired ceramic (LTCC) technology. Both, heating function and desired fluidic structures were integrated monolithically into the processor. Using catalytic partial oxidation of a hydrocarbon fuel (propane) as a reaction model, a thermally self-sustaining hydrogen production was achieved. The output flow is sufficiently high to drive an optimized single membrane μSOFC cell of about the same footprint as the micro reactor. Microsystem design, fabrication, catalyst integration as well as the chemical characterization are discussed in detail.

  4. Medium-Power Lead-Alloy Fast Reactor Balance-of-Plant Options

    SciTech Connect

    Dostal, Vaclav; Hejzlar, Pavel; Todreas, Neil E.; Buongiorno, Jacopo

    2004-09-15

    Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature ({approx}550 deg. C) compared to that of light water reactors ({approx}300 deg. C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO{sub 2} (S-CO{sub 2}) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545 deg. C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312 deg. C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO{sub 2} cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be

  5. Summary Report on New Transmutation Analysis for the Evaluation of Homogeneous and Heterogeneous Options in Fast Reactors

    SciTech Connect

    R. M. Ferrer; S. Bays; M. Pope; B. Forget; W. Skerjanc; M. Asgari

    2008-08-01

    A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design was selected as the baseline in this scenario study. Traditional approaches to Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) transuranic waste (TRU) burning in a fast spectrum system have typically focused on the continual homogeneous recycling (reprocessing) of the discharge fast reactor fuel. The effective reduction of transuranic inventories has been quantified through the use of the transuranics conversion ratio (TRU CR). The implicit assumption in the use of this single parameter is a homogeneous fast reactor option where equal weight is given to the destruction of transuranics, either by fission or eventual fission via transmutation. This work explores the potential application of alternative fast reactor fuel cycles in which the minor actinide (MA) component of the TRU is considered ‘waste’, while the plutonium component is considered as fuel. Specifically, a set of potential designs that incorporate radial heterogeneous target assemblies is proposed and results relevant to transmutation and system analysis are presented. In this work we consider exclusively minor actinide-bearing radial targets in a continual reprocessing scenario (as opposed to deep-burn options). The potential use of targets in a deep burn mode is not necessarily ruled out as an option. However, due to work scope constraints and material limit considerations, it was preferred to leave the target assemblies reach either the assumed limit of 200 DPA at discharge or maximum allowable gas pressure caused by helium production from transmutation. The number and specific design of the target assemblies was chosen to satisfy the necessary core symmetry and physical dimensions (available space for a certain amount of mass in an assembly based on an iterated mass density).

  6. Fast Neutron Spectrum Potassium Worth for Space Power Reactor Design Validation

    SciTech Connect

    Bess, John D.; Marshall, Margaret A.; Briggs, J. Blair; Tsiboulia, Anatoli; Rozhikhin, Yevgeniy; Mihalczo, John T.

    2015-03-01

    A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, one was performed that consisted of uranium metal annuli surrounding a potassium-filled, stainless steel can. The outer diameter of the annuli was approximately 13 inches (33.02 cm) with an inner diameter of 7 inches (17.78 cm). The diameter of the stainless steel can was 7 inches (17.78 cm). The critical height of the configurations was approximately 5.6 inches (14.224 cm). The uranium annulus consisted of multiple stacked rings, each with radial thicknesses of 1 inch (2.54 cm) and varying heights. A companion measurement was performed using empty stainless steel cans; the primary purpose of these experiments was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs.The experimental measurements were performed on July 11, 1963, by J. T. Mihalczo and M. S. Wyatt (Ref. 1) with additional information in its corresponding logbook. Unreflected and unmoderated experiments with the same set of highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in the International Handbook for Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) with the identifier HEU MET FAST 051. Thin

  7. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    SciTech Connect

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  8. 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

  9. Continuous Fiber Wound Ceramic Composite (CFCC) for Commercial Water Reactor Fuel. Technical progress report for period ending April 1, 2000

    SciTech Connect

    2000-04-01

    Our program began on August 1, 1999. As of April 1, 2000, the progress has been in materials selection and test planning. Three subcontracts are in place (McDermott Technologies Inc. for continuous fiber reinforced ceramic tubing fabrication, Swales Aerospace for LOCA testing of tubes, and Massachusetts Institute of Technology for In Reactor testing of tubes). With regard to materials selection we visited McDermott Technologies Inc. a number of times, including on February 23, 2000 to discuss the Draft Material Selection and Fabrication Report. The changes discussed at this meeting were implemented and the final version of this report is attached (attachment 1). McDermott Technologies Inc. will produce one type of tubing: Alumina oxide (Nextel 610) fiber, a carbon coating (left in place), and alumina-yttria matrix. A potentially desirable CFCC material of silicon carbide fiber with spinel matrix was discussed. That material selection was not adopted primarily due to material availability and cost. Gamma Engineering is exploring the available tube coatings at Northwestern University as a mechanism for reducing the permeability of the tubes, and thus, will use coating as a differentiating factor in the testing of tubing in the LOCA test as well as the In-Reactor Test. The conclusion of the Material Selection and Fabrication Report lists the possible coatings under evaluation. With regard to Test Planning, the MIT and Swales Aerospace have submitted draft Test Plans. MIT is attempting to accommodate an increased number of test specimens by evaluating alternative test configurations. Swales Aerospace held a design review at their facilities on February 24, 2000 and various engineering alternatives and safety issues were addressed. The final Test Plans are not expected until just before testing begins to allow for incorporation of changes during ''dry runs.''

  10. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    SciTech Connect

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  11. Data Collection Methods for Validation of Advanced Multi-Resolution Fast Reactor Simulations

    SciTech Connect

    Tokuhiro, Akiro; Ruggles, Art; Pointer, David

    2015-01-22

    In pool-type Sodium Fast Reactors (SFR) the regions most susceptible to thermal striping are the upper instrumentation structure (UIS) and the intermediate heat exchanger (IHX). This project experimentally and computationally (CFD) investigated the thermal mixing in the region exiting the reactor core to the UIS. The thermal mixing phenomenon was simulated using two vertical jets at different velocities and temperatures as prototypic of two adjacent channels out of the core. Thermal jet mixing of anticipated flows at different temperatures and velocities were investigated. Velocity profiles are measured throughout the flow region using Ultrasonic Doppler Velocimetry (UDV), and temperatures along the geometric centerline between the jets were recorded using a thermocouple array. CFD simulations, using COMSOL, were used to initially understand the flow, then to design the experimental apparatus and finally to compare simulation results and measurements characterizing the flows. The experimental results and CFD simulations show that the flow field is characterized into three regions with respective transitions, namely, convective mixing, (flow direction) transitional, and post-mixing. Both experiments and CFD simulations support this observation. For the anticipated SFR conditions the flow is momentum dominated and thus thermal mixing is limited due to the short flow length associated from the exit of the core to the bottom of the UIS. This means that there will be thermal striping at any surface where poorly mixed streams impinge; rather unless lateral mixing is ‘actively promoted out of the core, thermal striping will prevail. Furthermore we note that CFD can be considered a ‘separate effects (computational) test’ and is recommended as part of any integral analysis. To this effect, poorly mixed streams then have potential impact on the rest of the SFR design and scaling, especially placement of internal components, such as the IHX that may see poorly mixed

