WorldWideScience

Sample records for burner fast reactors

  1. Fabrication of particulate metal fuel for fast burner reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Lee, Sun Yong; Kim, Jong Hwan; Woo, Yoon Myung; Ko, Young Mo; Kim, Ki Hwan; Park, Jong Man; Lee, Chan Bok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    U Zr metallic fuel for sodium cooled fast reactors is now being developed by KAERI as a national R and D program of Korea. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. Therefore, innovative fuel concepts should be developed to address the fabrication challenges pertaining to TRU while maintaining good performances of metallic fuel. Particulate fuel concepts have already been proposed and tested at several experimental fast reactor systems and vipac ceramic fuel of RIAR, Russia is one of the examples. However, much less work has been reported for particulate metallic fuel development. Spherical uranium alloy particles with various diameters can be easily produced by the centrifugal atomization technique developed by KAERI. Using the atomized uranium and uranium zirconium alloy particles, we fabricated various kinds of powder pack, powder compacts and sintered pellets. The microstructures and properties of the powder pack and pellets are presented.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Some aspects of risk reduction strategy by multiple recycling in fast burner reactors of the plutonium and minor actinide inventories

    International Nuclear Information System (INIS)

    The paper shows the impact of recycling LWR-MOX fuel in a fast burner reactor on the plutonium (Pu) and minor actinide (MA) inventories and on the related radio activities. Reprocessing of the targets for multiple recycling will become increasingly difficult as the burn up increases. Multiple recycling of Pu + MA in fast reactors is a feasible option which has to be studied very carefully: the Pu (except the isotopes Pu-238 and Pu-240), Am and Np levels decrease as a function of the recycle number, while the Cm-244 level accumulates and gradually transforms into Cm-245. Long cooling times (10 + 2 years) are necessary with aqueous processing. The paper discusses the problems associated with multiple reprocessing of highly active fuel types and particularly the impact of Pu-238, Am-241 and Cm-244 on the fuel cycle operations. The calculations were performed with the zero-dimensional ORIGEN-2 code. The validity of the results depends on that of the code and its cross section library. The time span to reduce the initial inventory of Pu + MA by a factor of 10, amounts to 255 years when average burn ups are limited to 150 GWd t-1. (orig.)

  4. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  5. Advanced Burner Reactor Preliminary NEPA Data Study.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  6. Fast Reactor Development Strategy in China

    International Nuclear Information System (INIS)

    As one of the largest developing countries, China needs a reliable energy supplement. At the same time, China should improve the energy structure to decrease CO2 emissions. Nuclear and renewable energies are the main solutions to these issues. According to the research results, the nuclear capacity should increase to 400 GW(e) up to 2050. Fast reactors must be developed considering the limitation of uranium resources. In order to deploy fast reactor technology, the ‘experimental reactor, demonstration reactor and commercial reactor’ strategy has been suggested. China has finished the construction of the China Experimental Fast Reactor (CEFR) and gained necessary experience about fast reactors. The China Institute of Atomic Energy (CIAE) has begun to design the CFR-600, a 600 MW(e) demonstration fast reactor. This reactor will be put into operation before 2025. After that, a larger commercial reactor will be constructed. Besides fast reactors, all of other key sectors of fuel cycle will be developed at the same time such as reprocessing, fast reactor fuel, etc. There are two main tasks of fast reactors, one of which is to raise the utility ratio of uranium, and the other one is to transmute the long life waste of light water reactors. The fast reactor will be designed as a breeder and burner, respectively. (author)

  7. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  8. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    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

  9. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Analysis of Reactor Deployment Scenarios with Introduction of SFR Breakeven Reactors and Burners Using DANESS Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2008-01-15

    Using the DANESS code newly employed for future scenario analysis, reactor deployment scenarios with the introduction of sodium cooled fast reactors(SFRs) having different conversion ratios in the existing PWRs dominant nuclear fleet have been analyzed to find the SFR deployment strategy for replacing PWRs with the view of a spent fuel reduction and an efficient uranium utilization through its reuse in a closed nuclear fuel cycle. Descriptions of the DANESS code and how to use are briefly given from the viewpoint of its first application. The use of SFRs and recycling of TRUs by reusing PWR spent fuel leads to the substantial reduction of the amount of PWR spent fuel and environmental burden by decreasing radiotoxicity of high level waste, and a significant improvement on the natural uranium resources utilization. A continuous deployment of burners effectively decreases the amount of PWR spent fuel accumulation, thus lightening the burden for PWR spent fuel management. An introduction of breakeven reactors effectively reduces the uranium demand through producing excess TRU during the operation, thus contributing to a sustainable nuclear power development. With SFR introduction starting in 2040, PWRs will remain as a main power reactor type till 2100 and SFRs will be in support of waste minimization and fuel utilization.

  11. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    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

  12. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  13. Fast breeder reactor research

    International Nuclear Information System (INIS)

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  14. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarizes the fast reactor research carried out by ECN during the period covering the year 1980. This research is mainly concerned with the cores of sodium-cooled breeders, in particular the SNR-300, and its related safety aspects. It comprises six items: A programme to determine relevant nuclear data of fission- and corrosion-products; A fuel performance programme comprising in-pile cladding failure experiments and a study of the consequences of loss-of-cooling and overpower; Basic research on fuel; Investigation of the changes in the mechanical properties of austenitic stainless steel DIN 1.4948 due to fast neutron doses, this material has been used in the manufacture of the reactor vessel and its internal components; Study of aerosols which could be formed at the time of a fast reactor accident and their progressive behaviour on leaking through cracks in the concrete containment; Studies on heat transfer in a sodium-cooled fast reactor core. As fast breeders operate at high power densities, an accurate knowledge of the heat transfer phenomena under single-phase and two-phase conditions is sought. (Auth.)

  15. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    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 ampersand D

  16. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarises the fast reactor research carried out at the Netherlands Energy Research Centre during the year 1981. The neutron and fission product cross sections of various isotopes have been evaluated. In the fuel performance programme, some preliminary results are given and irradiation facilities described. Creep experiments on various stainless steel components are reported

  17. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  18. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  19. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  20. Use of freeze-casting in advanced burner reactor fuel design

    Energy Technology Data Exchange (ETDEWEB)

    Lang, A. L.; Yablinsky, C. A.; Allen, T. R. [Dept. of Engineering Physics, Univ. of Wisconsin Madison, 1500 Engineering Drive, Madison, WI 53711 (United States); Burger, J.; Hunger, P. M.; Wegst, U. G. K. [Thayer School of Engineering, Dartmouth College, 8000 Cummings Hall, Hanover, NH 03755 (United States)

    2012-07-01

    This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models

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

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); 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.

  2. Sodium fast reactor safety and licensing research plan - Volume II

    International Nuclear Information System (INIS)

    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.

  3. Fast reactors and nuclear nonproliferation

    International Nuclear Information System (INIS)

    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. (author)

  4. Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Wegst, Ulrike G.K. [Dartmouth College, Hanover, NH (United States). Thayer School of Engineering; Allen, Todd [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States)

    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.

  5. Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

    International Nuclear Information System (INIS)

    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.

  6. Fast reactors using molten chloride salts as fuel

    International Nuclear Information System (INIS)

    This report deals with a rather exotic ''paper reactor'' in which the fuel is in the form of molten chlorides. (a) Fast breeder reactor with a mixed fuel cycle of thorium/uranium-233 and uranium 238/plutonium in which all of the plutonium can be burned in situ and in which a denatured mixture of uranium-233 and uranium-238 is used to supply further reactors. The breeding ratio is relatively high, 1.58 and the specific power is 0.75 GW(th)/m3 of core. (b) Fast breeder reactor with two and three zones (internal fertile zone, intermediate fuel zone, external fertile zone) with an extremely high breeding ratio of 1.75 and a specific power of 1.1 GW(th)/m3 of core. (c) Extremely high flux reactor for the transmutation of the fission products: strontium-90 and caesium-137. The efficiency of transmutation is approximately 15 times greater than the spontaneous beta decay. This high flux burner reactor is intended as part of a complex breeder/burner system. (d) Internally cooled fast breeder in which the cooling agent is the molten fertile material, the same as in the blanket zone. This reactor has a moderate breeding ratio of 1.38, a specific power of 0.22 GW(th)/m3 of core and very good inherent safety properties. All of these reactors have the fuel in the form of molten chlorides: PuCl3 as fissile, UCl3 as fertile (if needed) and NaCl as dilutent. The fertile material can be 238UCl3 as fertile and NaCl as dilutent. In mixed fuel cycles the 233UCl3 is also a fissile component with 232ThCl4 as the fertile constituent

  7. Heterogeneous Transmutation Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  9. Status of national programmes on fast reactors

    International Nuclear Information System (INIS)

    Based on the International Working Group on Fast reactors (IWGFR) members' request, the IAEA organized a special meeting on Fast Reactor Development and the Role of the IAEA in May 1993. The purpose of the meeting was to review and discuss the status and recent development, to present major changes in fast reactor programmes and to recommend future activities on fast reactors. The IWGFR took note that in some Member States large prototypes have been built or are under construction. However, some countries, due to their current budget constraints, have reduced the level of funding for research and development programmes on fast reactors. The IWGFR noted that in this situation the international exchange of information and cooperation on the development of fast reactors is highly desirable and stressed the importance of the IAEA's programme on fast reactors. These proceedings contain important and useful information on national programmes and new developments in sodium cooled fast reactors in Member States. Refs, figs and tabs

  10. On fast reactor kinetics studies

    Energy Technology Data Exchange (ETDEWEB)

    Seleznev, E. F.; Belov, A. A. [Nuclear Safety Inst. of the Russian Academy of Sciences IBRAE (Russian Federation); Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F. [Inst. for Physics and Power Engineering IPPE (Russian Federation)

    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)

  11. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

  12. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  13. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  14. Heterogeneous Recycling in Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Fast breeder reactor protection system

    Science.gov (United States)

    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)

  16. A review of the UK fast reactor programme. March 1977

    International Nuclear Information System (INIS)

    This paper reports on the Fast Reactor Programme of United Kingdom. These are the main lines: Dounreay Fast Reactor; Prototype Fast Reactor; Commercial Fast Reactor; engineering development; materials development; chemical engineering/sodium technology; fast reactor fuel; fuel cycle; safety; reactor performance study

  17. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Interfacial effects in fast reactors

    International Nuclear Information System (INIS)

    The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed to measure U-238 capture rates near th blanket--reflector interface in the MIT Blanket Test Facility. Prior MIT experiments on a thorium--uranium interface in a blanket assembly were also reanalyzed. Extremely localized fertile capture rate increases of on the order of 50% were measured immediately at the interfaces relative to extrapolation of asymptotic interior traverses, and relative to state-of-the-art (LIB-IV, SPHINX, ANISN/2DB) calculations which employ infinite-medium self-shielding throughout a given zone. A method was developed to compute a spatially varying background scattering cross section per absorber nucleus which takes into account both homogeneous and heterogeneous effects on the interface flux transient

  19. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

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

    International Nuclear Information System (INIS)

    Highlights: • Thorium as support fertile material for TRU transmutation in Fast Reactors. • Comparative analysis of Th and U based breakeven and burner Fast Reactors. • Thorium fosters significant advantages in terms of safety parameters. • Inherent safety is investigated through quasi-static reactivity and energy balances. • Th use in low-CR Fast Reactors does not reduce fuel decay heat and neutron sources. - Abstract: 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. Computational Neutronics Methods and Transmutation Performance Analyses for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

    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

  3. A CFD Simulation Process for Fast Reactor Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Kurt D. Hamman; Ray A. Berry

    2010-09-01

    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.

  4. A CFD simulation process for fast reactor fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hamman, Kurt D., E-mail: Kurt.Hamman@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Berry, Ray A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States)

    2010-09-15

    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-{epsilon} and SST (Menter) k-{omega} 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.

  5. Sodium fast reactors with closed fuel cycle

    CERN Document Server

    Raj, Baldev; Vasudeva Rao, PR 0

    2015-01-01

    Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

  6. Decommissioning of fast reactors after sodium draining

    International Nuclear Information System (INIS)

    Acknowledging the importance of passing on knowledge and experience, as well mentoring the next generation of scientists and engineers, and in response to expressed needs by Member States, the IAEA has undertaken concrete steps towards the implementation of a fast reactor data retrieval and knowledge preservation initiative. Decommissioning of fast reactors and other sodium bearing facilities is a domain in which considerable experience has been accumulated. Within the framework and drawing on the wide expertise of the Technical Working Group on Fast Reactors (TWG-FR), the IAEA has initiated activities aiming at preserving the feedback (lessons learned) from this experience and condensing those to technical recommendations on fast reactor design features that would ease their decommissioning. Following a recommendation by the TWG-FR, the IAEA had convened a topical Technical Meeting (TM) on 'Operational and Decommissioning Experience with Fast Reactors', hosted by CEA, Centre d'Etudes de Cadarache, France, from 11 to 15 March 2002 (IAEA-TECDOC- 1405). The participants in that TM exchanged detailed technical information on fast reactor operation and decommissioning experience with various sodium cooled fast reactors, and, in particular, reviewed the status of the various decommissioning programmes. The TM concluded that the decommissioning of fast reactors to reach safe enclosure presented no major difficulties, and that this had been accomplished mainly through judicious adaptation of processes and procedures implemented during the reactor operation phase, and the development of safe sodium waste treatment processes. However, the TM also concluded that, on the path to achieving total dismantling, challenges remain with regard to the decommissioning of components after sodium draining, and suggested that a follow-on TM be convened, that would provide a forum for in-depth scientific and technical exchange on this topic. This publication constitutes the Proceedings of

  7. Liquid Metal Coolant Technology for Fast Reactors

    International Nuclear Information System (INIS)

    In the paper presented are results of comparative analysis and the choice of liquid metal coolants for fast reactors, the current status of studies on the physical chemistry and technology of sodium coolants for fast neutron reactors and heavy liquid metal coolants, namely, lead-bismuth and lead for fast reactors and accelerator driven systems. There are descriptions of devices designed for control of the impurities in sodium coolants and their removal as well as methods of heavy liquid metal coolant quality control, removal of impurities from heavy liquid metal coolants and the steel surface of components of nuclear power plants (NPPs) and relevant equipment. Attention is given to the issues of modelling of impurity mass transfer in liquid metal coolants and designing new liquid metal coolants for NPPs. Results of the analysis of NPP abnormal operating conditions are presented. The adopted design approaches assure reliable protection against accidents. Up to now, about 200 reactor-years of sodium cooled fast reactor operation and about 80 reactor-years of submarine reactor operation have been gained. The new goals for sodium and heavy liquid metal coolant technology have been formulated as applied to the new generation fast reactors. (author)

  8. Fast molten salt reactor-transmuter for closing nuclear fuel cycle on minor actinides

    International Nuclear Information System (INIS)

    Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle is the most perspective and actual direction. The reactor on melts salts - molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed nuclear fuel cycle, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides in molten salt reactor is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF - BeF2; NaF - LiF - BeF2; NaF-LiF ; NaF-ZrF4 ; LiF-NaF -KF; NaCl. RRC 'Kurchatov institute' together with other employees have developed the basic design reactor installations with molten salt reactor - burner long-living nuclides for fluoride fuel composition with the limited solubility minor actinides (MAF3 10 mol %) allows to develop in some times more effective molten salt reactor with fast neutron spectrum - burner/ transmuter of the long-living radioactive waste. In high-flux fast reactors on melts salts within a year it is possible to burn ∼300 kg minor actinides per 1 GW thermal power of reactor. The technical and economic estimation given power

  9. Review of fast reactor activities in India

    International Nuclear Information System (INIS)

    It may be recalled that In the presentation at the last meeting of the IWGFR (13th Annual meeting), a broad outline of India's nuclear energy programme and the role of fast breeders in the programme has been provided. The steps taken to enable the fast breeders to fulfil their role have also been described. In brief, fast breeder reactors are considered as an essential and integral part of the programme of nuclear energy and constitute the second step in the programme, the first being the construction of natural uranium heavy water moderated reactors which will consume natural uranium but will produce plutonium to fuel fast breeder reactors. This basic position has remained unchanged and the Government is now taking steps to build a large number of heavy water reactors, say 10 million Kw capacity in the next 20 years. This defines the time frame for developing the fast breeder technology in the country. It has therefore been decided to mobilise the efforts towards design, construction and operation of a medium sized (about 500 M We) reactor by mid-nineties. Thus, the climate for fast breeder reactors is good and there is a good deal of enthusiasm amongst the scientists and engineers working in the field although the actual implementation of the programme during the year had to face certain difficulties

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

    Science.gov (United States)

    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.

  11. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)

    2015-11-04

    This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and

  12. Stationary Liquid Fuel Fast Reactor

    International Nuclear Information System (INIS)

    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

  13. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

    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

  14. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

  15. Spatial Kinetics in Fast Reactors

    International Nuclear Information System (INIS)

    Reactor neutronic calculations designed for calculating of unsteady processes in a real 3D geometry require processing of a large amount of information. They cannot consist of simple models, as they should reflect the processes of variations of all local reactor characteristics. The model complexity and the significant time needed for numerical solution of neutron-transport equations limit the choice of methods that can achieve the required accuracy. Thus there is an urgent need for the development of various methods enabling the solution of unsteady neutron-transport equations and estimates of their errors, spent time and consistency with the experimental data. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

    A general description of expert systems is given. The operation of a fast reactor is reviewed. The expert system to the diagnosis of breakdowns limited to the reactor core. The structure of the system is described: specification of the diagnostics; structure of the data bank and evaluation of the rules; specification of the prediagnostics and evaluation; explanation of the diagnostics; time evolution of the system; comparison with other expert systems. Applications to some cases of faults are finally presented

  18. Current status of fast reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, H.H.

    1979-01-01

    The subject of calculation of reactivity coefficients for fast reactors is developed, starting with a discussion of the status of relevant nuclear data and proceeding to the subjects of group cross section generation and of methods of obtaining reactivity coefficients from group cross sections. Reactivity coefficients measured in critical experiments are compared with calculated values. Dependence of reactivity coefficients on reactor design is discussed. Finally, results of the recent international comparison of calculated reactivity coefficients are presented.