  12. Development and Validation of ARKAS cellule: An Advanced Core-Bowing Analysis Code for Fast Reactors

    SciTech Connect

    Ohta, Hirokazu; Yokoo, Takeshi; Nakagawa, Masatoshi; Matsuyama, Shinichiro

    2004-05-15

    An advanced analysis code, ARKAS cellule, has been developed to determine the core distortion and the mechanical behavior of fast reactors. In this code, each hexagonal subassembly duct is represented by a folded thin plate structure divided into a user-specified number of shell elements so that the interduct contact forms and the cross-sectional distortion effect of each duct are properly taken into account. In this paper, the numerical model of the ARKAS cellule code is introduced, and the analytical results for two validation problems are presented. From a single duct compaction analysis, the first validation problem, it is clarified that the new analytical model is applicable to simulating the change of duct compaction stiffness that depends on the loading conditions such as the loading pad forms and the number of contact faces. The second validation analysis has been conducted by comparison with the experimental values obtained by the National Nuclear Corporation Limited in the United Kingdom using the core restraint uniplanar experimental rig (CRUPER), an ex-reactor rig in which a cluster of 91 short ducts is compressed by 30 movable peripheral rams toward the center of the cluster in seven stages. The analysis clarified that the predictions obtained using ARKAS cellule agree well with the measured ram loads and interwrapper gap widths during the compaction sequence. One may conclude that ARKAS cellule is valid for quantitative analysis of the core mechanical behavior and will be particularly useful for the evaluation of transient deformation of core assemblies during accidents in which the distortion of loading pads have important effects on obtaining favorable reactivity feedback.

  13. Decontamination of liquid-metal fast breeder reactor components for reuse; The French experience

    SciTech Connect

    Michaille, P. ); Moroni, J.C. ); Lambert, I. )

    1991-02-01

    Decontamination of stainless steel liquid-metal fast breeder reactor components for reuse in France began with the decontamination of Rapsodie components. At that time, dilute phosphoric acid was used. To cope with additional irradiated components after Phenix came into operation, an extensive study was performed, which led to the selection of a procedure involving two baths. The first bath, alkaline permanganate (AP), is applied for 3 h; the second bath, sulfo-phosphoric acid (SP), is applied for 6 h, both at 60{degrees}C. Up to three cycles are repeated until the residual dose rate is sufficiently low. Eight intermediate heat exchangers (IHXs) and two primary pumps from Phenix were decontaminated using this method. This paper reports that because SP can pickle only a limited depth ({approximately} 3{mu}m), due to the passivation effect of phosphoric acid, and because of the waste treatment problems associated with phosphates, new solutions were explored. One possibility involves improvement of the AP-SP procedure: In the SPm procedure, the AP bath is omitted and the phosphoric concentration is reduced by a factor of 4. A second approach is the use of a new formula, called SECA, a mixture of maleic and citric acid used in reducing conditions (imposed by hydrazine). Since the Phenix and Superphenix waste treatment facilities are not designed to reprocess maleic-citric acid, only the SPm procedure has been used on reactor components. A low-contaminated IHX from Rapsodie served as a test benchmark, not only for the decontamination procedure, but also for the requalification criteria, before the SPm procedure was applied to a highly contaminated IHX from Phenix. Recent results are presented.

  14. Minor Actinide Transmutation Physics for Low Conversion Ratio Sodium Fast Reactors

    SciTech Connect

    Mehdi Asgari; Samuel E. Bays; Benoit Forget; Rodolfo Ferrer

    2007-09-01

    The effects of varying the reprocessing strategy used in the closed cycle of a Sodium Fast Reactor (SNF) prototype are presented in this paper. The isotopic vector from the aqueous separation of transuranic (TRU) elements in Light Water Reactor (LWR) spent nuclear fuel (SNF) is assumed to also vary according to the reprocessing strategy of the closed fuel cycle. The decay heat, gamma energy, and neutron emission of the fuel discharge at equilibrium are found to vary depending on the separation strategy. The SFR core used in this study corresponds to a burner configuration with a conversion ratio of ~0.5 based on the Super-PRISM design. The reprocessing strategies stemming from the choice of either metal or oxide fuel for the SFR are found to have a large impact on the equilibrium discharge decay heat, gamma energy, and neutron emission. Specifically, metal fuel SFR with pyroprocessing of the discharge produces the largest amount of TRU consumption (166 kg per Effective Full Power Year or EFPY), but also the highest decay heat, gamma energy, and neutron emission. On the other hand, an oxide fuel SFR with PUREX reprocessing minimizes the decay heat and related parameters of interest to a minimum, even when compared to thermal Mixed Oxide (MOX) or Inert Matrix Fuel (IMF) on a per mass basis. On an assembly basis, however, the metal SFR discharge has a lower decay heat than an equivalent oxide SFR assembly for similar minor actinide consumptions (~160 kg/EFPY.) Another disadvantage in the oxide PUREX reprocessing scenario is that there is no consumption of americium and curium, since PUREX reprocessing separates these minor actinides (MA) and requires them to be disposed of externally.

  15. Fabrication technological development of the oxide dispersion strengthened alloy MA957 for fast reactor applications

    SciTech Connect

    ML Hamilton; DS Gelles; RJ Lobsinger; GD Johnson; WF Brown; MM Paxton; RJ Puigh; CR Eiholzer; C Martinez; MA Blotter

    2000-03-27

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material, in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report. The alloy is a ferritic stainless steel developed by International Nickel Company specifically for structural reactor applications. It is strengthened by a very fine, uniformly distributed yttria dispersoid. Its fabrication involves a mechanical alloying process and subsequent extrusion, which ultimately results in a highly elongated grain structure. While the presence of the dispersoid produces a material with excellent strength, the body centered cubic structure inherent to the material coupled with the high aspect ratio that results from processing operations produces some difficulties with ductility. The alloy is very sensitive to variations in a number of processing parameters, and if the high strength is once lost during fabrication, it cannot be recovered. The microstructural evolution of the alloy under irradiation falls into two regimes. Below about 550 C, dislocation development, {alpha}{prime} precipitation and void evolution in the matrix are observed, while above about 550 C damage appears to be restricted to cavity formation within oxide particles. The thermal expansion of the alloy is very similar to that of HT9 up to the temperature where HT9 undergoes a phase transition to austenitic. Pulse magnetic welding of end caps onto MA957 tubing can be accomplished in a manner similar to that in which it is performed on HT9, although the welding parameters appear to be very sensitive to variations in the tubing that result from small changes in fabrication conditions. The tensile and stress rupture behavior of the alloy are acceptable in the unirradiated condition, being comparable to HT9 below about 700 C and exceeding those of HT9

  16. Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

    NASA Astrophysics Data System (ADS)

    Li, Bo-Shiuan

    Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure at higher burnup and insignificant cladding dimensional change. However, severe degradation of SiC thermal conductivity with neutron irradiation will lead to higher fuel temperature with larger fission gas release. High stiffness of SiC has a drawback of accumulating large interfacial pressure upon pellet-cladding mechanical interactions (PCMI). This large stress will eventually reach the flexural strength of SiC, causing failure of SiC cladding instantly in a brittle manner instead of the graceful failure of ductile metallic cladding. The large interfacial pressure causes phenomena that were previously of only marginal significance and thus ignored (such as creep of the fuel) to now have an important role in PCMI. Consideration of the fuel pellet creep and elastic deformation in PCMI models in FRAPCON provide for an improved understanding of the magnitude of accumulated interfacial pressure. Outward swelling of the pellet is retarded by the inward irradiation-induced creep, which then reduces the rate of interfacial pressure buildup. Effect of PCMI can also be reduced and by increasing gap width and cladding thickness. However, increasing gap width and cladding thickness also increases the overall thermal resistance which leads to higher fuel temperature and larger fission

  17. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  18. Investigation of Nuclear Data Libraries with TRIPOLI-4 Monte Carlo Code for Sodium-cooled Fast Reactors

    NASA Astrophysics Data System (ADS)

    Lee, Y.-K.; Brun, E.

    2014-04-01

    The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.

  19. Analysis of Sodium Fire in the Containment Building of Prototype Fast Breeder Reactor Under the Scenario of Core Disruptive Accident

    SciTech Connect

    Rao, P.M.; Kasinathan, N.; Kannan, S.E.

    2006-07-01

    The potential for sodium release to reactor containment building from reactor assembly during Core Disruptive Accident (CDA) in Fast Breeder Reactors (FBR) is an important safety issue with reference to the structural integrity of Reactor Containment Building (RCB). For Prototype Fast Breeder Reactor (PFBR), the estimated sodium release under a CDA of 100 MJ energy release is 350 kg. The ejected sodium reacts easily with air in RCB and causes temperature and pressure rise in the RCB. For estimating the severe thermal consequences in RCB, different modes of sodium fires like pool and spray fires were analyzed by using SOFIRE -- II and NACOM sodium fire computer codes. Effects of important parameters like amount of sodium, area of pool, containment air volume and oxygen concentration have been investigated. A peak pressure rise of 7.32 kPa is predicted by SOFIRE II code for 350 kg sodium pool fire in 86,000 m{sup 3} RCB volume. Under sodium release as spray followed by unburnt sodium as pool fire mode analysis, the estimated pressure rise is 5.85 kPa in the RCB. In the mode of instantaneous combustion of sodium, the estimated peak pressure rise is 13 kPa. (authors)

  20. Evaluation of Homogeneous Options: Effects of Minor Actinide Exclusion from Single and Double Tier Recycle in Sodium Fast Reactors

    SciTech Connect

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

    2008-03-01

    The Systems Analysis Campaign under the Global Nuclear Energy Partnership (GNEP) has requested the fuel cycle analysis group at the Idaho National Laboratory (INL) to analyze and provide isotopic data for four scenarios in which different strategies for Minor Actinides (MA) management are investigated. A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design was selected as the baseline in this scenario study. Two transuranic (TRU) conversion ratios, defined as the ratio of the amount of TRU produced over the TRU destroyed in the reactor core, along with different fuel-types were investigated.

  1. ASTRID sodium cooled fast reactor: Program for improving in service inspection and repair

    SciTech Connect

    Jadot, F.; De Dinechin, G.; Augem, J. M.; Sibilo, J.

    2011-07-01

    In the frame of the CEA, EDF, AREVA coordinated research program for the development of Generation IV sodium-cooled fast reactors (SFR), the ASTRID project was launched in 2010. For the future prototype, the improvement of in-service inspection and repair (ISI and R) capabilities was identified as a major issue. Following the pluri-annual SFR research program, the ISI and R main R and D axes remain: i) improvement of the primary system conceptual design, ii) development of measurement and inspection techniques (continuous monitoring instrumentation and periodic inspection tools), iii) accessibility and associated robotics, and iv) development and validation of repair processes. Associated ISI and R needs are being defined through an iterative method between designers and instrumentation specialists: adaptation of the Design to ISI and R requirements, fission chamber development, validation of the ultrasonic and chemical transducers, of ultrasonic non destructive simulation, of acoustic surveillance, of laser repair intervention processes, of connected robotic equipment. Moreover, CEA, as leader of the ASTRID Project, is willing to find new contributors, partners or suppliers, in order to get innovative, diversified, exhaustive and efficient solutions. (authors)

  2. New concept of designing Pu and MA containing fuel for fast reactors

    NASA Astrophysics Data System (ADS)

    Savchenko, A. M.; Konovalov, I. I.; Vatulin, A. V.; Glagovsky, E. M.

    2009-03-01

    New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO 2 powder of <50 μm fractions prepared by pyrochemical or other methods. In this way a high uranium content fuel meat metallurgically bonded to cladding forms a heat conducting framework, pores of which contain PuO 2 powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment.

  3. Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

    SciTech Connect

    Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E.; Lovera, P.; Fleche, J. L.; Lacroix, M.; Carra, O.; Prele, G.

    2012-07-01

    Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO{sub 2} interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

  4. Management of transuranics using the Integral Fast Reactor (IFR) fuel cycle

    SciTech Connect

    Wade, D.C.

    1994-01-01

    The 50 years of activities following the discovery of self-sustaining fission chains have given rise to a buildup of roughly 900 tons of manmade transuranics. Of the total, about 260 tons of Pu{sup 239} were generated for use in weapons while the remainder were generated as a byproduct of electrical power produced worldwide by the commercial thermal nuclear power industry. What is to be done with these actinides? The options for disposition include interminable storage, burial, or recycle for use. The pros and cons of each option are being vigorously debated regarding the impact upon the issues of human and ecological risk -- both current and future; weapons proliferation potential -- both current and future; and total life cycle benefits and costs. As to the options for utilization, commercial uses for actinides (uranium and transuranics) are of limited diversity. The actinides have in the past and will in the future find application in large scale mostly by virtue of their ability to release energy through fission, and here their utility is unmatched -- whether the application be in commercial electricity generation or in armaments. The integral Fast Reactor (IFR) fuel cycle offers a number of features for management of the current and future burden of manmade transuranic materials and for capturing the energy content of the U{sup 238}. These features are discussed here.

  5. Change in physical properties of high density isotropic graphites irradiated in the ?JOYO? fast reactor

    NASA Astrophysics Data System (ADS)

    Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

    1995-08-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor "JOYO" to fluences from 2.11 to 2.86 × 10 26 n/m 2 ( E > 0.1 MeV) at temperatures from 549 to 597°C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens.

  6. Multiple lead seal assembly for a liquid-metal-cooled fast-breeder nuclear reactor

    DOEpatents

    Hutter, Ernest; Pardini, John A.

    1977-03-15

    A reusable multiple lead seal assembly provides leak-free passage of stainless-steel-clad instrument leads through the cover on the primary tank of a liquid-metal-cooled fast-breeder nuclear reactor. The seal isolates radioactive argon cover gas and sodium vapor within the primary tank from the exterior atmosphere and permits reuse of the assembly and the stainless-steel-clad instrument leads. Leads are placed in flutes in a seal body, and a seal shell is then placed around the seal body. Circumferential channels in the body and inner surface of the shell are contiguous and together form a conduit which intersects each of the flutes, placing them in communication with a port through the wall of the seal shell. Liquid silicone rubber sealant is injected into the flutes through the port and conduit; the sealant fills the space in the flutes not occupied by the leads themselves and dries to a rubbery hardness. A nut, threaded onto a portion of the seal body not covered by the seal shell, jacks the body out of the shell and shears the sealant without damage to the body, shell, or leads. The leads may then be removed from the body. The sheared sealant is cleaned from the body, leads, and shell and the assembly may then be reused with the same or different leads.

  7. Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel

    SciTech Connect

    Eliseev, V.A.; Malisheva, I.V.; Matveev, V.I.; Egorov, A.V.; Maslov, P.A.