  19. Fast Reactor Knowledge Management at IGCAR, India

    International Nuclear Information System (INIS)

    The Process Architecture: → Acquire: Solicitation; Voluntary submission; Mandatory requirements; Interview/Observation; → Quality Control: Review/Editing; Certification; Quality index; → Disseminate: Publish through the Technology architecture; Formal/Informal Meetings; COPs; → Utilize: Projects; Day-to-day activities; → Maintenance; → Retirement. Mission: To conduct a broad based multidisciplinary programme of scientific research and advanced engineering development, directed towards the establishment of the technology of Sodium Cooled Fast Breeder Reactors (FBR) and associated fuel cycle facilities in the Country. The mission includes the development and applications of new and improved materials, techniques, equipment and systems for FBRs, pursue basic research to achieve breakthroughs in Fast Reactor technology

  20. TITAN program and direct cycle fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Yasuyoshi; Yoshizawa, Yoshio; Nitawaki, Takeshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

    2000-07-01

    In December 1999, the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (TIT) started a new program for the development of advanced nuclear reactors with small and medium size. TITAN is the acronym for the program. A novel concept of a carbon dioxide cooled direct cycle fast reactor with a Rankin cycle has been proposed as the advanced nuclear reactors and evaluated for an alternative option to liquid metal cooled fast reactors (LMFRs). The use of carbon dioxide as coolant eliminates major safety related problems of sodium cooled fast reactors: positive sodium void reactivity, hazardous reaction between sodium and water or air. The decay heat is passively removed by allocating a storage tank of liquidized carbon dioxide between the regenerator and the condenser, and by introducing naturally the carbon dioxide vaporized from the tank into the core in the event of the depressurization accident. The direct cycle results in considerable simplification of the heat transport system owing to the absence of intermediate cooling and water-steam loops comparing with the LMFRs. The thermal efficiency of the direct cycle is evaluated as 34.3 %, which is slightly higher than those in the current BWRs and PWRs. (author)

  1. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    A review of the United Kingdom Fast Reactor Programme is introduced. Operational experience with the Prototype Fast Reactor (PFR) is briefly summarized. The design concept of the Commercial Demonstration Fast Reactor (CDFR) is given in some detail. The emphasis is on materials development, chemical engineering/sodium technology, fuel reprocessing and fuel cycle, engineering component development and reactor safety

  2. Progress of China Experimental Fast Reactor in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    1 Background Fast reactor is the reactor which realized the chain fission with fast neutron.As an optional type of generation Ⅳ reactor,fast reactor has three characters:1) It can change 238U to 239Pu and raise the uranium resource utilization

  3. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    International Nuclear Information System (INIS)

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated 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 overcome 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 thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

  5. Liquid metal fast reactor transient design

    International Nuclear Information System (INIS)

    An examination has been made of how the currently available computing capabilities could be used to reduce Liquid Metal Fast Reactor design, manufacturing, and construction cost. While the examination focused on computer analyses some other promising means to reduce costs were also examined. (author)

  6. Fast Reactor and ADS development in China

    International Nuclear Information System (INIS)

    Conclusion: • The Fukushima accident influence China deeply. “The 12th five years plan and 2020 perspective goal of nuclear safety and radioactive pollution prevention” has been approved which means the nuclear may restart in the near future. • A demonstration fast reactor is under design. • More and more research works will be executed on CEFR

  7. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

    The invention, in which the United States Department of Energy has participated as co-inventor, relates to breeder reactor fuel elements, and specifically to such elements incorporating 'getters', hereafter designated as fission product traps. The main object of the invention is the construction of a fast breeder reactor fuel pin, free from local stresses induced in the cladding by reactions with cesium. According to the invention, the fast breeder fuel element includes a cladding tube, sealed at both ends by a plug, and containing a fissile stack and a fertile stack, characterized by the interposition of a cesium trap between the fissile and fertile stacks. The trap is effective at reactor operating temperatures in retaining and separating the cesium generated in the fissile material and preventing cesium reaction with the fertile stack. Depending on the construction method adopted, the trap may consists of a low density titanium oxide or niobium oxide pellet

  8. The development of fast reactors in France

    International Nuclear Information System (INIS)

    Only minor changes were introduced in the French nuclear programme by the new government in 1981. The operating conditions of Rapsodie were very satisfactory up to January 1982. After a leak in the double primary jacket (nitrogen circuit) the reactor was shut down for investigations. Phenix is continuing to operate smoothly. Construction of Super Phenix (Creys Malville power plant) is proceeding normally though with some delay. The studies for the future (after Creys Malville) are following their way both for the Project 1500 (Super Phenix 2) and for the specific plants of the fuel cycle. Research and development are largely directed toward Super Phenix 1 needs and the prospects of Super Phenix 2. International cooperation remains very intensive. The financial resources devoted to the development of fast reactors are globally stable. Including fuel cycle and safety (but excluding the Phenix operation) about 1300 millions of francs will be devoted to fast reactors by the C.E.A. in 1982. (author)

  9. Fast Reactor Fuel Development in Europe

    International Nuclear Information System (INIS)

    Research and development of minor-actinide-bearing fuels in Europe has made significant progress, with a number of scoping irradiation tests made on a number of candidate fuels foreseen for fast reactors and dedicated minor actinide transmutation systems, e.g. the accelerator driven system. Currently, efforts concentrate on uranium based fuels, as the deployment of fast reactor fleets requires Pu generation in order to achieve sustainability. Both homogeneous and heterogeneous concepts for minor actinide reactor recycling are considered. In the former, the minor actinides are added in small quantities to the mixed oxide fuel, while in the latter, the minor actinides are loaded in significant quantities in UO2. Irradiation programmes to test these concepts for pellet and SPHEREPAC fuel configurations are under way. (author)

  10. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity

  11. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  12. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

    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.

  13. Intrinsically secure fast reactors with dense cores

    Energy Technology Data Exchange (ETDEWEB)

    Slessarev, Igor [29, Res. Tivoli, Allee des Peupliers, 13090 Aix-en-Provence (France)], E-mail: igor.slessarev@free.fr

    2007-11-15

    Secure safety, resistance to weapons material proliferation and problems of long-lived wastes remain the most important 'painful points' of nuclear power. Many innovative reactor concepts have been developed aimed at a radical enhancement of safety. The promising potential of innovative nuclear reactors allows for shifting accents in current reactor safety 'strategy' to reveal this worth. Such strategy is elaborated focusing on the priority for intrinsically secure safety features as well as on sure protection being provided by the first barrier of defence. Concerning the potential of fast reactors (i.e. sodium cooled, lead-cooled, etc.), there are no doubts that they are able to possess many favourable intrinsically secure safety features and to lay the proper foundation for a new reactor generation. However, some of their neutronic characteristics have to be radically improved. Among intrinsically secure safety properties, the following core parameters are significantly important: reactivity margin values, reactivity feed-back and coolant void effects. Ways of designing intrinsically secure safety features in fast reactors (titled hereafter as Intrinsically Secure Fast Reactors - ISFR) can be found in the frame of current reactor technologies by radical enhancement of core neutron economy and by optimization of core compositions. Simultaneously, respecting resistance to proliferation, by using non-enriched fuel feed as well as a core breeding gain close to zero, are considered as the important features (long-lived waste problems will be considered in a separate paper). This implies using the following reactor design options as well as closed fuel cycles with natural U as the reactor feed: {center_dot}Ultra-plate 'dense cores' of the ordinary (monolithic) type with negative total coolant void effects. {center_dot}Modular type cores. Multiple dense modules can be embedded in the common reflector for achieving the desired NPP total

  14. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide, or Tribaloy 700 (a nickel-base hard-facing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping

  15. Fuel management codes for fast reactors

    International Nuclear Information System (INIS)

    The CAPHE code is used for managing and following up fuel subassemblies in the Phenix fast neutron reactor; the principal experimental results obtained since this reactor was commissioned are analyzed with this code. They are mainly concerned with following up fuel subassembly powers and core reactivity variations observed up to the beginning of the fifth Phenix working cycle (3/75). Characteristics of Phenix irradiated fuel subassemblies calculated by the CAPHE code are detailed as at April 1, 1975 (burn-up steel damage)

  16. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

    Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

  17. Sodium fast reactor evaluation: Core materials

    Science.gov (United States)

    Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon; Raison, J. P.; Mizuno, T.; Delage, F.; Carmack, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

  18. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  19. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward

    International Nuclear Information System (INIS)

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high

  20. Alternative concept for a fast energy amplifier accelerator driven reactor

    International Nuclear Information System (INIS)

    Recently Rubbia et al. introduced a conceptual design of a Fast Energy Amplifier (EA) as an advanced innovative reactor which utilizes a neutron spallation source induced by protons as an external source in a subcritical array imbibed a molten lead coolant which, besides being breeder and waste burner, generates energy. This paper introduces some qualitative changes in Rubbia's concept such as more than one point of spallation, in order to reduce the requirement in the energy and current of the accelerator, and mainly to make a more flat neutron distribution. The subcritical core which in Rubbia's concept is an hexagonal array of pins immersed in a molten lead coolant is replaced by a concept of a solid lead calandria with the fuel elements in channels cooled by helium, allowing on line refueling or shuffling, and the utilization of a direct thermodynamic cycle (Brayton), which is more efficient than a vapor cycle. Although the calculations to demonstrate the feasibility of the EA alternative concept are underway and not yet finished, these ideas do not violate the basic physics of the EA, as showed in this paper, with evident advantages in the fuel cycle (on line refueling); reduced requirements in the accelerator complex, which is more realistic and economical in today accelerators technology; and finally the utilization of He as coolant compared with molten Pb is more close to the proved technology given the know how of gas cooled reactors and more efficient from the thermodynamic point of view, allowing simplification and the utilization in other process, besides electricity generation, as hydrogen generation. (author)

  1. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

    Liquid Metal Cooled Fast Breeder Reactors (LMFBR) require mechanisms operating in various sodium liquid and sodium vapor environments for extended periods of time up to temperatures of 900 K under different chemical properties of the fluid. The design of tribological systems in those reactors cannot be based on data and past experience of so-called conventional tribology. Although basic tribological phenomena and their scientific interpretation apply in this field, operating conditions specific to nuclear reactors and prevailing especially in the nuclear part of such facilities pose special problems. Therefore, in the framework of the R and D-program accompanying the construction phase of SNR 300 experiments were carried out to provide data and knowledge necessary for the lay-out of friction systems between mating surfaces of contacting components. Initially, screening tests isolated material pairs with good slipping properties and maximum wear resistance. Those materials were subjected to comprehensive parameter investigations. A multitude of laboratory scale tests have been performed under largely reactor specific conditions. Unusual superimpositions of parameters were analyzed and separated to find their individual influence on the friction process. The results of these experiments were made available to the reactor industry as well as to factories producing special tribo-materials. (orig.)

  2. Risk Management for Sodium Fast Reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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.

  3. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    In 1985, the International Atomic Energy Agency (IAEA) published a report titled 'Status of Liquid Metal Cooled Fast Breeder Reactors' (Technical Reports Series No. 246). The report was a general review of the status of fast reactor development at that time, covering some aspects of design and operation and reviewing experience from the earliest days. It summarized the programmes and plans in all countries which were pursuing the development of fast reactors. In 1999, the IAEA published a follow-up report titled 'Status of Liquid Metal Cooled Fast Reactor Technology' (IAEA-TECDOC-1083), necessitated by the substantial advances in fast reactor technology development and changes in the economic and regulatory environment which took place during the period of 1985-1998. Chief among these were the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at a high burnup and the launch of new fast reactor programmes by some additional Member States. In 2006, the Technical Working Group on Fast Reactors (TWG-FR) identified the need to update its past publications and recommended the preparation of a new status report on fast reactor technology. The present status report intends to provide comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction, operation and decommissioning; various areas of research and development; engineering; safety; and national strategies and public acceptance of fast reactors.

  4. A review of the UK fast reactor programme, March 1979

    International Nuclear Information System (INIS)

    The Status report of the UK activities related to fast-breeder reactor activities includes the following: summary of the operating experience of the prototype Fast Reactor (PFR) during 1978; design studies of the commercial demonstration fast reactor (CDFR); design studies of later advanced LMFBR; engineering developments of high temperature sodium loop, steam generators and instrumentation; materials development; corrosion problems; sodium technology; fuel elements development; PFR fuel reprocessing; safety issues molten fuel-coolant interaction; core structure test; accident analysis; reactor performance studies; experimental reactor physics; fuel management and general neutronics calculation for CDFR; reactor instruments

  5. Fuel systems for compact fast space reactors

    International Nuclear Information System (INIS)

    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 UO2 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 UO2-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

  6. Fast reactor technology innovation and visualization

    International Nuclear Information System (INIS)

    Innovations in safety, operations, and maintenance for improving the availability, reliability, and capital cost of the sodium fast reactor are described. Concerning safety these innovations deal with on-line limiting safety settings, inherent core protection, detection of subassembly coolant mis-allocation. Concerning reactor operations these innovations deal with advanced energy conversion, adapting non-base load nuclear plants and on-line diagnostics. Other innovations concern inspection, servicing, refueling. The development of these innovations rely on visualization technology for their use and for demonstration of improvements achievable. A visualization platform for running these innovations and the nuclear plant thermal-hydraulic, structure, and process codes that underlie them are described. The platform hardware consists of a large-scale tiled display and a haptic hand-controller and in the future will grow to include a high-speed network and multiple graphics-client systems

  7. Safe Management Of Fast Reactors: Towards Sustainability

    International Nuclear Information System (INIS)

    An interdisciplinary systemic approach to socio-technical optimization of nuclear energy management is proposed, by recognizing a) the rising requirements to nuclear safety being realized using fast reactors (FR), b) the actuality to maintain and educate qualified workforce for fast reactors, c) the reactor safety and public awareness as the keystones for improving attitude to implement novel reactors. Knowledge management and informational support firstly is needed in: 1) technical issues: a) nuclear energy safety and reliability, b) to develop safe and economic technologies; 2) societal issues: a) general nuclear awareness, b) personnel education and training, c) reliable staff renascence, public education, stakeholder involvement, e).risk management. The key methodology - the principles being capable to manage knowledge and information issues: 1) a self-organization concept, 2) the principle of the requisite variety. As a primary source of growth of internal variety is considered information and knowledge. Following questions are analyzed indicating the ways of further development: a) threats in peaceful use of nuclear energy, b) basic features of nuclear risks, including terrorism, c) human resource development: basic tasks and instruments, d) safety improvements in technologies, e) advanced research and nuclear awareness improvement There is shown: public education, social learning and the use of mass media are efficient mechanisms forming a knowledge-creating community thereby reasoning to facilitate solution of key socio-technical nuclear issues: a) public acceptance of novel nuclear objects, b) promotion of adequate risk perception, and c) elevation of nuclear safety level and adequate risk management resulting in energetic and ecological sustainability. (author)

  8. Actinide management with commercial fast reactors

    International Nuclear Information System (INIS)

    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

  9. Actinide management with commercial fast reactors

    Science.gov (United States)

    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.

  10. The Argentine-Brazilian fast reactor programme

    International Nuclear Information System (INIS)

    This paper summarizes the Argentine-Brazilian Fast Reactor Programme and gives reasons for the decision of a binational venture. The work carried out by both countries is described, showing how they complement each other, with the corresponding saving of resources. The main objectives of the Programme and tentative schedules in three progressing integrating stages are given and the present nuclear know-how in each country is identified as a good starting point. The paper also gives some details regarding the economical and human resources involved. (author). 1 graph

  11. Actinide management with commercial fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ohki, Shigeo [Japan Atomic Energy Agency, 4002, Narita-cho, O-arai-machi, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

    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.

  12. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  13. Fast reactors and advanced light water reactors for sustainable development

    International Nuclear Information System (INIS)

    Complete text of publication follows: The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been re-acknowledged worldwide. In response to this international movement, the papers compiling the most recent findings in the fields of fast reactors (FR) and advanced light water reactors (LWR) were gathered and published in this special issue. This special issue compiles six articles, most of which are very meticulously performed studies of the multi year development of design and assessment methods for large sodium-cooled FRs (SFRs), and two are related to the fuel cycle options that are leading to a greater understanding on the efficient utilization of energy resources. The Japanese sodium-cooled fast reactor (JSFR) is addressed in two manuscripts. H. Yamano et al. reviewed the current design which adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. Their safety assessments of both design basis accidents and severe accidents indicate that the devised JSFR satisfies well their risk target. T. Takeda et al. discussed the improvement of the modeling accuracy for the detailed calculation of JSFR's features in three areas: neutronics, fuel materials, and thermal hydraulics. The verification studies which partly use the measured data from the prototype FBR Monju are also described. Two of these manuscripts deal with those aspects of advanced design of SFR that have hitherto not been explored in great depth. The paper by G. Palmiotti et al. explored the possibility of using the sensitivity methodologies in the reactor physics field. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described. F. Baque et al. reviewed the evolution of the in

  14. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants

  15. Research activities on fast reactors in Switzerland

    International Nuclear Information System (INIS)

    The current domestic Swiss electricity supply is primarily based on hydro power (approximately 61%) and nuclear power (about 37%). The contribution of fossil systems is, consequently, minimal (the remaining 2%). In addition, long-term (but limited in time) contracts exist, securing imports of electricity of nuclear origin from France. During the last two years, the electricity consumption has been almost stagnant, although the 80s recorded an average annual increase rate of 2.7%. The future development of the electricity demand is a complex function of several factors with possibly competing effects, like increased efficiency of applications, changes in the industrial structure of the country, increase of population, further automation of industrial processes and services. Due to decommissioning of the currently operating nuclear power plants and expiration of long-term electricity import contracts there will eventually open a gap between the postulated electricity demand and the base supply. The assumed projected demand cases, high and low, as well as the secured yearly electric energy supply are shown. The physics aspects of plutonium burning fast reactor configurations are described including first results of the CIRANO experimental program. Swiss research related to residual heat removal in fast breeder reactors is presented. It consists of experimental ana analytic investigations on the mixing between two horizontal fluid layers of different velocities and temperatures. Development of suitable computer codes for mixing layer calculation are aimed to accurately predict the flow and temperature distribution in the pools. A satisfactory codes validation based on experimental data should be done

  16. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    International Nuclear Information System (INIS)

    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.

  17. History of fast reactor fuel development

    Science.gov (United States)

    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.

  18. State of the art and prospects of fast neutron reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zrodnikov, A.V.; Mittenkov, F.M.; Poplavsky, V.M.; Kiryushin, A.I. [Physics and Power Eng. Inst., Obninsk (Russian Federation). State Sci. Centre

    1997-10-01

    On the basis of experience of fast reactor design, construction and operation gained in Russia, this paper outlines their state of the art. The high maturity and efficiency of this type of nuclear power development in Russia and the equalization of the economic characteristics of thermal and fast reactors is shown, as well as the expediency of improvement of nuclear power environmental characteristics owing to fast reactors incorporation. (orig.) 7 refs.

  19. Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

    Energy Technology Data Exchange (ETDEWEB)

    Tucek, Kamil [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)]. E-mail: kamil.tucek@jrc.nl; Carlsson, Johan [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands); Wider, Hartmut [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)

    2006-08-15

    A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies. In this paper, two fast reactor systems are discussed-the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries. First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600 MW{sub e}) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems. We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating

  20. Status of National Programmes on Fast Breeder Reactors. International Working Group on Fast Reactors Twenty-First Annual Meeting, Seattle, USA, 9-12 May 1988

    International Nuclear Information System (INIS)

    The following papers on the status of national programmes on fast breeder reactors are presented in this report: Fast breeder reactor development in France during 1987; Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands; A review of the Indian fast reactor programme; A review of the Italian fast reactor programme; A review of the fast reactor programme in Japan; Status of fast reactor activities in the USSR; A review of the United Kingdom fast reactor programme; Status of liquid metal reactor development in the United States of America; Review of activities of the Commission of European Communities relating to fast reactors in 1987; European co-operation in the field of fast reactor research and development — 1987 progress report; A review of fast reactor activities in Switzerland

  1. LFR "Lead-Cooled Fast Reactor"

    Energy Technology Data Exchange (ETDEWEB)

    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

  2. Indian fast reactor technology: Current status and future programme

    Indian Academy of Sciences (India)

    S C Chetal; P Chellapandi

    2013-10-01

    The paper brings out the advantages of fast breeder reactor and importance of developing closed nuclear fuel cycle for the large scale energy production, which is followed by its salient safety features. Further, the current status and future strategy of the fast reactor programme since the inception through 40 MWt/13 MWe Fast Breeder Test Reactor (FBTR), is highlighted. The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast Reactor (SFRs) and timely introduction of metallic fuel reactors with emphasis on breeding gain and enhanced safety are being brought out in this paper.

  3. Direct Energy Conversion for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.; Cooper, J.; Vogt, D.; Chapline, G.; Turchi, P.; Barbee Jr., T.; Farmer, J.

    2000-07-01

    Strategic Computing Initiative (ASCI), should improve the speed and decrease the cost of developing new TEGs. The system concept to be evaluated is shown in Figure 1. Liquid metal is used to transport heat away from the nuclear heat source and to the TEG. Air or liquid (water or a liquid metal) is used to transport heat away from the cold side of the TEG. Typical reactor coolants include sodium or eutectic mixtures of lead-bismuth. These are coolants that have been used to cool fast neutron reactors. Heat from the liquid metal coolant is rejected through the thermal electric materials, thereby producing electrical power directly. The temperature gradient could extend from as high as 1300 K to 300 K, although fast reactor structural materials (including those used to clad the fuel) currently used limit the high temperature to about 825K.

  4. Sodium technology for fast breeder reactors

    International Nuclear Information System (INIS)

    Sodium, because of its good heat transfer and nuclear properties, is used as a coolant in fast reactors. It is also used largely as a reducing agent in pharmaceutical, perfumery and general chemical industries. Its affinity to react with air and water is a strong disadvantage. However, this is fully understood and the design of engineering systems take care of this aspect. With several experimental and test facilities established over the years in this country as well as abroad, the 'sodium technology' has reached a level of maturity. The design of sodium systems considering all the physical and chemical properties and the developmental work carried out at Indira Gandhi Centre for Atomic Research are broadly covered in this report. (author)

  5. Trial visualization of fast reactor design knowledge

    International Nuclear Information System (INIS)

    In design problems of large-scale systems like fast breeder reactors, inter-relations among design specifications are very important where a selected specification option is transferred to other specification selections as a premise to be taken account in engineering judgments. These inter-relations are also important in design case studies with the hypothetical adoption of rejected design options for the evaluation of deviation propagations among design specifications. Some of these rejected options have potential worth for future reconsideration by some circumstance changes (e.g., advanced simulations to exclude needs for mock-up tests, etc.), to contribute to flexibility in system designs. In this study, a computer software is built to visualize a design problem structure by representing engineering knowledge nodes on individual specification selections along with inter-relations of design specifications, to validate the knowledge representation method and to derive open problems. (author)

  6. Fast Reactor Fuel Development in Japan

    International Nuclear Information System (INIS)

    The future fast reactor and its fuel cycle system under development in Japan uses oxide fuel with simplified pelletizing fuel fabrication technology as a reference concept. Its driver fuel consists of large diameter annular fuel pellets, oxide dispersion strengthened ferritic steel cladding fuel pins with a ferritic-martensitic steel subassembly wrapper tube and minoractinide- bearing oxide fuel. The target burnup of the driver fuel is 150 GW.d/t in discharge average, which corresponds to 250 GW.d/t of peak burnup and 250 dpa of peak neutron dose. Fuel developmental efforts, including out-of-pile studies such as material characteristics experimental evaluation and fuel property measurements, various irradiation tests and fuel fabrication technology developments were planned and are in progress. Future fuels will be realized through Joyo irradiation tests and Monju demonstrations. International collaborative efforts are also an important part of such activities. (author)

  7. Safeguards in Prototype Fast Breeder Reactor Monju

    International Nuclear Information System (INIS)

    The assemblies loaded in the core and stored in the ex-vessel storage tank (EVST) are in liquid sodium in the Japanese prototype fast breeder reactor (FBR) Monju. Since it is difficult to apply a direct verification procedure for the fuel assemblies in these areas, a dual containment and surveillance system consisting of two monitoring devices such as surveillance camera and radiation monitor that are functionally independent has been applied. In addition, the Monju Remote Monitoring System was developed to strengthen the continuous surveillance and to reduce the load of the inspection activities. Furthermore, the ex-vessel transfer machine radiation monitor (EVRM) and the exit gate monitor (EXGM) were upgraded to strengthen the monitoring of spent blanket fuel assemblies and to improve the reliability of distinguishing between fuel assemblies and non-fuel items. As the result, the integrated safeguards was introduced in November 2009, and the effective safeguards activities have been implemented in Monju. (author)

  8. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; 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 event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions 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.

  9. Methane reforming with fast nuclear reactor steam

    International Nuclear Information System (INIS)

    The paper considers the concept of utilizing nuclear fast reactor (FR) with a sodium coolant for methane steam reforming. Steam conditions of a power FR, e.g. the BN-600 now operating in Russia: steam pressure P=13.2 MPa and steam temperature T=500degC, do not absolutely comply with the catalytic reactor working parameters, which produces a synthetic gas (syngas), a mix of hydrogen and carbon oxide. In this connection, the present paper addresses a possibility of utilizing steam produced in one of three independent the BN-600 loops in an amount of 640 t/h for preparing a gas-steam mixture with T=500degC and its additional heating in a converter up to the operating temperature, T=850degC, at the expense of natural gas burning or electrical energy supplying. In this case, the fraction of burned natural gas burning or electrical energy supplying. In this case, the fraction of burned natural gas significantly decreases. It is estimated that steam parameters of the BN-600 afford to obtain ∼3·105 nm3/h of hydrogen. It is also considered a concept of nuclear heat transfer to remote regions to be achieved with the aid of syngas incoming from the converter, its cooling further and transmitting through a pipeline to the place of its utilization, where it is restored into methane with the heat extraction. (author)

  10. A review of fast reactor progress in Japan

    International Nuclear Information System (INIS)

    The fast reactor development project in Japan is continuing at a slightly increased scale of effort in budget. The total budget for LMFBR development for fiscal year 1978 was 24 billion yen. In August 1977 major industries engaged in LMFBR have set up an office where design work can be jointly conducted. Highlights and topics of the fast reactor development activities cover description of JOYO reactor, its first criticality experiment, and the prototype fast breeder MONJU. Research and development programmes dealt with fission products release and its possible interaction with the soodium coolant, inspection of reactor components, experiments simulating sodium leakage, development of steam generator

  11. Implications of Fast Reactor Transuranic Conversion Ratio

    Energy Technology Data Exchange (ETDEWEB)

    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

  12. Development of studies on helium cooled fast reactors

    International Nuclear Information System (INIS)

    A necessity is shown of developing breeders with high reproductive properties. Helium cooled fast reactor is considered. The reactor performances, heating circuit with the use of a steam turbine unit in the secondary circuit is outlined. The reactor design and fuel assemblies are described

  13. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing fast reactor electricity at an affordable and competitive price.

  14. A review of the U.K. fast reactor programme: March 1978

    International Nuclear Information System (INIS)

    The review of the UK fast reactor programme covers the description of Dounreay Fast Reactor shut down after seventeen years of successful operation; description of prototype fast reactor (PFR); core design parameters safety features and plant design for commercial demonstration fast reactor (CDFR). Engineering development is related to large sodium rigs, coolant circuit hydraulics and vibration, instrumentation and components. The subjects of interest are material development, sodium technology, fast reactor fuel, fuel cycle, reactor safety, reactor performance studies

  15. The United States of America fast breeder reactor program

    International Nuclear Information System (INIS)

    The reasons for the development of the fast breeder reactor in the United States are outlined, and the LMFBR program is discussed in detail, under the following headings: program objectives, reactor physics, fuel and materials development, fuel recycle, safety, components, plant experience program (Near Commercial Breeder Reactor). The special facilities to be used at each stage of the program are described. It is planned that the Near Commercial Breeder Reactor will be complete in 1986, and commercial plants should follow in rapid succession. An alternate fast reactor concept (Gas Cooled Fast Reactor) is outlined. The Environmental Impact Statement for the proposed program is summarized, and the cost benefit analysis supplied as part of the Environment Statement is also summarized. (U.K.)

  16. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  17. Neutron spectrometer for fast nuclear reactors

    CERN Document Server

    Osipenko, M; Ricco, G; Caiffi, B; Pompili, F; Pillon, M; Angelone, M; Verona-Rinati, G; Cardarelli, R; Mila, G; Argiro, S

    2015-01-01

    In this paper we describe the development and first tests of a neutron spectrometer designed for high flux environments, such as the ones found in fast nuclear reactors. The spectrometer is based on the conversion of neutrons impinging on $^6$Li into $\\alpha$ and $t$ whose total energy comprises the initial neutron energy and the reaction $Q$-value. The $^6$LiF layer is sandwiched between two CVD diamond detectors, which measure the two reaction products in coincidence. The spectrometer was calibrated at two neutron energies in well known thermal and 3 MeV neutron fluxes. The measured neutron detection efficiency varies from 4.2$\\times 10^{-4}$ to 3.5$\\times 10^{-8}$ for thermal and 3 MeV neutrons, respectively. These values are in agreement with Geant4 simulations and close to simple estimates based on the knowledge of the $^6$Li(n,$\\alpha$)$t$ cross section. The energy resolution of the spectrometer was found to be better than 100 keV when using 5 m cables between the detector and the preamplifiers.

  18. Creep buckling problems in fast reactor components

    International Nuclear Information System (INIS)

    Creep buckling analyses for two important components of 500 M We Prototype Fast Breeder Reactor (PFBR), viz. Intermediate Heat Exchanger (IHX) and Inner Vessel (IV), are reported. The INCA code of CASTEM system is used for the large displacement elasto-plastic-creep analysis of IHX shell. As a first step, INCA is validated for a typical benchmark problem dealing with the creep buckling of a tube under external pressure. Prediction of INCA is also compared with the results obtained using Hoff's theory. For IV, considering the prohibitively high computational cost for the actual analysis, a simplified analysis which involves only large displacement elastoplastic buckling analysis is performed using isochronous stress strain curve approach. From both of these analysis is performed using isochronous stress strain curve approach. From both of these analysis, it has been inferred that creep buckling failure mode is not of great concern in the design of PFBR components. It has also been concluded from the analysis that Creep Cross Over Curve given in RCC-MR is applicable for creep buckling failure mode also. (author). 8 refs., 9 figs., 1 tab

  19. Immobilization of Fast Reactor First Cycle Raffinate

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Accuracy of helium accumulation fluence monitor for fast reactor dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Chikara; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    A helium (He) accumulation fluence monitor (HAFM) has been developed for fast reactor dosimetry. In order to evaluate the measurement accuracy of neutron fluence by the HAFM method, the HAFMs of enriched boron (B) and beryllium (Be) were irradiated in the Fast Neutron Source Reactor `YAYOI`. The number of He atoms produced in the HAFMs were measured and compared with the calculated values. As a result of this study, it was confirmed that the neutron fluence could be measured within 5 % by the HAFM method, and that met the required accuracy for fast reactor dosimetry. (author)

  1. Analysis of fast reactor scenario with different conversion ratios

    International Nuclear Information System (INIS)

    Korean fast reactor scenarios have been analyzed for various kinds of conversion ratio by the DANESS system dynamic analysis code. The once-through fuel cycle analysis was modeled based on the Korean 'National Energy Basic Plan' up to 2030 and a postulated nuclear demand growth rate until 2150. The fast reactor scenario analysis has been performed for three kinds of conversion ratios such as 0.3, 0.61 and 1.0. Through the calculations, the nuclear reactor deployment scenario, front-end cycle, back-end cycle, and long-term heat load have been investigated. From the once-through results, it is shown that the nuclear power demand would be ∼70 GWe and the total amount of the spent fuel accumulated by 2150 would be ∼168000 t. Also, the fast reactor scenario analysis results show that the spent fuel inventory and out-pile transuranic element can be reduced by increasing the fast reactor conversion ratio. Furthermore, the long-term heat load of spent fuel decreases with increasing the conversion ratio. However, it is known that the deployment of a fast reactor of low conversion ratio does not much reduce the spent fuel and out-pile transuranic element inventory due to the fast reactor deployment limitation which is related to the availability of transuranic elements. (author)

  2. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    The fast breeder reactor development project in Japan made progress in the past year, and will be continued in the next fiscal 1981. The scale of efforts both in budget and personnel will be similar to those in fiscal 1980. The budget for R and D works and for the construction of the fast breeder prototype reactor ''Monju'' will be approximately 20 billion yen and 27 billion yen, respectively, excluding the wage of the personnel concerned. The number of the technical personnel currently engaging in fast breeder reactor development in the Power Reactor and Nuclear Fuel Development Corp. is about 530. As for the experimental fast reactor ''Joyo'', three operational cycles at 75 MWt have been completed in August, 1980, and the fourth cycle has started in March, 1981. As for the prototype reactor ''Monju'', progress was made toward the construction, and the environmental impact statement on the reactor was approved by the authorities concerned. The studies on the preliminary design of large LMFBRs have been made by the PNC and also by power companies. The design study carried out by the PNC is concerned with a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of the commissioning of ''Monju''. The highlights and topics in the development activities for fast breeder reactors in the past twelve months are summarized in this report. (Kako, I.)

  3. Integral physics data for fast-reactor design

    International Nuclear Information System (INIS)

    Integral physics data for fast-reactor design. The recent compilation of the section on fast-reactor physics for the forthcoming second edition of 'Reactor Physics Constants' has necessitated a survey of the available experimental integral data. The choice of fast-reactor-physics integral data to be included in the compilation was based upon two criteria besides availability: (a) the data arise from relatively simple systems which lend themselves to simple theoretical analyses; and (b) complicated systems representing prototypes or mock-ups having general interest in terms of fast-power reactors. The first criterion was decided upon so as to list integral data for those systems of most general utility for the verification of cross-section parameters and calculational procedures. The second criterion is based upon presentation of current data on actual fast power breeder reactor systems. These are too complicated for simple theoretical analysis. They demonstrate the complexity of the actual reactor versus the more idealized and easily analysed critical experiment. Integral physics data for reactor design refer to measurements on reactor systems, critical or otherwise, of the various reactor physics quantities of practical and/or theoretical importance. These characterize and lead to an understanding of the system. The measurements are represented by critical mass, core shape factor, detector ratios, neutron spectra, material replacement experiments, reflector savings, neutron lifetime, Rossi-α, and similar quantities. These data are reviewed and the range of applicability is described. Limitations of experimental and analytical results are shown to exist in certain spectral and criticality analyses. Experimental and analytical investigations are suggested for future work. These will tend to narrow the gap between theory and experiment on 'known' systems. They also include investigations to 'firm up' the physics of large conceptual, fast power-breeder reactor

  4. Coupled hydro-neutronic calculations for fast burst reactor accidents

    International Nuclear Information System (INIS)

    Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by either single point kinetics or in a decoupled manner. This methodology is enabled by the use of modem supercomputers (CM-200). Two examples of this capability are presented: an unreflected metal fast burst assembly, and a reflected fast burst assembly typical of the Skua or SPR-III class of fast burst reactor

  5. An option for the Brazilian nuclear project: necessity of fast breeder reactors and core design for an experimental fast reactor

    International Nuclear Information System (INIS)

    Aiming to assure the continued utilization of fission energy, the development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed, as the best type for the Brazilian nuclear system in the future. The inherent safety characteristics are superior to current fast breeder reactors and an efficient utilization of thorium can be realized. The construction and operation of an experimental fast reactor is the first step and a basic tool for the development of FBRs technologies. A serie of core design for an 90 MW FBR is studied and the possible options and sizes of the main parameters are identified. (E.G.)

  6. Recent progress of Gas Fast Reactor program

    International Nuclear Information System (INIS)

    The GFR is considered by the French Atomic Energy Commission as a promising concept which combines the benefits of fast spectrum and high temperature, using helium as coolant. He properties are interesting with respect to safety: it is single phase (no threshold effect due to phase changing), chemically inert, and non toxic. It affords an optical transparency allowing potential improvements in temperature measurement, management for dismantling, and in-service-inspection. The voiding effect is limited, less than 1$, providing quasi- decoupling of the reactor physics from the state of the coolant. Nevertheless, Helium is a poor coolant, so that the GFR viability includes development of a refractory and dense fuel, and robust management of accidental transients, especially cooling accidents. GFR feasibility is essentially linked to three demonstrations: the feasibility (fabrication, thermo-mechanical behaviour) of a refractory fuel; the safety architecture with appropriate systems for the prevention and a robust mitigation of accidental scenarios (especially depressurization); economic competitiveness. The first one includes an experimental activity at the laboratory scale: completion of the results is expected by 2012-2015. The next step afterward will be the design, construction and the operation of a 50-100 MWth experimental reactor, the Allegro project (former ETDR), possibly as a European Joint Undertaking. The full paper will recall the 2007 design choices and it will give an overview of the progress performed so far regarding the safety architecture and the safety evaluation. The 2007 reference fuel technology is a ceramic plate type fuel element. It combines a high enough core power density (minimization of the Pu inventory), plutonium and minor actinides recycling capabilities. Innovative to many aspects, the fuel element is a key issue in the GFR feasibility. It is supported already by a significant R and D effort also applicable to a pin concept that is

  7. Review of fast reactor operating experience gained in 1998 in Russia. General trends of future fast reactor development

    International Nuclear Information System (INIS)

    Review of the general state of nuclear power in Russia as for 1998 is given in brief in the paper. Results of operation of BR-10, BOR-60 and BN-600 fast reactors are presented as well as of scientific and technological escort of the BN-350 reactor. The paper outlines the current status and prospects of South-Urals and Beloyarskaya power unit projects with the BN-800 reactors. The main planned development trends on fast reactors are described concerning both new projects and R and D works. (author)

  8. New modelling method for fast reactor neutronic behaviours analysis

    International Nuclear Information System (INIS)

    Due to safety rules running on fourth generation reactors' core development, neutronics simulation tools have to be as accurate as never before. First part of this report enumerates every step of fast reactor's neutronics simulation implemented in current reference code: ECCO. Considering the field of fast reactors that meet criteria of fourth generation, ability of models to describe self-shielding phenomenon, to simulate neutrons leakage in a lattice of fuel assemblies and to produce representative macroscopic sections is evaluated. The second part of this thesis is dedicated to the simulation of fast reactors' core with steel reflector. These require the development of advanced methods of condensation and homogenization. Several methods are proposed and compared on a typical case: the ZONA2B core of MASURCA reactor. (author)

  9. Shuffling strategy study of breeding-burning integrated fast reactor

    International Nuclear Information System (INIS)

    The breeding-burning integrated fast reactor uses burning assemblies to generate thermal power, meanwhile, converts 238U into 239Pu in the fertile assemblies. With periodical shuffling of assemblies, the reactor can maintain criticality for decades of years. To maintain long-term stability of the core reactivity, the core layout and shuffling strategy should balance the burning and the breeding of the assemblies. The scattered core layout and shuffling strategy ensures fast breeding of the fertile assemblies, and keeps stable core power distribution in whole life of the reactor. Moreover, at the end of the reactor life, the discharge burnups of different fuel assemblies are close to each other, which are about 250300 GW · d/t. This is important for breeding-burning integrated fast reactor to achieve very efficient utilization of uranium resource without reprocessing. (authors)

  10. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    The fast reactor programme in the United Kindom is reviewed under the following headings: Progress with PFR; Reprocessing: Commercial Design Studies; Structural Integrity; Engineering and Components; Materials; Sodium Chemistry; Core and Fuel; Safety; Plant Performance. (author)

  11. Advances in sodium technology, testing and diagnostics of fast reactors

    International Nuclear Information System (INIS)

    The collection contains a selection of 29 papers from three international specialists' meetings: the CMEA conference ''Control and measuring instruments and diagnostic systems of fast reactors'' held in the GDR in April 1983; the IAEA conference on nuclear power experience held in Austria in September 1982; and the conference ''Problems of technology and corrosion in sodium coolant and protective gas'' held in the GDR in April 1977. Three papers on operating experience with Soviet fast reactors and their safety have a general character; they are followed up by three papers on sodium technology. Five papers deal with the diagnostics of fast sodium cooled reactors and nine papers are devoted to the diagnostics of steam generators. Eight papers relate to detectors for the diagnostics of fast reactors. Safety regulations for work with alkali metals are added. (A.K.)