    2013-07-01

    With the development and research into the generation IV fast sodium reactors, great attention is paid to the enhancement of the core inherent self-protection characteristics. One of the problems dealt here is connected with the reduction of the reactivity margin so that the control rods running should not result in the core overheating and melting. In this paper we consider the possibilities of improving the core of BN-1200 with oxide fuel by a known method of introducing an axial fertile layer into the core. But unlike earlier studies this paper looks at the possibility of using such a layer not only for improving breeding, but also for reducing sodium void reactivity effect (SVRE). This proposed improvement of the BN-1200 core does not solve the problem of strong interference in control and protection system (CPS) rods of BN-1200, but they reduce significantly the reactivity margin for burn-up compensation. This helps compensate all the reactivity balances in the improved core configurations without violating constraints on SVRE value.

  8. Status of advanced fuel candidates for Sodium Fast Reactor within the Generation IV International Forum

    SciTech Connect

    F. Delage; J. Carmack; C. B. Lee; T. Mizuno; M. Pelletier; J. Somers

    2013-10-01

    The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxide and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.

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

    SciTech Connect

    Wenner, Michael; Franceschini, Fausto; Ferroni, Paolo; Sartori, Alberto; Ricotti, Marco

    2012-07-01

    Westinghouse Electric Company (referred to as 'Westinghouse' in the rest of this paper) is proposing a 'back-to-front' approach to overcome the stalemate on nuclear waste management in the US. In this approach, requirements to further the societal acceptance of nuclear waste are such that the ultimate health hazard resulting from the waste package is 'as low as reasonably achievable'. Societal acceptability of nuclear waste can be enhanced by reducing the long-term radiotoxicity of the waste, which is currently driven primarily by the protracted radiotoxicity of the transuranic (TRU) isotopes. Therefore, a transition to a more benign radioactive waste can be accomplished by a fuel cycle capable of consuming the stockpile of TRU 'legacy' waste contained in the LWR Used Nuclear Fuel (UNF) while generating waste which is significantly less radio-toxic than that produced by the current open U-based fuel cycle (once through and variations thereof). Investigation of a fast reactor (FR) operating on a thorium-based fuel cycle, as opposed to the traditional uranium-based is performed. Due to a combination between its neutronic properties and its low position in the actinide chain, thorium not only burns the legacy TRU waste, but it does so with a minimal production of 'new' TRUs. The effectiveness of a thorium-based fast reactor to burn legacy TRU and its flexibility to incorporate various fuels and recycle schemes according to the evolving needs of the transmutation scenario have been investigated. Specifically, the potential for a high TRU burning rate, high U-233 generation rate if so desired and low concurrent production of TRU have been used as metrics for the examined cycles. Core physics simulations of a fast reactor core running on thorium-based fuels and burning an external TRU feed supply have been carried out over multiple cycles of irradiation, separation and reprocessing. The TRU burning capability as well as the core isotopic content have been characterized

  10. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    NASA Technical Reports Server (NTRS)

    Schnitzler, Bruce G.; Borowski, Stanley K.

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were

  11. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    SciTech Connect

    Bruce G. Schnitzler; Stanley K. Borowski

    2012-07-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine

  12. Generation IV nuclear energy system initiative. Large GFR core subassemblydesign for the Gas-Cooled Fast Reactor.

    SciTech Connect

    Hoffman, E. A.; Kulak, R. F.; Therios, I. U.; Wei, T. Y. C.

    2006-07-31

    Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics

  13. Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Analysis and Incident Management Guidelines (Risk Management for Sodium Fast Reactors)

    SciTech Connect

    Denman, Matthew R.; Groth, Katrina M.; Cardoni, Jeffrey N.; Wheeler, Timothy A.

    2015-04-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self-correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayesian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The authors would like to acknowledge the U.S. Department of Energy's Office of Nuclear Energy for funding this research through Work Package SR-14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at Argonne National Laboratory, Oak Ridge National Laboratory, and Idaho National Laboratory for their continue d contributions to the advanced reactor PRA mission area.

  14. Stability analysis of a natural circulation lead-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Lu, Qiyue

    This dissertation is aimed at nuclear-coupled thermal hydraulics stability analysis of a natural circulation lead cooled fast reactor design. The stability concerns arise from the fact that natural circulation operation makes the system susceptible to flow instabilities similar to those observed in boiling water reactors. In order to capture the regional effects, modal expansion method which incorporates higher azimuthal modes is used to model the neutronics part of the system. A reduced order model is used in this work for the thermal-hydraulics. Consistent with the number of heat exchangers (HXs), the reactor core is divided into four equal quadrants. Each quadrant has its corresponding external segments such as riser, plenum, pipes and HX forming an equivalent 1-D closed loop. The local pressure loss along the loop is represented by a lumped friction factor. The heat transfer process in the HX is represented by a model for the coolant temperature at the core inlet that depends on the coolant temperature at the core outlet and the coolant velocity. Additionally, time lag effects are incorporated into this HX model due to the finite coolant speed. A conventional model is used for the fuel pin heat conduction to couple the neutronics and thermal-hydraulics. The feedback mechanisms include Doppler, axial/radial thermal expansion and coolant density effects. These effects are represented by a linear variation of the macroscopic cross sections with the fuel temperature. The weighted residual method is used to convert the governing PDEs to ODEs. Retaining the first and second modes, leads to six ODEs for neutronics, and five ODEs for the thermal-hydraulics in each quadrant. Three models are developed. These are: 1) natural circulation model with a closed coolant flow path but without coupled neutronics, 2) forced circulation model with constant external pressure drop across the heated channels but without coupled neutronics, 3) coupled system including neutronics with

  15. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    NASA Astrophysics Data System (ADS)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  16. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    SciTech Connect

    Afifah, Maryam Su’ud, Zaki; Miura, Ryosuke; Takaki, Naoyuki; Sekimoto, H.

    2015-09-30

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  17. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  18. Bimodal space nuclear power system with fast reactor and Topaz II-type single-cell TFE

    NASA Astrophysics Data System (ADS)

    Ponomarev-Stepnoi, N. N.; Usov, V. A.; Ogloblin, B. G.; Shalaev, A. I.; Klimov, A. V.; Kirillov, E. Ya.; Shumov, D. P.; Radchenko, I. S.; Nicolaev, Y. V.

    1996-03-01

    The paper deals with characteristics and conceptual studies of a bimodal space thermionic system with a fast reactor and single-cell TFEs which is designed to operate in two modes: rated power mode providing power supply to space vehicle-mounted systems with energy consumption level of 10-80 kW(e) and forced thermal propulsion mode with thrust of 2200 N.

  19. Bimodal space nuclear power system with fast reactor and Topaz II-type single-cell TFE

    SciTech Connect

    Ponomarev-Stepnoi, N.N.; Usov, V.A.; Ogloblin, B.G.; Shalaev, A.I.; Klimov, A.V.; Kirillov, E.Y.; Shumov, D.P.; Radchenko, I.S.; Nicolaev, Y.V.

    1996-03-01

    The paper deals with characteristics and conceptual studies of a bimodal space thermionic system with a fast reactor and single-cell TFEs which is designed to operate in two modes: rated power mode providing power supply to space vehicle-mounted systems with energy consumption level of 10{endash}80 kW(e) and forced thermal propulsion mode with thrust of 2200 N. {copyright} {ital 1996 American Institute of Physics.}

  20. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    SciTech Connect

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  1. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    SciTech Connect

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2012-07-01

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

  2. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    SciTech Connect

    Su'ud, Zaki; Sekimoto, H.