  12. Recycle Strategies for Fast Reactors and Related Fuel Cycle Technologies

    International Nuclear Information System (INIS)

    Fast reactors and related fuel cycle (hereafter referred to as 'fast reactor cycle') technologies have the potential to contribute to long term energy security owing to their effective use of uranium and plutonium resources, and to a reduction in the heat generation and potential toxicity of high level radioactive wastes by burning long lived minor actinides recovered from spent fuel from light water reactors and fast reactors. Further, it is likely that fast reactor cycle technologies can play a certain role in non-proliferation as addressed in the Global Nuclear Energy Partnership. With these features, the research and development towards their commercialization has been promoted vigorously and globally as a future vision of nuclear energy. The introduction of fast reactor cycle systems will be carried out independently in each country according to its national conditions and nuclear energy policy. It should then be considered important to have a globally common consensus relating to safety philosophy, concepts of proliferation resistance, transuranic element burnup and recycling and so on. For the development and utilization of fast reactor cycle systems, while respecting each country's concept, it is essential to organize the technologies and concepts which countires should have in common globally and build a framework to make them standardized. The use of existing frameworks such as the Generation IV International Forum and the International Project on Innovative Nuclear Reactors and Fuel Cycles is considered effective to achieving this. Furthermore, a vigorous promotion such as international cooperative developments enables the formation of international consensus on major technologies for the fast reactor cycle as well as the saving of resources by infrastructure sharing. (author)

  13. Numerical modelling of the CHEMREC black liquor gasification process. Conceptual design study of the burner in a pilot gasification reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marklund, Magnus

    2001-02-01

    The work presented in this report is done in order to develop a simplified CFD model for Chemrec's pressurised black liquor gasification process. This process is presently under development and will have a number of advantages compared to conventional processes for black liquor recovery. The main goal with this work has been to get qualitative information on influence of burner design for the gas flow in the gasification reactor. Gasification of black liquor is a very complex process. The liquor is composed of a number of different substances and the composition may vary considerably between liquors originating from different mills and even for black liquor from a single process. When a black liquor droplet is gasified it loses its organic material to produce combustible gases by three stages of conversion: Drying, pyrolysis and char gasification. In the end of the conversion only an inorganic smelt remains (ideally). The aim is to get this smelt to form a protective layer, against corrosion and heat, on the reactor walls. Due to the complexity of gasification of black liquor some simplifications had to be made in order to develop a CFD model for the preliminary design of the gasification reactor. Instead of modelling droplets in detail, generating gas by gasification, sources were placed in a prescribed volume where gasification (mainly drying and pyrolysis) of the black liquor droplets was assumed to occur. Source terms for the energy and momentum equations, consistent with the mass source distribution, were derived from the corresponding control volume equations by assuming a symmetric outflow of gas from the droplets and a uniform degree of conversion of reactive components in the droplets. A particle transport model was also used in order to study trajectories from droplets entering the reactor. The resulting model has been implemented in a commercial finite volume code (AEA-CFX) through customised Fortran subroutines. The advantages with this simple

  14. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The twenty-second Annual Meeting of the International Working Group on Fast Reactors took place in Vienna, 18-21 April 1989. Nineteen representatives from twelve Member States and International Organizations attended the Meeting. This publication is a collection of presentations in which the participants reported the status of their national programmes on fast breeder reactors. A separate abstract was prepared for each of the twelve papers from this collections. Refs, figs, tabs and 1 graph

  15. Improve Design of Fuel Shear for Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    GAO; Wei; OUYANG; Ying-gen; LI; Wei-min

    2012-01-01

    <正>Due to the deeper burnup and higher fuel swelling, fast reactor metal fuel rod using 316 stainless steel cladding, replacing the traditional zirconia cladding. The diameter of fuel rod of fast reactor is much longer than that of PWR, and the cladding of stainless steel has better ductility than zirconia cladding. Using the existing shear still will cause several aspects of problem: 1) Longer diameter of rod leads to

  16. Methods for quantifying uncertainty in fast reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. High burnup fast reactor fuel: processing and waste management experiences

    International Nuclear Information System (INIS)

    The routine processing of mixed Plutonium/Uranium oxide fuels from the Prototype Fast Reactor (PFR) at Dounreay began in September 1980 and the design features of the modified Dounreay Fast Reactor (DFR) reprocessing plant and experience of the first active campaign were described in a paper to the British Nuclear Engineering Society in November 1981 (1). Since then progress in processing the fuel discharged from PFR has been covered briefly in a number of papers to international conferences and the Public Inquiry held in 1986 into the outline planning application for the proposed European Demonstration Reprocessing Plant. During this decade considerable experience in the operation of fast reactors and associated fuel plants has been accumulated providing confidence in the system before entering the next development phase - that of its commercial demonstration. Confidence in the UK draws on the successful operation of the PFR and the associated Dounreay fuel reprocessing and BNF Sellafield fabrication plants. Of equal importance is public confidence in safe operation and in the management of wastes generated by a fast reactor system. The present paper is a review of fast reactor reprocessing and waste management at the Dounreay Nuclear Establishment (DNE) as a contribution to the present status of the fast reactor system

  18. Progress of Research on Demonstration Fast Reactor Main Pipe Material

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The main characteristics of the sodium pipe system in demonstration fast reactor are high-temperature, thin-wall and big-caliber, which is different from the high-pressure and thick-wall of the pressurized water reactor system, and the system is long-term

  19. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  20. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  1. CP ESFR: Collaborative Project for a European Sodium Fast Reactor

    International Nuclear Information System (INIS)

    The Collaborative Project for a European Sodium Fast Reactor (CP ESFR) is performed (2009-2012) in the 7th European Framework Programme. It is devoted to the identification and study of innovations to be considered for the future in the core design, safety, reactor architecture, components and the dissemination of knowledge related to this technology among young European professionals. (author)

  2. Recycle strategies for fast reactors and related fuel cycle technologies

    International Nuclear Information System (INIS)

    Full text: 1. Introduction Fast reactors and related fuel cycle (hereinafter referred to as 'Fast reactor cycle') technologies have the potential of contributing to long-term energy security due to effective use of uranium and plutonium resources, and reduction of the heat generation and potential toxicity of high-level radioactive wastes by burning long-lived minor actinides (MA) recovered from spent fuels of light-water reactors and fast reactors. Further, it is likely that fast reactor cycle technologies can play a certain role in non- proliferation as addressed in GNEP (Global Nuclear Energy Partnership). With these features, R and Ds toward their commercialization have been promoted vigorously and globally as a future vision of nuclear energy. 2. Recycle strategies in each country In Japan, it is determined that after burning uranium in light water reactors, plutonium is recovered from spent fuel and used for light water reactors at the moment and for fast reactors in the future. In order to make it possible, Fast Reactor Cycle Technology Development (FaCT) Project has been promoted with a combination of oxide-fueled sodium-cooled reactors, advanced aqueous reprocessing, and simplified pelletizing fuel fabrication adopted as a main concept aiming at startup of a demonstration reactor around 2025 and commercialization before around 2050. In France, a comparison of the basic specifications between an oxide-fueled sodium-cooled reactor and a carbide (or nitride)-fueled gas-cooled reactor has currently been promoted towards technological selection for a prototype reactor in 2012 in accordance with 'The 2006 planning act on the sustainable management of radioactive materials and waste (Act 2006- 739)' enacted in 2006. Based on the results, France aims at startup of the prototype reactor in 2020 and commercialization in around 2040. For reprocessing, methods which extract actinides collectively such as GANEX has been developed to enhance proliferation resistance

  3. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  4. Fast reactors: the future of nuclear energy

    International Nuclear Information System (INIS)

    The main problems to be solved for FBR type reactors become viable economically, presenting the research programs of Europe, United States of America, Japan and Brazil are described. The cooperations between interested countries for improving FBR type reactors, and the financial and human resources necessaries for the development of programs, are evaluated. The fuel cycle is also analysed. (M.C.K.)

  5. Status of national programmes on fast reactors in Korea

    International Nuclear Information System (INIS)

    The role of nuclear power plants in electricity generation in Korea is expected to become more important in the years to come due to poor natural resources and green house gases. This heavy dependence on nuclear power eventually raises the issues of efficient utilization of uranium resources and of spent fuel storage. Fast reactors can resolve these issues. Korea Atomic Energy Research Institute started development of a Liquid Metal Reactor design in 1997 and completed the Conceptual Design in March of 2002. Efforts are currently directed toward the development of advanced fast reactor concepts and basic key technologies. (author)

  6. Review of the United Kingdom fast reactor programme - March 1986

    International Nuclear Information System (INIS)

    The UK programme in the field of fast reactors has continued successfully towards the following main objectives, details of which are contained in subsequent sections of this report: (2) progress with the prototype fast reactor (PFR) which achieved its design power on 4 March 1985; (3) nuclear fuel reprocessing; (4) commercial design studies; (5) structural integrity of LMFBR during its lifetime; (6) R and D work on components of LMFBR; (7) materials study; (8) sodium chemistry; (9) reactor core and fuel design philosophy; (10) safety problems; (11) plant performance studies

  7. Fast Reactor Physics Vol. I. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

    International Nuclear Information System (INIS)

    Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  8. Fast Reactor Physics. Vol. II. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

    International Nuclear Information System (INIS)

    Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  9. Regenerative burner

    Energy Technology Data Exchange (ETDEWEB)

    Davies, T.E.; Quinn, D.E.; Watson, J.E.

    1986-08-05

    A regenerative burner is described operable in fire and flue modes comprising: a burner shell having first and second internal chambers, the first chamber being disposed on the flame axis of the burner and the second chamber surrounding the radial perimeter of the first chamber; a gas permeable annular regenerative bed separating the first and second chambers such that gas flow between the first and second chambers must travel through the regenerative bed in a generally radial direction with respect to the flame axis; means for supplying combustion air to the second chamber when the burner is in the fire mode and for exhausting the products of combustion from the second chamber when the burner is in the flue mode; and means for supplying fuel in the vicinity of the flame axis for mixing with combustion air to support combustion when the burner is in the fire mode.

  10. Research on the usage of a deep sea fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira; Kowata, Yasuki [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-09-01

    Many new types of fast reactors have been studied in PNC. A deep sea fast reactor has the highest realization probability of the reactors studied because its development is desired by many specialists of oceanography, meteorology, deep sea bottom oil field, seismology and so on and because the development does not cost big budget and few technical problems remain to be solved. This report explains the outline and the usage of the reactor of 40 kWe and 200 to 400 kWe. The reactor can be used as a power source at an unmanned base for long term climate prediction and the earth science and an oil production base in a deep sea region. On the other hand, it is used for heat and electric power supply to a laboratory in the polar region. In future, it will be used in the space. At the present time, a large FBR development plan does not proceed successfully and a realization goal time of FBR has gone later and later. We think that it is the most important to develop the reactor as fast as possible and to plant a fast reactor technique in our present society. (author)

  11. Fast-power-reactor optimization by the game theory

    International Nuclear Information System (INIS)

    In the first stage of the use of fast breeder reactor - because fissile-material amounts are small - we are interested in fast breeder reactors which achieve minimum fissile-material mass, with maximum power. This problem shows a two-matrix-game structure. First, we determine a competive-game solution and second, a cooperative-game solution, obtaining in this way the optimum distribution of the fissile and fertile materials in the multizone fast reactors. Another optimization problem which is solved in this paper is finding the reactor structure for which the power non-uniformity factor and the flux non-uniformity factor are minimum. This is, also, a mathematical two-matrix game and it is solved as above. The two optimization problems have different solutions. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

    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.

  13. Regenerative burner

    Energy Technology Data Exchange (ETDEWEB)

    Gitman, G.M.

    1990-05-08

    This patent describes a method of combusting fuel in a furnace having a pair of regenerative burners, each burner having a combustion chamber. It comprises: supplying fuel and oxygen alternatively to each burner to create alternating firing burners wherein the oxygen is supplied from two sources providing first and second oxidizing gases having different oxygen concentrations and simultaneously alternating the application of negative pressure to the remaining non-firing burner to recover heat from flue gases exhausted by the regenerative bed of the non-firing burner to be used further to preheat at least part of the oxygen being supplied to the firing burner; mixing the fuel with a fraction of the oxygen under substoichiometric combustion condition to create products of incomplete combustion to form a hot, luminous flame core containing partially pyrolized fuel; and mixing the partially pyrolyzed fuel with a remaining fraction of the oxygen to complete combustion of the pyrolized fuel; and controlling the total flow of fuel and oxygen supplied to each burner to provide each burner with a desired flame stoichiometry.

  14. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

    An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

  15. Assumed mode approach to fast reactor core seismic analysis

    International Nuclear Information System (INIS)

    The need for a time history approach, rather than a response spectrum approach, to the seismic analysis of fast breeder reactor core structures is described. The use of a Rayleigh-Ritz/Assumed Mode formalism for developing mathematical models of reactor cores is presented. Various factors including structural nonlinearity, fluid inertia, and impact which necessitate abandonment of response spectrum methods are discussed. The use of the assumed mode formalism is described in some detail as it applies to reactor core seismic analysis. To illustrate the use of this formal approach to mathematical modeling, a sample reactor problem with increasing complexities of modeling is presented. Finally, several problem areas--fluid inertia, fluid damping, coulomb friction, impact, and modal choice--are discussed with emphasis on research needs for use in fast reactor seismic analysis

  16. OECD Nuclear Energy Agency Activities Related to Fast Reactor Development

    International Nuclear Information System (INIS)

    The OECD Nuclear Energy Agency (NEA), whose role is to assist its member countries to develop, through international cooperation, the scientific and technological bases required for the safe, environmentally friendly and economical use of nuclear energy, conducts work related to fast reactor systems in two areas of activity: one focused on scientific research and technology development needs and one dedicated to strategic and policy issues. Recent, scientifically oriented, fast reactor related activities coordinated by the NEA comprise: -A coordinated effort to evaluate basic nuclear data needed for the development of fast reactor systems; -A recently initiated review of Integral Experiments for Minor Actinide Management; -An ongoing study on Homogeneous versus Heterogeneous Recycle of Transuranic Isotopes in Fast Reactors; -A comparative analysis of the safety characteristics of sodium cooled fast reactors; -A series of workshops on Advanced Reactors with Innovative Fuels; -A series of information exchange meetings on actinide and fission product partitioning and transmutation. The NEA has also conducted two reviews on issues related to the transition from thermal to fast neutron nuclear systems. One study was devoted to technical issues, including benchmark studies on: (i) the performance of scenario analysis codes, (ii) a regional (European) scenario and (iii) a global transition scenario. The other study emphasized issues of interest to policymakers, such as key parameters affecting the cost-benefit analysis of transitioning, including the size and age of the nuclear reactor fleet, the expected future reliance on nuclear energy, access to uranium resources, domestic nuclear infrastructure and technology development, and radioactive waste management policy in place. The NEA is also an active player in many other international activities related to fast neutron systems, such as the Generation IV International Forum, where the NEA acts as technical secretariat for

  17. A Review of the UK Fast Reactor Programme: March 1980

    International Nuclear Information System (INIS)

    Towards the end of 1979 the Government announced a new programme of thermal reactor stations to be built over ten years (totalling 15GW), in addition to the two AGR stations at Torness and Heysham 'B' which had been approved by the previous Government. The first station of the new programme will be based on a Westinghouse PWR, subject to safety clearance and the outcome of a public inquiry, and it is envisaged that the remaining stations of the programme would be split between PWRs and AGRs. The AEA Chairman wrote formally to the Secretary of State for Energy in December 1979, putting forward on behalf of the Electricity Supply Authorities, NNC, BNFL and the AEA a recommended strategy for building the Commercial Demonstration Fast Reactor (CDFR), subject to normal licensing procedure and to public inquiry, so as to ensure that the key options for introducing commercial fast reactors, when required, should remain open. A Government statement is expected during the next few months. Meanwhile the level of effort on fast reactor research and development in the UK has been maintained, the fast reactor remaining the largest of the UKAEA's reactor development projects with expenditure totalling somewhat over £80M per annum. The main feature of the UK fast reactor programme has continued to be the operation of PFR (Sections 2 and 7) which is yielding a wealth of experience and of information relevant to the design of commercial fast reactors. Bum-up of standard driver fuel has reached 6-7% by heavy atoms, while specially enriched lead fuel pins have reached 11 % without failure. An extensive programme of work in the reactor and its associated steam plant was completed in March 1980 and the reactor then started its fifth power run. The fuel reprocessing plant at DNE is being commissioned and has reprocessed some of the spent fuel remaining from the DFR. It will start soon on reprocessing fuel discharged from the PFR. During the year improvements to the design of the future

  18. Preliminary evaluation of alternate-fueled gas cooled fast reactors

    International Nuclear Information System (INIS)

    A preliminary evaluation of various alternative fuel cycles for the Gas-Cooled Fast Reactor (GCFR) is presented. Both homogeneous and heterogeneous oxide-fueled GCFRs are considered. The scenario considered is the energy center/dispersed reactor concept in which proliferation-resistant denatured reactors are coupled to 233U production reactors operating in secure energy centers. Individual reactor performance characteristics and symbiotic system parameters are summarized for several possible alternative fuel concepts. Comparisons are made between the classical homogeneous GCFR and the advanced heterogeneous concept on the basis of breeding ratio, doubling time, and net fissile gain. In addition, comparisons are made between a three-dimensional reactor model and the R-Z heterogeneous configuration utilized for the depletion and fuel management calculations. Lastly, thirty-year mass balance data are given for the various GCFR fuel cycles studied

  19. Knowledge management in fast reactors and related fuel cycles

    International Nuclear Information System (INIS)

    Full text: The 21st century is ushering in a new phase of economic and social development which can be referred as 'Knowledge Economy', in which knowledge has become the key asset in determining the organization's success or failure. The IAEA defines knowledge management as an integrated, systematic approach to identify, manage and share an organization's knowledge collectively in order to help achieve the objectives of the organization. Nuclear technology is very complex and a highly technical endeavor. It relies on innovative creation, storage and dissemination of knowledge. The nuclear energy is characterized by long time scales and technological excellence. Nuclear knowledge management is a critical input to nuclear power industry, the associated fuel cycle activities and nuclear applications in medicine, industry and agriculture. Realizing the importance of knowledge preservation in the area of fast reactor technology, IAEA had given a consultancy work to Argonne National Laboratory to study and suggest the means of knowledge management. The IAEA initiative seeks to establish a comprehensive inventory of fast reactor data and knowledge for the fast reactor development in the coming years. It was suggested that the knowledge regarding important disciplines like fuels and materials, reactor physics and core design, operations, the demonstration of safety should be preserved. Various countries have initiated the fast reactor knowledge preservation activities. In France, CEA, EDF and Framatome ANP have initiated liquid metal cooled fast reactor knowledge preservation project that deals with R and D aspects and Superphenix design. European Fast Reactor collaboration (MASURCA,SNEAK,ZEBRA) has preserved the zero power critical experimental data in the SNEDAX database. Japan has started a comprehensive knowledge preservation program including the capture of 'Human Knowledge' based on interviews. In Russia steps are initiated to preserve fast reactor knowledge

  20. Transmutation of /sup 90/Sr and /sup 137/Cs in a high-flux fast reactor with a thermalized central region

    Energy Technology Data Exchange (ETDEWEB)

    Taube, M.