    2014-09-30

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  3. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  4. Thermohydraulic model experiments on the transition from forced to natural circulation for pool-type fast reactors

    SciTech Connect

    Hoffmann, H.; Marten, K.; Weinberg, D. )

    1992-09-01

    In this paper, thermohydraulic studies on the transition from forced to natural convection are carried out using the 1:20 scale RAMONA three-dimensional reactor model with water as the simulant fluid. In the investigations, a scram from 40% load operation of a fast reactor is simulated. The core mass flows and the core as well as the hot plenum temperatures are measured as a function of time for various core power levels, coastdown curves of the primary- and secondary-side pumps, and for various delay times for the start of the immersion coolers after a scram. These parameters influence the onset of the natural circulation in the reactor tank. The main result is that the longer the intermediate heat exchanger coolability is ensured and the later the immersion coolers start to operate, the higher is the natural-circulation flow and, hence, the lower are the core temperatures.

  5. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    NASA Astrophysics Data System (ADS)

    Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

  6. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    SciTech Connect

    Malathi, N.; Sahoo, P. Ananthanarayanan, R.; Murali, N.

    2015-02-15

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.

  7. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    SciTech Connect

    Meriyanti; Su'ud, Zaki; Rijal, K.; Zuhair; Ferhat, A.; Sekimoto, H.

    2010-06-22

    In this study a feasibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850 deg. C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticality was obtained for this reactor.

  8. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    NASA Astrophysics Data System (ADS)

    Malathi, N.; Sahoo, P.; Ananthanarayanan, R.; Murali, N.

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ˜100 Hz/mm, ˜1 s, and ˜0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  9. Level monitoring system with pulsating sensor--application to online level monitoring of dashpots in a fast breeder reactor.

    PubMed

    Malathi, N; Sahoo, P; Ananthanarayanan, R; Murali, N

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation. PMID:25725884

  10. Renovation of CPF (Chemical Processing Facility) for Development of Advanced Fast Reactor Fuel Cycle System

    SciTech Connect

    Shinichi Aose; Takafumi Kitajima; Kouji Ogasawara; Kazunori Nomura; Shigehiko Miyachi; Yoshiaki Ichige; Tadahiro Shinozaki; Shinichi Ohuchi

    2008-01-15

    CPF (Chemical Processing Facility) was constructed at Nuclear Fuel Cycle Engineering Laboratories of JAEA (Japan Atomic Energy Agency) in 1980 as a basic research field where spent fuel pins from fast reactor (FR) and high level liquid waste can be dealt with. The renovation consists of remodeling of the CA-3 cell and the laboratory A, installation of globe boxes, hoods and analytical equipments to the laboratory C and the analytical laboratory. Also maintenance equipments in the CA-5 cell which had been out of order were repaired. The CA-3 cell is the main cell in which important equipments such as a dissolver, a clarifier and extractors are installed for carrying out the hot test using the irradiated FR fuel. Since the CPF had specialized originally in the research function for the Purex process, it was desired to execute the research and development of such new, various reprocessing processes. Formerly, equipments were arranged in wide space and connected with not only each other but also with utility supply system mainly by fixed stainless steel pipes. It caused shortage of operation space in flexibility for basic experimental study. Old equipments in the CA-3 cell including vessels and pipes were removed after successful decontamination, and new equipments were installed conformably to the new design. For the purpose of easy installation and rearranging the experimental equipments, equipments are basically connected by flexible pipes. Since dissolver is able to be easily replaced, various dissolution experiments is conducted. Insoluble residue generated by dissolution of spent fuel is clarified by centrifugal. This small apparatus is effective to space-saving. Mini mixer settlers or centrifugal contactors are put on to the prescribed limited space in front of the backside wall. Fresh reagents such as solvent, scrubbing and stripping solution are continuously fed from the laboratory A to the extractor by the reagent supply system with semi-automatic observation

  11. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    NASA Astrophysics Data System (ADS)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  12. Atmospheric Dispersion of Sodium Aerosol due to a Sodium Leak in a Fast Breeder Reactor Complex

    NASA Astrophysics Data System (ADS)

    Punitha, G.; Sudha, A. Jasmin; Kasinathan, N.; Rajan, M.

    Liquid sodium at high temperatures (470 K to 825 K) is used as the primary and secondary coolant in Liquid Metal cooled Fast Breeder Reactors (LMFBR). In the event of a postulated sodium leak in the Steam Generator Building (SGB) of a LMFBR, sodium readily combusts in the ambient air, especially at temperatures above 523 K. Intense sodium fire results and sodium oxide fumes are released as sodium aerosols. Sodium oxides are readily converted to sodium hydroxide in air due to the presence of moisture in it. Hence, sodium aerosols are invariably in the form of particulate sodium hydroxide. These aerosols damage not only the equipment and instruments due to their corrosive nature but also pose health hazard to humans. Hence, it is essential to estimate the concentration of sodium aerosols within the plant boundary for a sodium leak event. The Gaussian Plume Dispersion Model can obtain the atmospheric dispersion of sodium aerosols in an open terrain. However, this model does not give accurate results for dispersion in spaces close to the point of release and with buildings in between. The velocity field due to the wind is altered to a large extent by the intervening buildings and structures. Therefore, a detailed 3-D estimation of the velocity field and concentration has to be obtained through rigorous computational fluid dynamics (CFD) approach. PHOENICS code has been employed to determine concentration of sodium aerosols at various distances from the point of release. The dispersion studies have been carried out for the release of sodium aerosols at different elevations from the ground and for different wind directions.

  13. Analysis and Development of A Robust Fuel for Gas-Cooled Fast Reactors

    SciTech Connect

    Knight, Travis W

    2010-01-31

    The focus of this effort was on the development of an advanced fuel for gas-cooled fast reactor (GFR) applications. This composite design is based on carbide fuel kernels dispersed in a ZrC matrix. The choice of ZrC is based on its high temperature properties and good thermal conductivity and improved retention of fission products to temperatures beyond that of traditional SiC based coated particle fuels. A key component of this study was the development and understanding of advanced fabrication techniques for GFR fuels that have potential to reduce minor actinide (MA) losses during fabrication owing to their higher vapor pressures and greater volatility. The major accomplishments of this work were the study of combustion synthesis methods for fabrication of the ZrC matrix, fabrication of high density UC electrodes for use in the rotating electrode process, production of UC particles by rotating electrode method, integration of UC kernels in the ZrC matrix, and the full characterization of each component. Major accomplishments in the near-term have been the greater characterization of the UC kernels produced by the rotating electrode method and their condition following the integration in the composite (ZrC matrix) following the short time but high temperature combustion synthesis process. This work has generated four journal publications, one conference proceeding paper, and one additional journal paper submitted for publication (under review). The greater significance of the work can be understood in that it achieved an objective of the DOE Generation IV (GenIV) roadmap for GFR Fuel—namely the demonstration of a composite carbide fuel with 30% volume fuel. This near-term accomplishment is even more significant given the expected or possible time frame for implementation of the GFR in the years 2030 -2050 or beyond.

  14. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    SciTech Connect

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R.