    1976-10-01

    The fission products /sup 90/Sr and /sup 137/Cs produced by fission reactors of 30 GW(th) can be transmutated into stable nuclides by neutron irradiation with a thermal flux of 2 x 10/sup 16/ n cm/sup -2/ s/sup -1/. The rates of transmutation are 15 and 3.3 times greater, respectively, than that of spontaneous beta decay. The transmutation would take place in a central thermalized region of a high-flux fast burner reactor of 7 GW(th). In the case where the power reactors of 23 GW(th) are breeders with a high breeding gain of G = 0.38, the total system, inclusive of the high-flux burner, remains a breeding system, with G/sub total/ = 0.09. Details of the neutronics calculations and simplified thermohydraulics are given. The high-flux burner is fueled with a molten salt of chlorides of plutonium and sodium with a power density of 10 kW cm/sup -3/. The ''self-liquidation'' of such a system is discussed.

  1. Capital cost: gas cooled fast reactor plant

    International Nuclear Information System (INIS)

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design

  2. Small size modular fast reactors in large scale nuclear power

    International Nuclear Information System (INIS)

    The report presents an innovative nuclear power technology (NPT) based on usage of modular type fast reactors (FR) (SVBR-75/100) with heavy liquid metal coolant (HLMC) i. e. eutectic lead-bismuth alloy mastered for Russian nuclear submarines' (NS) reactors. Use of this NPT makes it possible to eliminate a conflict between safety and economic requirements peculiar to the traditional reactors. Physical features of FRs, an integral design of the reactor and its small power (100 MWe), as well as natural properties of lead-bismuth coolant assured realization of the inherent safety properties. This made it possible to eliminate a lot of safety systems necessary for the reactor installations (RI) of operating NPPs and to design the modular NPP which technical and economical parameters are competitive not only with those of the NPP based on light water reactors (LWR) but with those of the steam-gas electric power plant. Multipurpose usage of transportable reactor modules SVBR-75/100 of entirely factory manufacture assures their production in large quantities that reduces their fabrication costs. The proposed NPT provides economically expedient change over to the closed nuclear fuel cycle (NFC). When the uranium-plutonium fuel is used, the breeding ratio is over one. Use of proposed NPT makes it possible to considerably increase the investment attractiveness of nuclear power (NP) with fast neutron reactors even today at low costs of natural uranium. (authors)

  3. Advanced Multiphysics Modeling of Fast Reactor Fuel Behavior

    International Nuclear Information System (INIS)

    Evaluation of fast reactor fuel thermo-mechanical performance using fuel performance codes is a key aspect of advanced fast reactors designs. Those fuel performance codes capture the multiphysics nature of fuel behavior during irradiation where different, mostly interdependent, phenomena are taking place. Existing fuel performance codes do not fully capture those interdependencies and present the different phenomena through de-coupled models. Recent developments in multiphysics simulation capabilities and availability of advanced computing platforms led to advancements in simulation of nuclear fuel behavior. This paper presents current experiences in applying different multiphysics simulation platforms to evaluation of fast reactors metallic fuel behavior. Full 3D finite element simulation platforms that include capabilities to fully couple key fuel behavior models are discussed. Issues associated with coupling metallic fuels phenomena, such as fission gas models and constituent distribution models, with thermo-mechanical finite element platforms, as well as different coupling schemes are also discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, Stephen Rhead [Idaho National Lab. (INL), Idaho Falls, ID (United States); Medvedev, Pavel G. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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.

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

    International Nuclear Information System (INIS)

    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.

  6. Sodium components cleaning status in the Italian fast reactor program

    International Nuclear Information System (INIS)

    As a consequence of the Italian Fast Reactor Development, mainly aimed to the PEC project and to the participation in the French Superphenix project, it is of increasing importance to set up a reliable method for specific reactor components and related test loops. The first problem was the cleaning of the PEC fuelling machine. In order to perform the routine maintenance of the machine an alcohol cleaning method based on the use of 2-butoxyethanol-NN dimethylformamide mixture has been proposed

  7. Studies on the transient operation and stability of fast reactors

    International Nuclear Information System (INIS)

    These studies form part of the general programme of perfecting calculation methods for fast reactors. The basic formulae are given for the layouts used, i.e. the classic kinetic and thermal exchange equations, etc. A description is then given of the digital computer methods employed for studying the stable functioning of the reactor and of the methods used for transient operation studies. Finally, some examples of application are discussed and a comparison is made with parallel studies on the same subject. (author)

  8. UK fast reactor components - sodium removal decontamination and requalification

    International Nuclear Information System (INIS)

    Over the past two decades extensive experience on sodium removal techniques has been gained at the UKAEA's Dounreay Nuclear Establishment from both the Dounreay Fact Reactor (DFR) and the Prototype Fast Reactor (PFR). This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Part 2 of the paper, which describes recent operations associated with the PFR, demonstrates the background to these views. This past and continuing experience is being used in forming the basis of the plant to be provided for sodium removal, decontamination and requalification of components in the UK's future commercial fast reactors. Further improvements in techniques and in component designs can be expected in the course of the next few years. Consequently UK philosophy and approach with respect to maintenance and repair operations is sufficiently flexible to enable relevant improvements to be incorporated into the next scheduled fast reactor - the Commercial Demonstration Fast Reactor (CUR). This paper summarises the factors which are being taken into consideration in this continuously advancing field

  9. Spectrophotometric Procedure for Fast Reactor Advanced Coolant Manufacture Control

    Science.gov (United States)

    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.

  10. A new neutron noise technique for fast reactors

    International Nuclear Information System (INIS)

    This paper gives a new neutron noise technique for fast reactors, which is known as thermalization measurement technique of the neutron noise. The theoretical formulas of the technique were developed, and a digital delayed coincidence time analyzer consisted of TTL integrated circuits was constructed for the study of this technique. The technique has been tested and applied practically at Df-VI fast zero power reactor. It was shown that the provided technique in this work has a number of significant advantages in comparison with the conventional neutron noise method

  11. Fast ultrasonic visualisation under sodium. Application to the fast neutron reactors

    International Nuclear Information System (INIS)

    The fast ultrasonic visualization under sodium is in the programme of research and development on the inspection inside the fast neutron reactors. This work is about the development of a such system of fast ultrasonic imaging under sodium, in order to improve the existing visualization systems. This system is based on the principle of orthogonal imaging, it uses two linear antennas with an important dephasing having 128 piezo-composite elements of central frequency equal to 1.6 MHz. (N.C.)

  12. The Integral Fast Reactor: A practical approach to waste management

    International Nuclear Information System (INIS)

    This report discusses development of the method for pyroprocessing of spent fuel from the Integral Fast Reactor (or Advanced Liquid Metal Reactor). The technology demonstration phase, in which recycle will be demonstrated with irradiated fuel from the EBR-II reactor has been reached. Methods for recovering actinides from spent LWR fuel are at an earlier stage of development but appear to be technically feasible at this time, and a large-scale demonstration of this process has begun. The utilization of fully compatible processes for recycling valuable spent fuel materials promises to provide substantial economic incentives for future applications of the pyroprocessing technology

  13. Elements for evaluation of fast breeder reactor's potential in Argentina

    International Nuclear Information System (INIS)

    Fast Breeder Reactors (FBR) main features are presented in a general form, including their physical principles, the history of their evolution, their relevant technological aspects and the basis for their comparison to other energy sources. This is completed with descriptions of typical reactors and a model of FBR penetration in the Argentine electrical network. It is recommended to form a multidisciplinary board to study which position should be taken with respect to this type of reactors. In the author's opinion a Research activity should be started and gradually increased for passing to Development activities after a short while. (Author)

  14. Fast-Mixed Spectrum Reactor. Progress report for 1979

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, G.J.; Cerbone, R.J.

    1980-05-01

    This report summarizes the progress of the Fast Mixed Spectrum Reactor (FMSR) since the publication of the Interim Report in January 1979. The FMSR program was initiated to determine the feasibility of a breeder reactor concept which operated on a once-through-and-store fuel cycle and for which the only feed would be natural uranium. A first or startup core enriched to a maximum of about eleven percent in uranium-235 would be required. The concept has excellent antiproliferation advantages. In the once-through and store mode, the FMSR has a resource utilization which is a factor of four higher than a light water reactor.

  15. Fast-Mixed Spectrum Reactor. Progress report for 1979

    International Nuclear Information System (INIS)

    This report summarizes the progress of the Fast Mixed Spectrum Reactor (FMSR) since the publication of the Interim Report in January 1979. The FMSR program was initiated to determine the feasibility of a breeder reactor concept which operated on a once-through-and-store fuel cycle and for which the only feed would be natural uranium. A first or startup core enriched to a maximum of about eleven percent in uranium-235 would be required. The concept has excellent antiproliferation advantages. In the once-through and store mode, the FMSR has a resource utilization which is a factor of four higher than a light water reactor

  16. Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997

  17. Thermal and neutronic calculation for fast breeder reactor FBR

    International Nuclear Information System (INIS)

    This research included studying of thermal and neutronic calculation for fast breeder nuclear reactor, to putting the optimum design for this reactor. So a Soviet type (BN-350) was chosen, which has its core composed of two enrichment zones, and with blanket that contains depleted uranium. A group of thermal calculation programs was made by using personal computer, to obtain core and blanket reactor dimensions and volume fractions of reaction input material and number and dimensions of fuel rods which were used for neutron calculations. Several core and blanket enrichments were used to study neutron flux behaviour for two reactors different conditions. First when control rods exist in the core reactor and second when the rods are out of the core. Breeding ratio was also studied for different core and blanket enrichment. 30 tabs.; 24 figs.; 34 refs.; 3 apps

  18. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. New concept of proliferation resistant sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Eliseev, V.A.; Krivitski, I.Y.; Matveev, V.I.; Popov, E.P.; Savitski, V.I.; Tsikunov, A.G. [Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2001-07-01

    The full text follows. It is proposed the concept of BN-800 sodium cooled fast reactor operating in the closed fuel cycle with special reprocessing technology. The use of nitride fuel allows improving the parameters of reactor safety (internal breeding {approx}1, zero value of sodium void reactivity effect), economy (one refueling per year), ecology (use of nitride enriched by nitrogen-15) and non-proliferation (use of reprocessing without separating the plutonium from uranium). The main difficulty of this type reactor development is that the technical project of BN-800 reactor with MOX fuel was developed. When using the nitride fuel it is necessary to serve (in max extent) the mail technical decisions of this project. This report presents first results on development and justification of the BN-800 reactor with nitride fuel core. (authors)

  20. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    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. Status of national programmes on fast reactors 1997/98. 31. annual meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    The objective of the meeting was to co-ordinate the exchange of information on the status of fast reactor development and operational experience, including experience with experimental types of reactor; to consider meeting arrangements for 1998 and 1999; and to review the IAEA co-ordinated research activities in the field of fast reactor, as well as co-ordination of the International Working Group on Fast Reactors activities with other organizations

  2. The renaissance of fast sodium reactors 2007 assessment: situation and contributions from the Phenix experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J. [Phenix Plant (France)

    2007-07-01

    The first nuclear reactor to produce electrical current was the fast sodium/potassium reactor EBR-1 in Idaho (Usa). Following this pioneering experience, France, Germany, Great Britain, Usa, Japan, Russia and India launched construction of fast sodium reactors. In the post Chernobyl years, waves of protest against nuclear power grew and swelled, leading to a strong overall slowdown for this reactor type. The SNR-300 project in Germany never started up, and was shut down. In Great Britain, PFR was definitely shut down, operation of MONJU in Japan and BN-800 project in Russia were frozen, FFTF in the United States shut down, and finally the SPX-1 project in France was also stopped. When PHENIX started back up in 2003, there were only three other research reactors operating worldwide: FBTR in India, BOR-60 in Russia and JOYO in Japan, and one power reactor BN-600 in Russia. The Generation-IV initiative was the opportunity for global thinking about reactors for the future, referred to as fourth generation reactors. Six reactor designs were selected, including the fast sodium reactor. However, after several years, most of the countries have officially announced or confirmed that the fast sodium reactor is their priority reference design. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision to build a prototype scheduled for operation in 2020. These and other plans are all sustained in a very practical manner by the ongoing production in the field. PHENIX has been operating since 2003, contributing to the development of future systems and demonstrating the fast reactors ability to burn waste. Following the excellent results obtained by the BN-600, Russia has re-launched the BN-800 project. China is currently in the process of building a 65 MW research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU for divergence in 2008. In India, a 1200 MW (thermal) power reactor is under

  3. Fast breeder reactor-block antiseismic design and verification

    International Nuclear Information System (INIS)

    The Specialists' Meeting on ''Fast Breeder Reactor-Block Antiseismic Design and Verification'' was organized by the ENEA Fast Reactor Department in co-operation with the International Working Group (IWGFR) of the International Atomic Energy Agency (IAEA), according to the recommendations of the 19th IAEA/IWGFR Meeting. It was held in Bologna, at the Headquarters of the ENEA Fast Reactor Department, on October 12-15, 1987, in the framework of the Celebrations for the Ninth Centenary of the Bologna University. The proceedings of the meeting consists of three parts. Part 1 contains the introduction and general comments, the agenda of the meeting, session summaries, conclusions and recommendations and the list of participants. Part 2 contains 8 status reports of Member States participating in the Working Group. Contributed papers were published in Part 3 and were further subdivided into 5 sessions as follows: whole reactor-block analysis (4 papers); whole reactor-block analysis (sloshing and buckling, seismic isolation effects) (8 papers); detailed core analysis (6 papers); shutdown systems and core structural and functional verifications (6 papers); component and piping analysis (7 papers). A separate abstract was prepared for each of the 8 status reports and 31 contributed papers. Refs, figs and tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  6. Basic cable routing guidelines for a fast reactor plant

    International Nuclear Information System (INIS)

    In this paper the guidelines evolved for cable routing in 500 MWe Prototype Fast Breeder Reactor (PFBR) are presented. Safety related redundant system cables in a nuclear plant shall not become unavailable due to cable fire. This is ensured by proper cable routing in the plant in addition to the other general fire protection measures

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

    International Nuclear Information System (INIS)

    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 ampersand D

  8. Status of the fast breeder reactor technology in China

    International Nuclear Information System (INIS)

    According to the Chinese long-term energy strategy the FBR development is strongly supported. In the near term nuclear programme it is intended to build the experimental First Fast Reactor (FFR) in the year 2000. Design work is in progress. (author). 1 ref., 6 figs, 8 tabs

  9. The Programme for Fast Reactor Development in the Russian Federation

    International Nuclear Information System (INIS)

    The paper highlights the status and perspectives on the development of nuclear energy based on fast reactor and closed fuel cycle technologies in the Russian Federation. Information is presented on the new Federal Target Programme 'Nuclear Power Technologies of a New Generation for the Period 2010-2015 and the Outlook to 2020'. (author)

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

    International Nuclear Information System (INIS)

    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 ampersand D

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Symposium on key questions about the fast breeder reactor

    International Nuclear Information System (INIS)

    Except for several introductions on various aspects of the fast breeder reactor development this paper contains the full texts of the discussions held in the sub-groups panels on resp. technical matters, environment and health, society, politics and economics. The main issues of each discussion are summarized

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

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

    International Nuclear Information System (INIS)

    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

  17. Fast Reactor Knowledge Organization System: Implementation and challenges

    International Nuclear Information System (INIS)

    For three decades, several countries had large and vigorous fast breeder reactor development programmes, which had their peaks by 1980. From that time onward, Fast Reactor (FR) development generally began to decline and efforts for FR reactor development essentially disappeared by 1994. This development stagnation continued until 2003. In September 2003, in Resolution GC(47)/RES/10.B, the International Atomic Energy Agency (IAEA) General Conference recognised the vitality of nuclear knowledge. The loss of FR knowledge has been taken seriously and the IAEA took the initiative to coordinate the efforts of the member states in the preservation of knowledge in FRs. In the framework of this initiative, the IAEA intends to create an international inventory combining information from different member states on FRs and organized in the knowledge system in a systematic and structured manner

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

    International Nuclear Information System (INIS)

    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

  19. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    Directory of Open Access Journals (Sweden)

    Vladimir Petrochenko

    2012-09-01

    Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

  20. High temperature fast reactor for hydrogen production in Brazil

    International Nuclear Information System (INIS)

    The main nuclear reactors technology for the Generation IV, on development phase for utilization after 2030, is the fast reactor type with high temperature output to improve the efficiency of the thermo-electric conversion process and to enable applications of the generated heat in industrial process. Currently, water electrolysis and thermo chemical cycles using very high temperature are studied for large scale and long-term hydrogen production, in the future. With the possible oil scarcity and price rise, and the global warming, this application can play an important role in the changes of the world energy matrix. In this context, it is proposed a fast reactor with very high output temperature, ∼ 1000 deg C. This reactor will have a closed fuel cycle; it will be cooled by lead and loaded with nitride fuel. This reactor may be used for hydrogen, heat and electricity production in Brazil. It is discussed a development strategy of the necessary technologies and some important problems are commented. The proposed concept presents characteristics that meet the requirements of the Generation IV reactor class. (author)

  1. Investigation of small and modular-sized fast reactor

    International Nuclear Information System (INIS)