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  15. Regulatory Technology Development Plan Sodium Fast Reactor. Mechanistic Source Term Development

    SciTech Connect

    Grabaskas, David S.; Brunett, Acacia Joann; Bucknor, Matthew D.; Sienicki, James J.; Sofu, Tanju

    2015-02-28

    Construction and operation of a nuclear power installation in the U.S. requires licensing by the U.S. Nuclear Regulatory Commission (NRC). A vital part of this licensing process and integrated safety assessment entails the analysis of a source term (or source terms) that represents the release of radionuclides during normal operation and accident sequences. Historically, nuclear plant source term analyses have utilized deterministic, bounding assessments of the radionuclides released to the environment. Significant advancements in technical capabilities and the knowledge state have enabled the development of more realistic analyses such that a mechanistic source term (MST) assessment is now expected to be a requirement of advanced reactor licensing. This report focuses on the state of development of an MST for a sodium fast reactor (SFR), with the intent of aiding in the process of MST definition by qualitatively identifying and characterizing the major sources and transport processes of radionuclides. Due to common design characteristics among current U.S. SFR vendor designs, a metal-fuel, pool-type SFR has been selected as the reference design for this work, with all phenomenological discussions geared toward this specific reactor configuration. This works also aims to identify the key gaps and uncertainties in the current knowledge state that must be addressed for SFR MST development. It is anticipated that this knowledge state assessment can enable the coordination of technology and analysis tool development discussions such that any knowledge gaps may be addressed. Sources of radionuclides considered in this report include releases originating both in-vessel and ex-vessel, including in-core fuel, primary sodium and cover gas cleanup systems, and spent fuel movement and handling. Transport phenomena affecting various release groups are identified and qualitatively discussed, including fuel pin and primary coolant retention, and behavior in the cover gas and

  16. Conceptual Design of Passive Safety System for Lead-Bismuth Cooled Fast Reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, A. G.; Nandiyanto, A. B. D.

    2016-04-01

    This paper presents the results of the conceptual design of passive safety systems for reactor power 225 MWth using Pb-Bi coolant. Main purpose of this research is to design of heat removal system from the reactor wall. The heat from the reactor wall is removed by RVACS system using the natural circulation from the atmosphere around the reactor at steady state. The calculation is performed numerically using Newton-Raphson method. The analysis involves the heat transfer systems in a radiation, conduction and natural convection. Heat transfer calculations is performed on the elements of the reactor vessel, outer wall of guard vessel and the separator plate. The simulation results conclude that the conceptual design is able to remove heat 1.33% to 4.67% from the thermal reactor power. It’s can be hypothesized if the reactor had an accident, the system can still overcome the heat due to decay.

  17. Characterization of scintillator materials for fast-ion loss detectors in nuclear fusion reactors

    NASA Astrophysics Data System (ADS)

    Jiménez-Ramos, M. C.; García López, J.; García-Muñoz, M.; Rodríguez-Ramos, M.; Carmona Gázquez, M.; Zurro, B.

    2014-08-01

    In fusion plasma reactors, fast ion generated by heating systems and fusion born particles must be well confined. The presence of magnetohydrodynamic (MHD) instabilities can lead to a significant loss of these ions, which may reduce drastically the heating efficiency and may cause damage to plasma facing components in the vacuum vessel. In order to understand the physics underlying the fast ion loss mechanism, scintillator based detectors have been installed in several fusion devices. In this work we present the absolute photon yield and its degradation with ion fluence in terms of the number of photons emitted per incident ion of several scintillators thin coatings: SrGa2S4:Eu2+ (TG-Green), Y3Al5O12:Ce3+ (P46) and Y2O3:Eu3+ (P56) when irradiated with light ions of different masses (deuterium ions, protons and α-particles) at energies between approximately 575 keV and 3 MeV. The photon yield will be discussed in terms of the energy deposited by the particles into the scintillator. For that, the actual composition and thickness of the thin layers were determined by Rutherford Backscattering Spectrometry (RBS). A collimator with 1 mm of diameter, which defines the beam size for the experiments, placed at the entrance of the chamber. An electrically isolated sample holder biased to +300 V to collect the secondary electrons, connected to a digital current integrator (model 439 by Ortec) to measure the incident beam current. A home made device has been used to store the real-time evolution of the beam current in a computer file allowing the correction of the IL yields due to the current fluctuations. The target holder is a rectangle of 150 × 112 mm2 and can be tilted. The X and Y movements are controlled through stepping motors, which permits a fine control of the beam spot positioning as well as the study of several samples without venting the chamber. A silica optical fiber of 1 mm diameter fixed to the vacuum chamber, which collects the light from the scintillators

  18. Particle-bed gas-cooled fast reactor (PB-GCFR) design. Project final technical report (Sept 2001 - Aug 2003).

    SciTech Connect

    Taiwo, T. A.; Wei, T. Y. C.; Feldman, E. E.; Hoffman, E. A.; Fatone, M.; Holland, J. W.; Prokofiev, I. G.; Yang, W. S.; Palmiotti, G.; Hill, R. N.; Todosow, M.; Salvatores, M.; Gandini, A.

    2003-10-27

    The objective of this project is to develop a conceptual design of a particle-bed, gas-cooled fast reactor (PB-GCFR) core that meets the advanced reactor concept and enhanced proliferation-resistant goals of the US Department of Energy's NERI program. The key innovation of this project is the application of a fast neutron spectrum environment to enhance both the passive safety and transmutation characteristics of the advanced particle-bed and pebble-bed reactor designs. The PB-GCFR design is expected to produce a high-efficiency system with a low unit cost. It is anticipated that the fast neutron spectrum would permit small-sized units ({approx} 150 MWe) that can be built quickly and packaged into modular units, and whose production can be readily expanded as the demand grows. Such a system could be deployed globally. The goals of this two-year project are as follows: (1) design a reactor core that meets the future needs of the nuclear industry, by being passively safe with reduced need for engineered safety systems. This will entail an innovative core design incorporating new fuel form and type; (2) employ a proliferation-resistant fuel design and fuel cycle. This will be supported by a long-life core design that is refueled infrequently, and hence, reduces the potential for fuel diversion; (3) incorporate design features that permit use of the system as an efficient transmuter that could be employed for burning separated plutonium fuel or recycled LWR transuranic fuel, should the need arise; and (4) evaluate the fuel cycle for waste minimization and for the possibility of direct fuel disposal. The application of particle-bed fuel provides the promise of extremely high burnup and fission-product protection barriers that may permit direct disposal.

  19. Ceramic Electron Multiplier

    DOE PAGESBeta

    Comby, G.

    1996-10-01

    The Ceramic Electron Multipliers (CEM) is a compact, robust, linear and fast multi-channel electron multiplier. The Multi Layer Ceramic Technique (MLCT) allows to build metallic dynodes inside a compact ceramic block. The activation of the metallic dynodes enhances their secondary electron emission (SEE). The CEM can be used in multi-channel photomultipliers, multi-channel light intensifiers, ion detection, spectroscopy, analysis of time of flight events, particle detection or Cherenkov imaging detectors. (auth)

  20. The Application of Structural Materials Data From the BN-350 Fast Reactor to Life Extension of Light Water Reactors

    SciTech Connect

    Romanenko, O.G.; Kislitsin, S.B.; Maksimkin, O.P.; Shiganakov, Sh.B.; Chumakov, Ye.V.; Dumchev, I.V.

    2006-07-01

    This paper describes the results of investigations of 08Cr16Ni11Mo3 (AISI 316 steel analogue) austenitic stainless steel irradiated in BN-350 breeder reactor at irradiation conditions close to that for Light Water Reactor (LWR) Internals. The pores were found in 08Cr16Ni11Mo3 steel irradiated at temperature 280 deg. C up to rather low damage 1.3 dpa and with dose rate 3.9 x 10{sup -9} dpa/s. There were obtained dose rate dependencies of yield strength, ultimate strength and ductility for 08Cr16Ni11Mo3 steel irradiated up to 7-13 dpa at 302-311 deg. C. These dependencies show a decrease in both yield strength and ultimate strength when dose rate decreases. There was observed an apparent decrease in total elongation when dose rate decreases, which was presumably connected with the pores formation in the steel at low dose rates. (authors)

  1. Characterization of scintillator materials for fast-ion loss detectors in nuclear fusion reactors

    NASA Astrophysics Data System (ADS)

    Jiménez-Ramos, M. C.; García López, J.; García-Muñoz, M.; Rodríguez-Ramos, M.; Carmona Gázquez, M.; Zurro, B.

    2014-08-01

    In fusion plasma reactors, fast ion generated by heating systems and fusion born particles must be well confined. The presence of magnetohydrodynamic (MHD) instabilities can lead to a significant loss of these ions, which may reduce drastically the heating efficiency and may cause damage to plasma facing components in the vacuum vessel. In order to understand the physics underlying the fast ion loss mechanism, scintillator based detectors have been installed in several fusion devices. In this work we present the absolute photon yield and its degradation with ion fluence in terms of the number of photons emitted per incident ion of several scintillators thin coatings: SrGa2S4:Eu2+ (TG-Green), Y3Al5O12:Ce3+ (P46) and Y2O3:Eu3+ (P56) when irradiated with light ions of different masses (deuterium ions, protons and α-particles) at energies between approximately 575 keV and 3 MeV. The photon yield will be discussed in terms of the energy deposited by the particles into the scintillator. For that, the actual composition and thickness of the thin layers were determined by Rutherford Backscattering Spectrometry (RBS). A collimator with 1 mm of diameter, which defines the beam size for the experiments, placed at the entrance of the chamber. An electrically isolated sample holder biased to +300 V to collect the secondary electrons, connected to a digital current integrator (model 439 by Ortec) to measure the incident beam current. A home made device has been used to store the real-time evolution of the beam current in a computer file allowing the correction of the IL yields due to the current fluctuations. The target holder is a rectangle of 150 × 112 mm2 and can be tilted. The X and Y movements are controlled through stepping motors, which permits a fine control of the beam spot positioning as well as the study of several samples without venting the chamber. A silica optical fiber of 1 mm diameter fixed to the vacuum chamber, which collects the light from the scintillators

  2. Multigroup calculation of criticality and power distribution in a two-pass fast spectrum cermet-fueled reactor

    SciTech Connect

    Anghaie, S.; Feller, G.J. ); Peery, S.D.; Parsley, R.C. )

    1992-01-01

    The advanced propulsion group at Pratt Whitney has developed a nuclear thermal rocket concept, the XNR2000, for use on lunear, Mars, and deep-space planetary missions. The XNR2000 engine is powered by a fast spectrum cermet-fueled nuclear reactor that heats up hydrogen propellant to a maximum of 2850 K. An expander cycle is used to deliver 12 kg/s hydrogen to the core, producing 25,000 lb[sub f] thrust at 944 s of specific impulse. The reactor comprises a beryllium-reflected outer annulus core and an inner core with the hydrogen propellant entering from the bottom of the outer core and exiting from the bottom part of the inner core to the thrust chamber. Both the outer and inner cores are loaded with prismatic cermet fuel elements. The baseline XNR2000 reactor core consists of 90 fuel elements in the outer core and 61 in the inner core, arranged in the pattern. This paper focuses on the neutronic analysis of the baseline XNR2000 reactor.

  3. Fast start-up, performance and microbial community in a pilot-scale anammox reactor seeded with exotic mature granules.

    PubMed

    Ni, Shou-Qing; Gao, Bao-Yu; Wang, Chih-Cheng; Lin, Jih-Gaw; Sung, Shihwu

    2011-02-01

    The possibility to introduce the exotic anammox sludge to seed the pilot-scale anammox granular reactor and its fast start-up for treating high nitrogen concentration wastewater were evaluated in this study. The reactor was started up successfully in two weeks; in addition, high nitrogen removal was achieved for a long period. Stoichiometry molar ratios of nitrite conversion and nitrate production to ammonium conversion were calculated to be 1.26±0.02:1 and 0.26±0.01:1, respectively. The Stover-Kincannon model which was first applied in granular anammox process indicated that the granular anammox reactor possessed high nitrogen removal potential of 27.8 kg/m(3)/d. The anammox granules in the reactor were characterized via microscope observation and fluorescence in situ hybridization technique. Moreover, the microbial community of the granules was quantified to be composed of 91.4-92.4% anammox bacteria by real-time polymerase chain reaction. This pilot study can elucidate further information for industrial granular anammox application. PMID:21145732

  4. Thermohydraulic model experiments and calculations on the transition from forced to natural circulation for pool-type fast reactors

    SciTech Connect

    Hoffmann, H.; Marten, K.; Weinberg, D.; Kamide, H.

    1990-01-01

    After a reactor scram, the decay heat removal (DHR) is of decisive importance for the safety of the plant. A fully passive DHR system based on natural circulation alone is independent of any power source. The DHE system consists of immersion coolers (ICs) installed in the hot plenum and connected to air coolers, each via intermediate circuits. During the postscram phase, the decay heat is to be removed by natural circulation from the core into the hot plenum and via the ICs and intermediate loops to the air coolers. The function of this DHR system is investigated and demonstrated in model tests with a geometry similar to the reactor, though on a different scale RAMONA is such a three-dimensional model set up on a 1:20 scale. It is operated with water. The steady-state tests for natural-circulation DHR operations have been conducted over a wide range of operational and geometric parameters. To study the transition from nominal to DHR conditions, experiments were defined to investigate the onset of natural circulation in the postscram phase (transient tests). The experiments were analyzed using the one-dimensional LEDHER code. LEDHER is a network analysis code for the long-term DHR of a fast reactor developed at Power Reactor and Nuclear Fuel Development Corporation in Japan. The results of the experiments and conclusions are summarized.

  5. Application of GRS method to evaluation of uncertainties of calculation parameters of perspective sodium-cooled fast reactor

    SciTech Connect

    Peregudov, A.; Andrianova, O.; Raskach, K.; Tsibulya, A.

    2012-07-01

    A number of recent studies have been devoted to the estimation of errors of reactor calculation parameters by the GRS (Generation Random Sampled) method. This method is based on direct sampling input data resulting in formation of random sets of input parameters which are used for multiple calculations. Once these calculations are performed, statistical processing of the calculation results is carried out to determine the mean value and the variance of each calculation parameter of interest. In our study this method is used for estimation of errors of calculation parameters (K{sub eff}, power density, dose rate) of a perspective sodium-cooled fast reactor. Neutron transport calculations were performed by the nodal diffusion code TRIGEX and Monte Carlo code MMK. (authors)

  6. Phase Characteristics of a Number of U-Pu-Am-Np-Zr Metallic Alloys for Use as Fast Reactor Fuels

    SciTech Connect

    Douglas E. Burkes; J. Rory Kennedy; Thomas Hartmann; Cynthia A. Papesch; Denis D. Keiser, Jr.