    In this paper, feasibility of the multipurpose small fast reactor, which could be used for requirements concerned with various utilization of electricity and energy and flexibility of power supply site, is discussed on the basis of examination of literatures of various small reactors. And also, a possibility of economic improvement by learning effect of fabrication cost is discussed for the modular-sized reactor which is expected to be a base load power supply system with lower initial investment. (1) Multipurpose small reactor (a) The small reactor with 10MWe-150MWe has a potential as a power source for large co-generation, a large island, a middle city, desalination and marine use. (b) Highly passive mechanism, long fuel exchange interval, and minimized maintenance activities are required for the multipurpose small reactor design. The reactor has a high potential for the long fuel exchange interval, since it is relatively easy for FR to obtain a long life core. (c) Current designs of small FRs in Japan and USA (NERI Project) are reviewed to obtain design requirements for the multipurpose small reactor. (2) Modular-sized reactor (a) In order that modular-sized reactor could be competitive to 3200MWe twin plant (two large monolithic reactor) with 200kyenWe, the target capital cost of FOAK is estimated to be 260kyen/yenWe for 800MWe modular, 280kyen/yenWe for 400MWe modular and 290kyen/yenWe for 200MWe by taking account of the leaning effect. (b) As the result of the review on the current designs of modular-sized FRs in Japan and USA (S-PRISM) from the viewpoint of economic improvement, since it only be necessary to make further effort for the target capital cost of FOAK, since the modular-sized FRs requires a large amount of material for shielding, vessels and heat exchangers essentially. (author)

  2. The generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    The gas cooled fast reactor (GFR) is a helium-cooled fast spectrum reactor operating within a closed fuel cycle. It combines the advantages of fast reactors, in terms of a more sustainable use of uranium resources and waste minimisation, with the wider applicability of high temperature gas reactors, in terms of high efficiency electricity generation and the co-generation of high-quality process heat. Other advantages like the absence of threshold effect due to phase changing, the optical transparency and chemical inertness of the Helium coolant are also acknowledged. Within the European Union, GFR is one of the three fast reactors proposed for development to the demonstration stage within the European Sustainable Nuclear Industry Initiative (ESNII). On a wider global scale, GFR is one of the six systems proposed for further development within the Generation IV International Forum (GIF). In this respect, France, Switzerland, Japan and the European Union (through EURATOM) are signatories to the 'System Arrangement', the instrument through which the international research efforts are coordinated. This paper presents the current status of the development of the GFR system. The status of the GFR programme in each of the signatory countries is summarised including the intended contribution of the newly launched EURATOM 7. Framework Programme project - GoFastR. France has provided the bulk of the effort on conceptual design, safety assessment and fuel development. Switzerland makes significant contributions to the GFR system in the areas of core physics, uncertainty analysis, deterministic safety assessment and fuel development. Historically Japan has been very active in the development of the GFR system. Within the Generation IV GFR system, Japan contributes to the development of fuel and core materials

  3. Fast Neutron Detector for Fusion Reactor KSTAR Using Stilbene Scintillator

    CERN Document Server

    Lee, Seung Kyu; Kim, Gi-Dong; Kim, Yong-Kyun

    2011-01-01

    Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spectra with high radiation from a gamma-ray background. The signals from neutrons and gamma-rays are discriminated by the digital charge pulse shape discrimination (PSD) method, which uses total to partial charge ratio analysis. The signals are digitized by a flash analog-to-digital convertor (FADC). To evaluate the performance of the fabricated stilbene neutron diagnostic system, the efficiency of 10 mm soft-iron magnetic shielding and the detection efficiency of fast neutrons were tested experimentally using a 252Cf neutr...

  4. Reactivity changes in hybrid thermal-fast reactor systems during fast core flooding

    International Nuclear Information System (INIS)

    A new space-dependent kinetic model in adiabatic approximation with local feedback reactivity parameters for reactivity determination in the coupled systems is proposed in this thesis. It is applied in the accident calculation of the 'HERBE' fast-thermal reactor system and compared to usual point kinetics model with core-averaged parameters. Advantages of the new model - more realistic picture of the reactor kinetics and dynamics during local large reactivity perturbation, under the same heat transfer conditions, are underlined. Calculated reactivity parameters of the new model are verified in the experiments performed at the 'HERBE' coupled core. The model has shown that the 'HERBE' safety system can shutdown reactor safely and fast even in the case of highly set power trip and even under conditions of big partial failure of the reactor safety system (author)

  5. Status of fast reactor development in India

    International Nuclear Information System (INIS)

    The economic liberalization process has accelerated the industrial growth and this requires considerable energy input. Nuclear energy has to play an important role in supporting the increasing demand of energy. In this year FBTR was operated at 10.2 MWt power in a sustained manner. Several physics and engineering experiments were carried out. The mixed carbide fuel has achieved a peak burnup of 16,000 MW d/t without failure. First batch of irradiation experiments was completed and pins were delivered to RML for PIE. For PFBR, a thorough review of the conceptual design was carried out towards reducing the capital cost, construction time and for improving plant reliability. A 2 loop concept with 2 PSPs, 4 IHXs and 2 secondary loops having 4 integrated SG modules has been finally chosen with an expected savings of 15% in NSSS capital cost, 2-3 years in construction time and 10-12% in capacity factor. Number of materials to be used for major components was reduced to three to facilitate speedy development. Operating temperatures were finalized after optimization studies. Discussion is in progress to finalize fuel handling system. Research and development activities are continuing at ICGAR in the areas of reactor physics, engineering development, core engineering, thermal hydraulics, structural mechanics, metallurgy, post irradiation examination, instrumentation and electronics, chemistry, fuel reprocessing, safety research and health physics etc. (author). 9 figs

  6. BN800: The advanced sodium cooled fast reactor plant based on close fuel cycle

    International Nuclear Information System (INIS)

    As one of the advanced countries with actually fastest reactor technology, Russia has always taken a leading role in the forefront of the development of fast reactor technology. After successful operation of BN600 fast reactor nuclear power station with a capacity of six hundred thousand kilowatts of electric power for nearly 30 years, and after a few decades of several design optimization improved and completed on its basis, it is finally decided to build Unit 4 of Beloyarsk nuclear power station (BN800 fast reactor power station). The BN800 fast reactor nuclear power station is considered to be the project of the world's most advanced fast reactor nuclear power being put into implementation. The fast reactor technology in China has been developed for decades. With the Chinese pilot fast reactor to be put into operation soon, the Chinese model fast reactor power station has been put on the agenda. Meanwhile, the closed fuel cycle development strategy with fast reactor as key aspect has given rise to the concern of experts and decision-making level in relevant areas. Based on the experiences accumulated in many years in dealing the Sino-Russian cooperation in fast reactor technology, with reference to the latest Russian published and authoritative literatures regarding BN800 fast reactor nuclear power station, the author compiled this article into a comprehensive introduction for reference by leaders and experts dealing in the related fields of nuclear fuel cycle strategy and fast reactor technology development researches, etc. (authors)

  7. Innovations in Equipment Erection of Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is sodium cooled, pool type reactor with generating capacity of 1250 MWt/500 MWe. Reactor assembly consists of large dimensional vessels like Safety vessel (13.54 m diameter, 12.8 m height and weight approximately 155 MT) and Main vessel (12.9 m diameter, 12.94 m height and weight approximately 202 MT including core catcher, core support structure and cooling pipes) and Steam generator (26 m length, 1.5 m diameter, and weight approximately 35 MT). PFBR reactor equipment erection was a challenging task where thin walled vessels had transported and handled with utmost precaution to avoid radial forces on the vessels which could buckle the vessels. There was a real challenge in lifting the vessels without swing, placement of large size and heavy vessel at a distance of 57 m where the crane operator had no line of site to the equipment being erected. To handle such over dimensional reactor components many mock-up tests had been carried out before erection and gained lot of confidence. Lot of care had been taken during lifting, handling and erection of thin walled over dimensional reactor components with innovative methods used for lifting fixtures, guiding arrangements, alignment fixtures and achieved the stringent erection tolerances. This paper discusses the first ever experiences gained during the handling and erection of such thin walled, over dimensional reactor components at PFBR site. (author)

  8. Reactor shutdown system of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6/ ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

  9. Fast reactor development strategy targets study in China

    International Nuclear Information System (INIS)

    China is a big developing Country who needs a huge energy resources and a rapid growing rate. Considering energy resources limited and environment issues it is sure that the nuclear energy will be becoming one of the main energy resources. The Government has decided to develop the nuclear power capacity to 40 GW in 2020. It is envisaged that it will reach to 240 GW in 2050. It is stimulate us to consider conscientiously the development of the fast breeder reactor's and related closed nuclear fuel cycle by the limitation of Uranium resources and uncertainties of international Uranium market. Followings are the proposed strategic targets of fast reactor development in China. (1) To realize the operation of commercial fast breeder reactors with an unit size of 800-900 MWe and one site-multi reactors in 2030. (2) To develop the nuclear power capacity to 240 GW in 2050. (3) To replace step by step the fossil fuel utilization in large scale by nuclear energy beyond 2050. (authors)

  10. Physics aspects of metal fuelled fast reactors with thorium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Mohapatra, D.K., E-mail: dina@igcar.gov.in; Singh, S.S.; Riyas, A.; Mohanakrishnan, P.

    2013-12-15

    Metal fuelled fast breeder reactors (MFBR) with high breeding ratio will play a major role in meeting the high nuclear power growth envisaged in India. In this regard several conceptual reactor designs with alloys of U–Pu–Zr fuel have been suggested for commercial operations. This study focusses on the physics design aspects of a sodium cooled U–Pu–6%Zr fuelled 1000 MWe fast breeder reactor, which can attain a breeding ratio of nearly 1.5. The calculation results on reactor kinetics and safety parameters of the 1000 MWe MFBR are presented. The changes in the breeding ratio by introduction of thorium in the blankets of the MFBR are also investigated. Burnup analyses are carried out to compare the core burnup effects in MOX and metal fuelled FBRs. Since the MOX fuelled 500 MWe prototype fast breeder is getting constructed at IGCAR, for burnup comparisons a MFBR of similar design is considered. The results of this study indicate that the loss of reactivity in the metal core with burnup is less than half that of a MOX core and its breeding ratio remains nearly constant. It is also found that the isotopic composition of plutonium (Pu-vector composition) remains more steady with burnup in a metal core.

  11. Design of fuel fabrication plant of Fast Reactor Fuel Cycle Facility for reload requirement of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    India's economic growth is on a fast growth track. The energy demand is expected to grow rapidly in the coming decades. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear energy is best suited to meet this demand in a sustainable manner without causing undue environmental impact. Fast reactors are expected to be major contributors in sufficing this demand to a great extent. As an effort to achieve the objective, a Prototype Fast Breeder Reactor is being constructed at Kalpakkam. This paper also highlights the design features of FFP, unit operations, scheme of automation, branched layout of glove box train, shielding arrangement on glove boxes, accident consequence analysis etc.

  12. Performance and characteristics of a small-sized oxide-fuelled fast reactor (PRISM) for the burning of excess plutonium

    International Nuclear Information System (INIS)

    The increasing availability of fissile material and the worldwide perception of the need to reduce or control such material make a new approach towards nuclear reactors necessary. The Nuclear Fission Division at ENEA is presently interested in evaluating various options for the burning of excess plutonium, among which the LWR reactors using rock-like inert matrix fuel in a once-through cycle and fast reactors with recycling. This paper deals with the latter option. A fast system, rich in neutrons, can produce plutonium by breeding, or it can burn it (together with Minor Actinides) with a high degree of effectiveness: this flexibility makes the fast system worth considering. In the framework of a co-operation project between GE-USA and ENEA, the PRISM MOD D oxide-fuelled reactor (small size, 840 MW) was studied as a burner. A complete set of studies was carried out covering several fields: neutronics, thermohydraulics, dynamics and safety. A negative void coefficient was achieved thanks to a particular arrangement of the fuel subassemblies (leaky core) and to the enrichment distribution. The central zone, loaded with shielding subassemblies, accounts for a negative void coefficient even in the inner zone. Moreover, this zone could become a suitable place for burning long-lived fission products. As a burner, using a standard fuel, the expected performance is a consumption of about 60-70 kg Pu/TWhe. ULOF and TOP analysis show a begin transient evolution. Only half of the GEMs are sufficient to close the ULOF transient; the reactor power reaches its maximum at 1.35 of the nominal figure during a TOP accident. A method of detecting a failure in a subassembly and identifying the subassembly concerned was developed. It was demonstrated that it is possible to create a sufficient number of tags for them to be recognised as different by the detector system, by using 'cocktails' of only four tag gases. The total gas required to overcome the uncertainties is less than 15 ncm3

  13. Physical and technical aspects of lead cooled fast reactors safety

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    2001-07-01

    The safety analysis of lead-cooled fast reactors has been performed for the well-developed concept of BREST-OD-300 reactor. The most severe accidents have been considered. An ultimate design-basis accident has been defined as an event resulting from an external impact and involving a loss of leak-tightness of the lead circuit, loss of forced circulation of lead and loss of heat sink to the secondary circuit, failure of controls and of reactor scram with resultant insertion of total reactivity margin, etc. It was assumed in accident analysis that the protective feature available for accident mitigation was only reactivity feedback on the changes in the temperatures of the reactor core elements and coolant flow rate, and in some cases also actuation of passive protections of threshold action in response to low flow rate and high coolant temperature at the core outlet. It should be noted that the majority of the analyzed accidents could be overcame even without initiation of the above protections. It has been demonstrated that a combination of inherent properties of lead coolant, nitride fuel, physical and design features of fast reactors will ensure natural safety of BREST and are instrumental for avoiding by a deterministic approach the accidents associated with a significant release of radioactivity and requiring evacuation of people in any credible initiating event and a combination of events. (author)

  14. Characteristics of fast reactor core designs and closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N. [State Scientific Center of the Russian Federation, Institute for Physics and Power Engineering (IPPE), Obninsk, Kaluga Region (Russian Federation)

    2007-07-01

    On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

  15. Fast reactor programme. Annual progress report 1982

    International Nuclear Information System (INIS)

    The status of recent fast-capture cross sections for important fission-product nuclides has been reviewed; an intercomparison of evaluations for Eu-isotopes has been made and corrections have been applied to recent reported evaluations of neutron capture cross sections for Pd isotopes. An outline of the evaluation procedure for the nuclides sup(58g)Co and sup(58m)Co is given. The evaluation of the cover-gas nuclides has been completed with additional results for 36Ar and 38Ar. Some results of the latest fuel failure experiments under simulated reduced coolant conditions, the so-called SHOT experiments, are given. The first irradiation experiments with the prototype irradiation facility HFR-TOP 01 are described. Neutron flux calculations have been performed to determine the dimensions of a flux depression plate to achieve a symmetric flux distribution inside the fuel during pre-irradiation. The creep investigations on various heats and welded joints of DIN 1.4948 have been finished; the main findings are reported. A first project on the low-cycle fatigue behaviour of DIN 1.4948 has been completed. A three-dimensional finite element analysis has been performed on compact tension test specimen having a curved crack-front due to crack-tunneling. A code version of the VITESSE computer code has been developed to predict the thermohydraulic behaviour of distorted bundle geometries. Results from the LDA measuring programmes in the different test sections with respect to the secondary flow velocities are reported. Noise measurements in an unblocked 60 deg. reference bundle have been performed. (Auth.)

  16. Sodium fast reactor power monitoring using gamma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.; Normand, S.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A.M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, CEA - Saclay DRT/LIST/DETECS/SSTM, Batiment 516 - P.C. no 72, Gif sur Yvette, F-91191 (France); Montagu, T.; Dautremer, T.; Barat, E. [CEA, LIST, Laboratoire Processus Stochastiques et Spectres (France); Ban, G. [ENSICAEN (France)

    2009-06-15

    This work deals with the use of high flux gamma spectrometry to monitor the fourth generation of sodium fast reactor (SFR) power. The simulation study part of this work has shown that power monitoring in a short time response and with a good accuracy is possible. An experimental test is under preparation at the French SFR Phenix experimental reactor to validate simulation studies. First, physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as the sodium velocity, the atomic densities, Phenix neutron spectrum and incident neutron cross-sections of reactions producing gamma emitters. A thermal hydraulic transfer function was used for modeling primary sodium flow in our calculations. For the power monitoring problematic, use of a short decay period gamma emitter will allow to have a very fast response system without cumulative effect. We have determined that the best tagging agent is 20F which emits 1634 keV energy photons with a decay period of 11 s. The gamma spectrum was determined by flux point and a pulse high tally MCNP5.1.40 simulation and shown the possibility to measure the signal of this radionuclide. The experiment will be set during the reactor 'end life testing'. The Delayed Neutron Detection (DND) room has been chosen as the best available location on Phenix reactor to measure this kind of radionuclide due to a short transit time from reactor core to measurement sample. This location is optimum for global power measurement because homogenized sampling in the reactor hot pool. The main spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The HPGe diode signal will be processed by the Adonis digital signal processing due to high flux and fast activity measurement. Post-processing softwares will be used to limit statistical problems of the

  17. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    G Pandikumar; V Gopalakrishnan; P Mohanakrishnan

    2009-05-01

    In a thermal neutron reactor, multiple recycle of U–Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder reactor under construction at Kalpakkam. After about five recycles, the cycle-to-cycle variation in the above parameters is below 1%.

  18. Criteria for structural verification of fast reactor core elements

    International Nuclear Information System (INIS)

    Structural and functional criteria and relative verifications of PEC reactor fuel element are presented and discussed. Particular attention has been given to differentiate the structural verifications of low neutronic damage zones from those high neutronic damage ones. The structural verification criteria, which had already been presented at the 8th SMIRT Seminar Conference in Paris, have had some modifications during the Safety Report preparation. Finally some necessary activities are indicated for structural criteria validation, in particular for irradiated components, and for converging towards a European fast reactor code. (author). 3 refs, 6 tabs

  19. Technical meeting to 'Preserve fast reactor physics knowledge'. Working material

    International Nuclear Information System (INIS)

    The meeting extended its scope beyond reactor physics to include all the main areas of fast reactor data retrieval and knowledge preservation (FR KP). The participants presented the status of the national FR KP efforts and the progress achieved since the kick-off meeting of the IAEA initiative (meeting hosted by ANL-West in Idaho Falls, Idaho, 2-4 April 2002). Details are given in Section 2. The Scientific Secretary of the Technical Working Group on Fast Reactors (TWG-FR) presented the Agency activities (KNK II data and documentation retrieval and preservation), and recalled the Agency's role in this initiative: - Coordination of the national efforts - Ensuring the collaboration with other International Organizations (mainly OECD/NEA) - Establishing and maintaining the access means to the ultimate goal of the initiative, the 'fast reactor knowledge base'. The integration of specific activities relevant to the FR KP initiative, which are planned within the framework of the TWG-FR, was discussed. It was agreed to implement the following as TWG-FR tasks with clear relevance to FR KP initiative: - Japanese 'Proposal from Monju relevant to Fast Reactor Knowledge Preservation Activity in the framework of the IAEA TWG-FR' - Proposal of a CRP on 'Generalization and Analyses of Operational Experience with Fast Reactor Equipment and Systems' - TM on 'Handling of Sodium Coming from Decommissioned Fast Reactors and from the Shutdown of Experimental Facilities' (if not already covered by the TECDOC being prepared by IAEA's Nuclear Waste Technology Section). While the responsibility for fast reactor knowledge preservation, data retrieval and interpretation, as well as quality assurance will rest with the individual Member States joining the FR KP initiative, the participants confirmed the Agency's role (see above). More specifically, the participants in the meeting recommended that the IAEA - support and coordinate data retrieval and interpretation efforts by the fast reactor

  20. Technical committee meeting on Liquid Metal Fast Reactor (LMFR) developments. 33rd annual meeting of the International Working Group on Fast Reactors (IWG-FR). Working material

    International Nuclear Information System (INIS)

    Over the past 33 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1999/2000, as reported at the 33. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States

  1. Fast breeder reactors: Experience and trends. V. 2

    International Nuclear Information System (INIS)

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

  2. Modelling Homogeneous Nucleation in Sodium Fast Reactors under BDBA Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, M.; Herranz, L. E.; Kissane, M.

    2014-07-01

    During postulated Beyond Design Basis Accidents (BDBAs) in Sodium-cooled Fast Reactors (SFRs), the contaminated coolant discharge at high temperature into the containment is considered as a potential scenario during the severe accident progression. In this scenario, the vaporization of sodium and its subsequent combustion (oxidation) would result in supersaturated sodium oxide vapours and formation of large quantities of contaminated aerosols by nucleation of these combustion products. (Author)

  3. Fast neutron reactor fuel elements and power grid duty cycling

    International Nuclear Information System (INIS)

    The PHENIX power grid cycling operation in 1982-1983 will allow verification of the models and criteria developed in the interim. It will provide indispensible statistical data and will open the way to power grid duty for Super PHENIX beginning in 1986. Although at the present time it is impossible to resolve the question of weekly or daily load variations, it is felt that fast neutron reactor fuel subassemblies should provide satisfactory performance for primary and secondary frequency adjustments

  4. Evaluation of the breed/burn fast reactor concept

    International Nuclear Information System (INIS)

    A core design concept and fuel management strategy, designated breed/burn, has been evaluated for heterogeneous fast breeder reactors. In this concept internal blanket assemblies after fissile material is bred in over several incore cycles, are shuffled into a moderated radial blanket and/or central island. The most promising materials combination identified used thorium in the internal blankets (due to the superior performance of epithermal Th-U233 systems) and zirconium hydride (ZrH16) as the moderator

  5. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The present document contains information on the status of fast breeder reactor development and on worldwide activities in this advanced nuclear power technology during 1989 as reported at the 23rd meeting of the IWGFR in Vienna, April 1990. The publication is intended to provide information regarding the current status of LMFBR development in IAEA Member States. A separate abstract was prepared for each of the 11 papers presented by the participants of this meeting. Refs, figs and tabs

  6. A methodology of neutronic-thermodynamics simulation for fast reactor

    International Nuclear Information System (INIS)

    Aiming at a general optimization of the project, controlled fuel depletion and management, this paper develop a neutronic thermodynamics simulator, SIRZ, which besides being sufficiently precise, is also economic. That results in a 75% reduction in CPU time, for a startup calculation, when compared with the same calculation at the CITATION code. The simulation system by perturbation calculations, applied to fast reactors, which produce errors smaller than 1% in all components of the reference state given by the CITATION code was tested. (author)

  7. Knowledge management: knowledge and competence maintaining; problematics; Fast reactor example

    International Nuclear Information System (INIS)

    In this expose are examined two representative aspects, with a first part reserved to the general problematics of the knowledge and competence maintaining as it looks at Framatome, in its activities of pressurized water boiler supplier (PWR) and as provider of nuclear services, and a second part treating the solutions used by the different actors intervening in fast neutrons reactors, that is to say, the Cea, Framatome and EDF. (N.C.)

  8. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The present document contains information on the status of fast breeder reactor development and on worldwide activities in this advanced nuclear power technology during 1990 as reported at the 24th meeting of the IWGFR in Tsuruga, Japan, 15-18 April 1991. The publication is intended to provide information regarding the current status of LMFBR development in IAEA Member States and CEC. Figs and tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Computational intelligent systems for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Nearly 15000 process signals are digitized by physically and functionally distributed embedded systems in Prototype Fast Breeder Reactor (PFBR). Digitized signals are processed and relevant information is displayed through Large video display systems at Control Room. It is necessary that correct and reliable information need to be provided to the plant operator. Computational intelligent systems play a major role in enhancing the safe operation of the Nuclear reactor. The paper explains the features of three such systems, one for on-line validation of neutronic power channel through on-line thermal balance calculation and another for detection of anomalous reactivity addition through on-line reactivity balance computation and third for on-line computation of Reactor power from fluctuations of core thermocouple signals. (author)

  11. A review of the Italian fast reactor programme, March 1979

    International Nuclear Information System (INIS)

    Year 1978 in Italy was marked by a standstill in the nuclear energy field. The decisions previously made for the installation of eight 1000 MWe LWR-type reactors could not be acted upon because of the opposition of local authorities and lack of Government power. The construction site at Montalto di Castro (two BWR reactors) was ope ned with difficulty, whereas the decision to install a plant in Mo use equipped with two PWR reactor was postponed. The new presidents of ENEL and CNEN were appointed in January this year and the appointments of the new Boards of Directors are underway. With regard to CNEN, many political bodies are in agreement on an institutional change which would widen field of activity to include new energy sources, solar energy in particular. This will open a big problem: if CNEN will be no more a 'nuclear body, it could be necessary to transfer all the activities connected to the Regulatory Commission to another separate body to be instituted. In this context, the fast reactor programme has continued to develop under the directives of CIPE, and has concentrated its effort on the following three objectives: the PEC-Reactor, the Creys-Malville Power Plant and research and development, and industrial promotion. These objectives are being pursued with the participation of CNEN, ENEL and Italian industry. CNEN has the role of committing and operating the PEC reactor; it is also charged to perform part of the R and D Italian-French programme and to promote industrial development. ENEL participates in the NERSA Company, owner of the Creys-Malville Plant. Italian industry, with its activities of architect-engineering, designing and manufacturing will participate in the construction of the PEC and of the Italian part (33%) of the Creys-Malville Plant. During the last months of 1978 a consortium (COREY) was set up by CNEN and NIRA which has the purpose of integrating and ensuring the smooth running of Italian efforts in the field of long-term research and

  12. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  13. Delayed gamma power measurement for sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Graphical abstract: Display Omitted Research highlights: →20F and 23Ne tagging agents are produced by fast neutron flux. →20F signal has been measured at the SFR Phenix prototype. → A random error of only 3% for an integration time of 2 s could be achieved. →20F and 23Ne power measurement has a reduced temperature influence. → Burn-up impact could be limited by simultaneous 20F and 23Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,α) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  14. Romanian Contribution to the Development of Lead Cooled Fast Reactors

    International Nuclear Information System (INIS)

    In Romania the nuclear energy is considered an important component of energy mix and for a country sustainable development. Presently based on PHWR CANDU technology, the research and development activities, part of the national nuclear power programme, provide an increased effort towards generation 4, dedicated to support fast reactor lead technology. In the European framework (EU R&D Framework Programmes, European Sustainable Nuclear Industry Initiative) devoted to the development of GenerationIV technologies, Romania is contributing as a partner in EU R&D projects together with a large number of EU research organizations and in the ALFRED MoU, having ANSALDO Nucleare, ENEA and INR as initial members. ALFRED (Advanced Lead Fast Reactor European Demonstrator) project aims to build a 125MWe lead cooled fast reactor demonstrator, connected to the electrical network, with a target date for operation start-up in 2025. In February 2011 Romanian Government approved the option to host ALFRED demonstrator. Based on the access to European structural funds, existing nuclear experience and EU orientation to build the demonstrators in the new member states, Romania is an important option for siting process. An investigation of the existing national capabilities, identification of additional infrastructure and identification of the professional development needs in order to prepare the siting national support are presented in the paper. The main approaches and needed resources to meet expected requirements for ALFRED implementation are discussed as well. (author)

  15. Methods and tools to detect thermal noise in fast reactors

    International Nuclear Information System (INIS)

    The Specialists' Meeting on ''Methods and Tools to Detect Thermal Noise in Fast Reactors'' was held in Bologna on 8-10 October 1984. The meeting was hosted by the ENEA and was sponsored by the IAEA on the recommendation of the International Working Group on Fast Reactors. 17 participants attended the meeting from France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, Joint Research Centre of CEC and from IAEA. The meeting was presided over by Prof. Mario Motta of Italy. The purpose of the meeting was to review and discuss methods and tools for temperature noise detection and related analysis as a potential means for detecting local blockages in fuel and blanket subassemblies and other faults in LMFBR. The meeting was divided into four technical sessions as follows: 1. National review presentations on application purposes and research activities for thermal noise detection. (5 papers); 2. Detection instruments and electronic equipment for temperature measurements in fast reactors. (5 papers); 3. Physical models. (2 papers); 4. Signal processing techniques. (3 papers). A separate abstract was prepared for each of these papers

  16. Application of nitrogen alloyed steels for Indian Fast Reactor programme

    International Nuclear Information System (INIS)

    Towards building fast reactors for fulfilling energy requirements through second stage of nuclear power program planned by Department of Atomic Energy, a 500 MWe Prototype Fast Breeder Reactor (PFBR) is under advanced stage of construction at Kalpakkam, a coastal site. Nitrogen alloyed types 304LN and 316LN austenitic Stainless Steels have been selected for out of core components except for the steam generator primarily due to inclusion in the design codes favourable effect of nitrogen on mechanical strength and sensitization, and excellent weldability. For the once through steam generator design selected from economics and safety, modified 9Cr-1 Mo (Gr 91) has been selected from inclusion in the design codes, adequate mechanical strength, sound industrial experience and carbon transfer considerations. The presentation highlights the application of nitrogen alloyed types 304LN and 316LN SS, as well as modified 9Cr-1Mo steel for PFBR, and the influence of increased nitrogen alloying on mechanical properties on SS 316L for application to future fast reactors. (author)

  17. Optimization of ultra-long cycle fast reactor core

    International Nuclear Information System (INIS)

    An optimization of an ultra-long cycle fast reactor (UCFR) design with a power rate of 1000 MW (electric), UCFR-1000, has been performed to increase the safety of UCFR. Firstly, geometric optimization has been performed to decrease its peaking factors so that the peak temperatures measured by thermal hydraulic feedback are within the limit of design basis event (DBE). Secondly, fuel composition optimization has been performed by adopting Pressurized Water Reactor (PWR) spent fuel as a blanket material instead of natural uranium. Lastly, a small-size UCFR with a power rate of 100 MWe, UCFR-100, has been proposed for developing a short term deployable nuclear reactor. The major optimization process for UCFR-100 is decreasing maximum neutron flux and fast neutron fluence. The optimized UCFR-1000 has been enlarged radially and shortened axially from the initial UCFR design and this modification makes the burning speed of active core movement slower. It has been confirmed that a full-power operation of 60 years without refueling is feasible for both UCFR-1000 and UCFR-100 core designs by a breed-and-burn strategy. By the design optimization study, the reductions of maximum neutron flux, fast neutron fluence, and axial power peaking have been achieved, which are favorable for the safety of the UCFR. (author)

  18. Analysis of thorium/U-233 lattices and cores in a breeder/burner heavy water reactor

    International Nuclear Information System (INIS)

    Due to the inevitable dwindling of uranium resources, advanced fuel cycles in the current generation of reactors stand to be of great benefit in the future. Heavy water moderated reactors have much potential to make use of thorium, a currently unexploited resource. Core fuelling configurations of a Heavy Water Reactor based on the self-sufficient thorium fuel cycle were simulated using the DRAGON and DONJON reactor physics codes. Three heterogeneously fuelled reactors and one homogeneously fuelled reactor were studied. (author)

  19. Role of small lead-cooled fast reactors for international deployment in worldwide sustainable nuclear energy supply

    International Nuclear Information System (INIS)

    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) has been under ongoing development as part of the US advanced nuclear energy systems programs. Meeting future worldwide projected energy demands during this century (e.g., 1000 to 2000 GWe by 2050) in a sustainable manner while maintaining CO2 emissions at or below today's level will require massive deployments of nuclear reactors in non-fuel cycle states as well as fuel cycle states. The projected energy demands of non-fuel cycle states will not be met solely through the deployment of Light Water Reactors (LWRs) in those states without using up the world's resources of fissile material (e.g., known plus speculative virgin uranium resources = 15 million tonnes). The present U.S. policy is focused upon domestic deployment of large-scale LWRs and sodium-cooled fast spectrum Advanced Burner Reactors (ABRs) working in a symbiotic relationship that burns existing fissile material while destroying the actinides which are generated. Other major nuclear nations are carrying out the development and deployment of SFR breeders as witness the planning for SFR breeder deployments in France, Japan, China, India, and Russia. Small (less that 300 MWe) and medium (300 to 700 MWe) size reactors are better suited to the growing economies and infrastructures of many non-fuel cycle states and developing nations. For those deployments, fast reactor converters which are fissile self-sufficient by creating as much fissile material as they consume are preferred to breeders that create more fissile material than they consume. Thus

  20. A review of the UK fast reactor programme, March 1981

    International Nuclear Information System (INIS)

    A reduction in electricity sales over the last year had led to some fossil-fuelled stations being prematurely retired and has postponed the start of some new stations. Nevertheless the main programme for the building of 1 5 G We of thermal reactors during the next ten years remains unaltered and as summarised in last year's review. A formal request to build the first PWR at Sizewell in Suffolk has been presented by the CEGB to the government. This is expected to lead to a Public Inquiry within the next 12-18 months. The major contracts for building the AGR stations at Torness and Heysham were placed recently. Reduced projections for electricity demand up to the end of the century have also contributed to a delay in the government's response to the recommendation by the industry that the Commercial Demonstration Fast Reactor (CDFR) should be built to ensure that the option for commercial LMFBR should be demonstrated and maintained. A government statement is now expected before the end of 1981. Fast breeder reactors are expected to be required in the UK electrical supply system by about the turn of the century. Series ordering will be preceded by construction and operation of the CDFR, of a design suitable in all basic features for replication in programme reactors. The National Nuclear Corporation (NNC) has continued the development of a CDFR design having the required operational safety and economic characteristics. The basic design concept is now nearing completion following investigation of a number of alternatives. Some of the more important features of this design, namely the core, primary circuit and reactor cooling systems. steam cycle and boilers, and overall plant and station layout are described in this review. As a result of increased understanding of sodium/water reaction behaviour, development of manufacturing and inspection techniques and experience in plugging and repair of tubes containing leaking welds, coupled with the preference for a once through

  1. Sophistication of burnup analysis system for fast reactor (2)

    International Nuclear Information System (INIS)

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  2. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  3. A review of the United Kingdom fast reactor program - March 1983

    International Nuclear Information System (INIS)

    A review of the United Kingdom Fast Reactor Programme was given in March 1983. Operational experience with the Prototype Fast Reactor (PFR) is briefly summarized. The design concept of the Commercial Demonstration Fast Reactor (CDFR), including design codes, engineering components, materials and fuels development, chemical engineering/sodium technology, safety and reactor performance, is reviewed. The problems of PFR and CDFR fuel reprocessing are also discussed

  4. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    Baldev Raj; S L Mannan; P R Vasudeva Rao; M D Mathew

    2002-10-01

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of fissile material than in thermal reactors, with a matching increase in burn-up. The design of the fuel is an important aspect which has to be optimised for efficient, economic and safe production of power. FBR components operate under hostile and demanding environment of high neutron flux, liquid sodium coolant and elevated temperatures. Resistance to void swelling, irradiation creep, and irradiation embrittlement are therefore major considerations in the choice of materials for the core components. Structural and steam generator materials should have good resistance to creep, low cycle fatigue, creep-fatigue interaction and sodium corrosion. The development of carbide fuel and structural materials for the Fast Breeder Test Reactor at Kalpakkam was a great technological challenge. At the Indira Gandhi Centre for Atomic Research (IGCAR), advanced research facilities have been established, and extensive studies have been carried out in the areas of fuel and materials development. This has laid the foundation for the design and development of a 500 MWe Prototype Fast Breeder Reactor. Highlights of some of these studies are discussed in this paper in the context of our mission to develop and deploy FBR technology for the energy security of India in the 21st century.

  5. Review of fast reactor activities at OECD (NEA), March 1979

    International Nuclear Information System (INIS)

    continue to rise. These facts argue strongly not only for the continued development and commercial demonstration of the breeder, but also for the supporting services required to ensure its success. This report contains analysis of the following issues: gas cooled reactors; safety of LMFBRs, fast reactor physics; fast reactor nuclear data, and activities of the NEA Data Bank

  6. Minor actinides transmutation strategies in sodium fast reactors

    International Nuclear Information System (INIS)

    In minor actinides transmutation strategies for fast spectrum reactors, different possibilities regarding the core loading are considered. We study both homogeneous patterns (HOM) with various minor actinides (MA) content values and heterogeneous schemes (HET) with higher percentages of MA (Np, Am and Cm) at the periphery of reactor. We analyze the capability of transmutation of each design and the reactivity coefficients such as the Doppler constant, void worth and the fraction of delayed neutrons. The EVOLCODE2 code is the computational tool used in this study. It is based on MCNPX and ORIGEN/ACAB codes and allows carrying out burn-up calculations to get the isotopic evolution of fuel composition. Among the three strategies studied (HOM 2.5 %, HOM 4% and HET 20 %) for a possible design of a Sodium Cooled Fast Breeder Reactor, the one with better transmutation results is the HOM 4%, which shows higher absolute and relative values (12 Kg-MA/TWe, 29% respectively). Concerning transmutation in blankets with 20% MA content, results show a very little or no transmutation values when considering Np, Am and Cm together, though a positive small value for Np and Am is obtained

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

    Science.gov (United States)

    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.

  8. Status of fast breeder reactor development in the United States

    International Nuclear Information System (INIS)

    The energy policy of the United States is aimed at shifting as rapidly as practicable from an oil dependent economy to one that relies heavily on other fuels and energy sources. Nuclear power Is now and is expected to continue to be an important factor in achieving this goal. If nuclear power is to contribute to a solution of future energy needs, demonstration of the breeder reactor as a viable source of essentially inexhaustible energy supply is essential. The US DOE program for development of the fast breeder reactor has witnessed some notable events in the past year. Foremost among these Is the successful operational testing of the Fast Flux Test Facility (FFTF), located at.the Hanford Engineering Development Laboratory. The reactor reached full design power of 400 MW(t) on December 21, 1980, and has performed remarkably close to design specifications. Design of the Clinch River Breeder Reactor Plant (CRBRP), a 375 MW(e) LMFBR, is now over 80 percent complete. About $530 million in components have been ordered; component deliveries total approximately $124 million; work-in-process totals another $204 million. Construction of the plant, however, has been suspended since 1977. With the concurrence of the U.S. Congress and approvals from the appropriate authorities work on the safety review and site clearing for construction can resume. The Conceptual Design Study for a large, 1000 MW(e) LMFBR Large Developmental Plant was recently completed on a schedule commensurate with submission of a full report to the Congress at the end of March, 1981. This report is the culmination of a study which began in October, 1978 and involved contributions from U.S. reactor manufacturers and US DOE laboratories. The US DOE is carrying forward a comprehensive technology development program. This effort provides direct support to the FFTF and CRBRP projects and to the LDP. It also supports technology development which is generic to the overall LMFBR program. Funding for breeder

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

    International Nuclear Information System (INIS)

    Research and development in fast reactor reprocessing has been under way ∼ 20 yr in several countries. During the past decade, France and the United Kingdom have developed active programs 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 Experimental Breeder Reactor II (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. The Federal Republic of Germany (FRG) 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 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 focuses 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

  10. A review of the Indian fast reactor programme

    International Nuclear Information System (INIS)

    Development of Fast Breeder activities is being done mainly at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam and the total Scientific and Technical staff working at the Centre for development of FBRs is about 1200. The development work relating to the fuel fabrication and design and development for some of the fuel handling equipment is being done at the Bhabha Atomic Research Centre, Trombay, Bombay. Complete recovery from the fuel handling incident of FBTR was achieved during the beginning of 1989. Damaged guide tube and bent subassemblies were replaced, the incident was analysed in detail and appropriate remedial measures, viz., modifications in the fuel handling machine control logic and plug rotation logic were implemented to prevent its recurrence. Safety clearances for the restart of the reactor were obtained from the Atomic Energy Regulatory Board in May 1989. As steam generators were not valved in the secondary sodium system, the reactor power during this phase of operation was limited to 500 KWt. The main objectives during this phase were to complete the balance low power physics experiments and to operate the reactor for a sufficiently long time to assess the performance of various systems, in particular the neutronic instrumentation, control rod drive and safety logic system which were not in active service for the two years. From May to July, 1989, the reactor was successfully operated up to a power level of 500 KWt with 50% operating time. Design of PFBR is progressing intensively. (author). 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    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.