    2010-01-01

    Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. A series of test designs for the Advanced Fuel Cycle Initiative (AFCI) have been irradiated in the Advanced Test Reactor (ATR), designated as the AFC-1 and AFC-2 designs. Metal fuel compositions in these designs have included varying amounts of U, Pu, Zr, and minor actinides (Am, Np). Investigations into the phase behavior and relationships based on the alloy constituents have been conducted using x-ray diffraction and differential thermal analysis. Results of these investigations, along with proposed relationships between observed behavior and alloy composition, are provided. In general, observed behaviors can be predicted by a ternary U-Pu-Zr phase diagram, with transition temperatures being most dependent on U content. Furthermore, the enthalpy associated with transitions is strongly dependent on the as-cast microstructural characteristics.

  7. Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR). Metallic Fuels Program

    SciTech Connect

    Batte, G.L.; Pahl, R.G.; Hofman, G.L.

    1993-09-01

    This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II.

  8. Fast Pyrolysis of Poplar Using a Captive Sample Reactor: Effects of Inorganic Salts on Primary Pyrolysis Products

    SciTech Connect

    Mukarakate, C.; Robichaud, D.; Donohoe, B.; Jarvis, M.; Mino, K.; Bahng, M. K.; Nimlos, M.

    2012-01-01

    We have constructed a captive sample reactor (CSR) to study fast pyrolysis of biomass. The reactor uses a stainless steel wire mesh to surround biomass materials with an isothermal environment by independent controlling of heating rates and pyrolysis temperatures. The vapors produced during pyrolysis are immediately entrained and transported in He carrier gas to a molecular beam mass spectrometer (MBMS). Formation of secondary products is minimized by rapidly quenching the sample support with liquid nitrogen. A range of alkali and alkaline earth metal (AAEM) and transition metal salts were tested to study their effect on composition of primary pyrolysis products. Multivariate curve resolution (MCR) analysis of the MBMS data shows that transition metal salts enhance pyrolysis of carbohydrates and AAEM salts enhances pyrolysis of lignin. This was supported by performing similar separate studies on cellulose, hemicellulose and extracted lignin. The effect of salts on char formation is also discussed.

  9. On the use of a molten salt fast reactor to apply an idealized transmutation scenario for the nuclear phase out.

    PubMed

    Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara; Litskevich, Dzianis; Scholl, Susanne

    2014-01-01

    In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations--a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described. PMID:24690768

  10. On the Use of a Molten Salt Fast Reactor to Apply an Idealized Transmutation Scenario for the Nuclear Phase Out

    PubMed Central

    Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara; Litskevich, Dzianis; Scholl, Susanne

    2014-01-01

    In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations – a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described. PMID:24690768

  11. Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

    NASA Technical Reports Server (NTRS)

    Hsieh, T.-M.; Koenig, D. R.

    1977-01-01

    Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

  12. Nuclear data uncertainty propagation for neutronic key parameters of CEA's SFR V2B and CFV sodium fast reactor designs

    SciTech Connect

    Archier, P.; Buiron, L.; De Saint Jean, C.; Dos Santos, N.

    2012-07-01

    This paper presents a nuclear data uncertainty propagation analysis for two CEA's Sodium-cooled Fast Reactor designs: the SFR V2B and CFV cores. The nuclear data covariance matrices are provided by the DER/SPRC/LEPh's nuclear data team (see companion paper) for several major isotopes. From the current status of this analysis, improvements on certain nuclear data reactions are highlighted as well as the need for new specific integral experiments in order to meet the technological breakthroughs proposed by the CFV core. (authors)

  13. Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

    SciTech Connect

    Rowcliffe, A.F.; Tsai, H.C.; Smith, D.L.

    1997-08-01

    The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

  14. Comparison of measured and calculated air-transported radiation from a fast unshielded nuclear reactor. Technical report

    SciTech Connect

    Robitaille, H.A.; Hoffarth, B.E.

    1980-12-01

    Neutron and gamma-ray spectra have been measured at various distances up to 1100 metres from the fast-neutron reactor of the U.S. Army Pulse Radiation Division (Materiel Testing Directorate, Aberdeen Proving Ground, Md.) The spectra were obtained at a height of two metres above the air-ground interface and are compared to previous measurements performed by two other research laboratories, and also to the results of theoretical predictions based on two-dimensional discrete-ordinates transport theory. Integral quantities such as partial and total radiation kermas are generally in good agreement, however the theoretical calculations tend to predict somewhat softer neutron spectra than are observed experimentally.

  15. Reduced-Enrichment, Fast-Spectrum Lunar/Mars Surface Reactors

    SciTech Connect

    Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Sadasivan, Pratap; Amiri, Benjamin W.

    2006-07-01

    This report investigates the potential for reduced-enrichment, un-moderated surface reactors. The potential programmatic advantages of a reduced safeguards classification are discussed, along with design options that could allow a Cat III inventory surface reactor. Two potential reactor Cat III designs are presented, each with a thermal power of 100 kWt and a 5 year life: one is fueled with SS/UO{sub 2} fuel pins and the other with SS/UN fuel pins. The mass and performance parameters of these designs are compared to each other, and to a Cat III UZrH-fueled, moderated-spectrum design. (authors)

  16. Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Tsai, Hanchung

    2012-08-01

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 °C, cooling to 522 °C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ˜0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could

  17. Full-length U-xPu-10Zr (x=0, 8, 19 wt%) Fast Reactor Fuel Test in FFTF

    SciTech Connect

    D. L. Porter; H.C. Tsai

    2012-08-01

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt%) metallic fast reactor test with commercial-length (91.4 cm active fuel column length) conducted to date. With few remaining test reactors there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning of life (BOL) peak cladding temperature of the hottest pin was 608?C, cooling to 522?C at end of life (EOL). Selected fuel pins were examined non destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3 cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ~0.7 X/L axial location along the fuel column. This resulted from a lower production of rare earth fission products higher in the fuel column as well as a much smaller delta-T between fuel center and cladding, and therefore less FCCI, despite the higher cladding temperature. This behavior could

  18. Investigations on natural circulation in reactor models and shutdown heat removal systems for LMFBRs (liquid metal fast breeder reactors)

    SciTech Connect

    Hoffmann, H.; Weinberg, D.; Marten, K. ); Ieda, Yoshiaki )

    1989-11-01

    For sodium-cooled pool-type reactors, studies have been undertaken to remove the decay heat by natural convection alone, as in the case of failure of all power supplies. For this purpose, four immersion coolers (ICs), two each installed at a 180-deg circumferential position with respect to the others, are arranged within the reactor tank. They are connected with natural-drift air coolers through independent intermediate circuits. The primary sodium in the tank as well as the secondary sodium in the intermediate loop circulate by natural convection. The general functioning of this passive shutdown decay heat removal (DHR) system is demonstrated in 1:20 and 1:5 scale test models using water as a simulant fluid for sodium. The model design is based on the thermohydraulics similarity criteria. In the RAMONA three-dimensional 1:20 scale model, experiments were carried out to clarify the steady-state in-vessel thermohydraulics for different parameter combinations (core power, radial power distribution across the core, DHR by 2 or 4 ICs in operation, above-core structure geometry and position, different IC designs). For all mentioned parameters, temperatures and their fluctuations were measured and used to indicate isotherms and lines of identical temperature fluctuations. The flow patterns were observed visually. The experiments were recalculated by an updated version of the single-phase three-dimensional thermohydraulics code COMMIX.

  19. Ceramic breeder materials

    SciTech Connect

    Johnson, C.E.

    1990-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 32 refs., 1 fig., 1 tab.

  20. R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

    SciTech Connect

    Vanderhaegen, M.; Paumel, K.; Seiler, J. M.; Tourin, A.; Jeannot, J. P.; Rodriguez, G.

    2011-07-01

    In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boiling intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)