  12. SIMMER-III modeling of gas cooled fast reactor

    International Nuclear Information System (INIS)

    This paper deals with extension and application of the SIMMER-III code for safety studies of a gas cooled fast reactor. The equation of state of the helium gas and its thermal physical properties have been prepared and implemented in the code. The geometric, thermal hydraulic and neutronic models have been set up for the ALLEGERO reactor. The code and the associated model are verified by comparing steady state and unprotected loss of flow 20% remained flow rate (ULOF-20%) results with those done by other project partners. Reasonable or good agreements have been achieved for major physical variables. The unprotected loss of coolant accident (ULOCA) case is a severe transient case with core melting and degradation that was emulated only by SIMMER, in the project. In the initiating phase the clad becomes molten, this triggers the first power excursion. Then the fuel becomes more mobile and further power excursions take place, which lead to core melting and degradation. The fuel is ejected by power excursion and then moves relatively slowly to the lower part of vessel. Finally there are only a few kilograms of fuel escaping to the vessel outside (into reactor container) and the released thermal energy is about 6 GJ within a period of one minute. The final power stays below one MW and the reactor is in a deep sub-criticality state, since 1/2 fuel becomes noneffective. (author)

  13. Fast Pyrolysis of Lignin Using a Pyrolysis Centrifuge Reactor

    DEFF Research Database (Denmark)

    Trinh, Ngoc Trung; Jensen, Peter Arendt; Sárossy, Zsuzsa;

    2013-01-01

    at temperatures of 500−550 °C, reactor gas residence time of 0.8 s, and feed rate of 5.6 g/min. Gas chromatography mass spectrometry and size-exclusion chromatography were used to characterize the Chemical properties of the lignin oils. Acetic acid, levoglucosan, guaiacol, syringols, and p-vinylguaiacol are found...... to be major chemical components in the lignin oil. The maximal yields of 0.62, 0.67, and 0.38 wt % db were obtained for syringol, p-vinylguaiacol, and guaiacol, respectively. The reactor temperature effect was investigated in a range of 450−600 °C and has a considerable effect on the observed chemical......Fast pyrolysis of lignin from an ethanol plant was investigated on a lab scale pyrolysis centrifuge reactor (PCR) with respect to pyrolysis temperature, reactor gas residence time, and feed rate. A maximal organic oil yield of 34 wt % dry basis (db) (bio-oil yield of 43 wt % db) is obtained...

  14. Fast reactor development for sustainable nuclear energy supply in China

    International Nuclear Information System (INIS)

    China needs a very huge energy supply for national economy development and living standard improvement of 1.3 billion population. The nuclear energy is a new member of the energy supply family in China. A satisfied operation records of all 11 units of NPPs, especially with the total average load factor 85.8% of all NPPs in 67 reactor-years since commercial operation of each unit encourage the public to believe that the nuclear power is a safe, reliable, economically-acceptable, CO2 avoidable one and could be used in large scale. The government has decided in 2006 to accelerate the nuclear power development with the target of 40GWe in operation and 18GWe in construction in 2020.Right now 13 units with total capacity 13.05GWe are under construction and other 11 units of total capacity 12.05GWe have been approved by the government and the preparation for construction is underway. For the sustainable supply of nuclear energy, as the principle strategy, PWR-FBR matched with closed nuclear fuel cycle has been decided by the government for a long time. Three FBR development strategy targets suggested as following. (1) to realize FBR commercial utilization in small batch in 2030; (2) to increase nuclear capacity to 240GWe, sharing about 16%, mainly by FBRs in 2050, and (3) to replace coal fired plants by nuclear power in large scale, in the period about 2050-2100. For that, the suggested FBR development strategy is shown in Table 1. China Experimental Fast Reactor (CEFR) with the power 65MWt is a pool type sodium-cooled fast reactor. Pre-conceptual design started in 1990 with the first pot of concrete in 2000, and architecture engineering launched in 2001. Now it is under commissioning tests stage. The experiences of design fabrication and construction for this type of reactor have been gained. The possibility of large striding from 20MWe CEFR to 800MWe China Demonstration Fast Reactor (CDFR) has been studied. The favorableness is estimated mainly as following. (1) The

  15. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

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

    International Nuclear Information System (INIS)

    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.

  17. The story of the European fast reactor cooperation

    International Nuclear Information System (INIS)

    This report is a condensed history of European cooperation in the large breeder power plants with powers in excess of 1000 MWe. The beginning, in 1973, was marked by the so-called Utilities' Convention signed by EdF, RWE, and ENEL on the construction of Superphenix and SNR 2. In 1977, cooperation began among the reactor vendors and R and D organizations in France, Germany and Italy as well as Belgium and the Netherlands. After the British had joined in 1984, planning for the European Fast Reactor, EFR, was started in 1988. The conceptual design phase of the 1500 MWe breeder power plant covered a period of five years and was concluded with an economic assessment and a technical safety analysis of EFR in 1983. A number of ongoing studies are being conducted within a specific EFR program. (orig.)

  18. Thermophysical and thermochemical properties of fast reactor materials

    International Nuclear Information System (INIS)

    The physical and chemical properties of materials occurring within the core of a liquid sodium cooled fast breeder reactor (LMFBR) are reviewed. Properties particularly relevant to predicting the reactor's behaviour under various accident scenarios and during normal operations are considered and recommendations in a form suitable for use in computer codes which model such situations are put forward. Included in the review are the following properties: (a) Oxide fuels: density and thermal expansion, temperature of fusion, enthalpy and specific heat, vapour pressure, viscosity, diffusion and creep, emissivity, thermal conductivity, surface tension and energy. (b) AISI M316 stainless steel: composition, thermal expansion, temperature of fusion, enthalpy and specific heat, thermal conductivity, vapour pressure, viscosity. (c) Sodium: density, enthalpy and specific heat, vapour pressure, emissivity, diffusion, thermal conductivity, surface tension, viscosity, equation of state, critical constants

  19. Optimisation of safety parameters in fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Optimisation of safety parameters is an important aspect to be considered in the design of nuclear power plant and also becomes extremely important activity to be followed up during the commissioning and operating phases of the plant taking into account the operational feed back and review of incidental situations and available diversity and reliability. Otherwise, the spurious/ superfluous trips on the reactor besides affecting the availability of the plant, initiate plant transients causing stress for the plant equipment resulting in reduction of plant life. This activity has a significant role to play in attaining the maximum availability of the plant, without compromising safety. The study and evolution of optimisation process in fast breeder test reactor (FBTR); at Kalpakkam has been an interesting and rewarding experience

  20. Gas cooled fast reactor background, facilities, industries and programmes

    International Nuclear Information System (INIS)

    This report was prepared at the request of the OECD-NEA Coordinating Group on Gas Cooled Fast Reactor Development and it represents a contribution (Vol.II) to the jointly sponsored Vol.I (GCFR Status Report). After a chapter on background with a brief description of the early studies and the activities in the various countries involved in the collaborative programme (Austria, Belgium, France, Germany, Japan, Sweden, Switzerland, United Kingdom and United States), the report describes the facilities available in those countries and at the Gas Breeder Reactor Association and the industrial capabilities relevant to the GCFR. Finally the programmes are described briefly with programme charts, conclusions and recommendations are given. (orig.)

  1. Seismic response analysis of the PEC fast reactor building

    International Nuclear Information System (INIS)

    In order to compute the motion induced by the design earthquakes at the vessel supporting structure, a seismic response analysis was performed for the PEC fast reactor, taking into account the effects of soil-structure interaction by use of experimentally determined soil parameters. The main aim of he analysis was to evaluate the safety margins present in the design calculations. A detailed 3D finite element model was set up for fixed base analysis; from the results of the 3D model a simplified equivalent model of the structure was then derived for soil-structure interaction analysis. The mathematical model was validated and calibrated by using the results of the experimental dynamic tests performed on the reactor building. The results have shown the adequacy of the computation methodologies, and in particular of those on the use of the equivalent model. (author)

  2. Description of the prototype fast reactor at Dounreay

    International Nuclear Information System (INIS)

    The Prototype Fast Reactor (PFR) at Dounreay, UK, started operation in 1975 and was closed down in 1994. The present report contains a description of the PFR nuclear power plant, based on information available in literature and on information supplied during a visit to the plant. The report covers a description of the site and plant arrangement, the buildings and structures, the reactor core and other vessel internals, the control system, the main cooling system, the decay heat removal system, the emergency core cooling system, the containment system, the steam and power conversion system, the fuel handling system, plant safety features, the control and instrumentation systems and the sodium purification systems. (au) 16 refs

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

    International Nuclear Information System (INIS)

    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.

  4. Fast reactors, key elements of a sustainable nuclear energy

    International Nuclear Information System (INIS)

    The criteria that define the concept of sustainable energy are first the non-exhaustion of the resource, secondly the mitigation of the pollution during all the stages from the extraction/transformation of the resource to the use as energy and thirdly the right management of the waste in a view of avoiding any nuisance for future generations. In this article fossil energies, renewable energies and nuclear energie are judged by these 3 criteria. It appears that no energy can satisfy the 3 criteria but the introduction of fast reactors in a fleet of light water reactors associated with a program of spent fuel reprocessing turn nuclear energy into a sustainable energy on a long-term perspective. (A.C.)

  5. Safeguards in the prototype fast breeder reactor MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Usami, S.; Deshimaru, T.; Tomura, K. [Power Reactor and Nuclear Fuels Development Corporation, Ibaraki-ken (Japan)

    1995-12-31

    MONJU is a prototype fast breeder reactor in Japan designed to have a 280-MW(electric) output. The Power Reactor and Nuclear Fuel Development Corporation (PNC) started its construction in the autumn of 1985 in Tsuruga. The loading of the core fuel assemblies was started in October 1993, and the preoperational test is ongoing. MONJU uses 198 mixed-oxide (MOX) fuel assemblies as core fuel and 172 depleted uranium assemblies as blanket fuel. Assemblies loaded in-core and stored in the ex-vessel storage tank (EVST) reside in liquid sodium. These plutonium-containing fuel assemblies, MOX, and irradiated depleted uranium are regarded as in the difficult-to-access area, and the flows of fuel assemblies into and out of the area must be verified. Flow is verified by fuel flow monitors measuring radiation, which can limit inspector attendance during fuel handling.

  6. Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs

  7. Status of liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    This document represents a compilation of the information on the status of fast breeder reactor development. It is intended to provide complete and authoritative information for academic, energy, industrial and planning organizations in the IAEA Member States. The Report also provides extended reference and bibliography lists. A summarized overview of the national programmes of LMFBR development is given in Chapter II. Chapter III on LMFBR experience provides a brief description and purpose of all fast reactors - experimental, demonstration and commercial size - that have been or are planned for construction and operation. Fast reactor physics is dealt with in Chapter IV. Besides the basic facts and definitions of neutronics and the compilation and measurement of nuclear data, a broad range of the calculation methods, codes, and the state of the art is described. In Chapter V, fuels and materials are described. The emphasis is on the design and development experience gained with mixed oxide fuel pins and subassemblies. Structural materials, blanket elements and absorber materials are also discussed. Chaper VI presents a broad overview of the technical and engineering aspects of LMFBR power plants. LMFBR core design is described in detail, followed by the components of the main heat transport system, the refuelling equipment, and auxiliary systems. Chapter VII on safety is a compilation of the current safety design concepts of LMFBRs and new trends in safety criteria and safety goals. The chapter concludes with risk analyses of LMFBR technology. In Chapter VIII, the systems approach has been emphasized in the consideration of the whole LMFBR fuel cycle. Special emphasis is placed on safeguards aspects and the environmental impact of the LMFBR fuel cycle. Chapter IX describes deployment considerations of LMFBRs. Special emphasis is placed on economic aspects of the LMFBR power plant and its related fuel cycle. Finally, Chapter X provides an overall summary and a

  8. Knowledge Management in Fast Reactors and Related Fuel Cycles

    International Nuclear Information System (INIS)

    The 21st century is ushering in a new phase of economic and social development which can be referred as 'Knowledge Economy' in which knowledge has become the key asset in determining the organization's success or failure. Nuclear technology is very complex and a highly technical endeavor. It relies on innovative creation, storage and dissemination of knowledge. The nuclear technology is also characterized by long time scales and technological excellence. Nuclear Knowledge Management is a critical input to Nuclear Power Industry, the associated fuel cycle activities and nuclear applications in medicine, industry and agriculture. In an R and D Organization like Indira Gandhi Centre for Atomic Research (IGCAR) specializing in the areas of Fast Reactor Technology and associated Fuel Cycle Facilities, Knowledge Management plays a vital role. IGCAR is operating successfully the Fast Breeder Test Reactor (FBTR) for the last 24 years with a unique Pu-U Carbide Fuel. The paper highlights the various success stories, lessons learnt from FBTR, knowledge accrued, disseminated and reused. With intensive R and D and innovations, the processes have been developed and FBTR's spent carbide fuel of 155 GWd/t burn up has been reprocessed successfully. The paper covers the knowledge that has been created through extensive analysis and validation for the design of a 500 MWe Prototype Fast Breeder Reactor (PFBR) which is under construction at Kalpakkam. The Centre has developed world class expertise in the areas of sodium technology, material development, non-destructive evaluation, instrumentation etc. This paper gives some examples of how the knowledge generated is used for PFBR. (author)

  9. Fast Reactor Research in Europe: The Way Towards Sustainability (Summary)

    International Nuclear Information System (INIS)

    Full text: The European Union (EU) has taken the lead in responding to climate change, announcing far-reaching initiatives ranging from promoting energy efficient light bulbs and cars to new building codes, carbon trading schemes, development of low carbon technologies and greater competition in energy markets. Nuclear energy remains central to the energy debate in Europe. One third of EU electricity is produced via nuclear fission and eight new reactors are under construction. Traditionally non-nuclear countries are manifesting an interest in building nuclear power plants while the clock is ticking down on Belgium, Germany and the United Kingdom's decision to renew or close existing nuclear infrastructures. Sustainability in nuclear energy production is ensured in the medium term as a result of the large and diverse uranium resources available in politically stable countries around the world. The quantities available with high probability ensure more than one hundred years of nuclear energy production. This extrapolation depends, however, on the forecast for future nuclear energy production. The use of fast neutron breeder reactors would lead to a much more efficient utilization of the uranium, extending the sustainable energy production to several thousands of years. The presentation will outline the fast reactors of the new generation currently being developed within the Generation IV initiative. Broad conclusions of the presentation are that: - There is a growing nuclear renaissance in Europe for good reason; - Nuclear energy is a green and sustainable option for Europe and indeed the world's energy needs; - Nuclear energy is a competitive energy that makes economic sense; - Nuclear fission reactors have a safety and environmental track record that is second to none, yet public misperceptions persist and must be tackled; - Waste management solutions exist while new developments hold great promise; - The evolution and promise of nuclear technologies must also be

  10. Evaluation of the breed/burn fast reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Atefi, B.; Driscoll, M.J.; Lanning, D.D.

    1979-12-01

    A core design concept and fuel management strategy, designated breed/burn, has been evaluated for heterogeneous fast breeder reactors. In this concept internal blanket assemblies after fissile material is bred in over several incore cycles, are shuffled into a moderated radial blanket and/or central island. The most promising materials combination identified used thorium in the internal blankets (due to the superior performance of epithermal Th-U233 systems) and zirconium hydride (ZrH/sub 16/) as the moderator (because of the compact assembly and core designs it permitted).

  11. Fast reactor core concepts to improve transmutation efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Fujimura, Koji; Kawashima, Katsuyuki [Hitachi Research Laboratory, Hitachi, Ltd., 7-1-1, Omika-cho, Hitachi-shi, Ibaraki, 319-1221 Japan (Japan); Itooka, Satoshi [Hitachi-GE Nuclear Energy, Ltd., 3-1-1, Saiwai-cho, Hitachi-shi, Ibaraki, 317-0073 Japan (Japan)

    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.

  12. A review of the European collaborative programme on fast reactors

    International Nuclear Information System (INIS)

    Considerable progress has been made in the European Fast Reactor R and D Collaboration during 1988. In parallel with the technical achievements, 1988 has been notable for the determination and willingness to adapt and improve the various structures and procedures in the organization of the European Collaboration. Three important agreements on EFR were signed in Bonn, Federal Republic of Germany on 16 February 1989: Industrial agreement, as a basis for planning, design, future construction and marketing of EFR; R and D agreement on the content and extent of R and D support for EFR design; SERENA-FASTEC agreement on know-how pool and royalties. (author). 1 tab

  13. Fast Traveling-Wave Reactor of the Channel Type

    CERN Document Server

    Rusov, Vitaliy D; Vashchenko, Volodymyr N; Chernezhenko, Sergei A; Kakaev, Andrei A; Pantak, Oksana I

    2015-01-01

    The main aim of this paper is to solve the technological problems of the TWR based on the technical concept described in our priority of invention reference, which makes it impossible, in particular, for the fuel claddings damaging doses of fast neutrons to excess the ~200 dpa limit. Thus the essence of the technical concept is to provide a given neutron flux at the fuel claddings by setting the appropriate speed of the fuel motion relative to the nuclear burning wave. The basic design of the fast uranium-plutonium nuclear traveling-wave reactor with a softened neutron spectrum is developed, which solves the problem of the radiation resistance of the fuel claddings material.

  14. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    The paper presents the current status of the Gas cooled Fast Reactor system development which is shared within the Generation IV International Forum including EURATOM through the 7th Framework Programme project GoFastR. The various areas considered will include suitable fuel compounds and high temperature resistant cladding materials options, core design optimisation, primary system boundary, energy conversion. The safety approach, mainly oriented on core cooling for the moment, will be recalled together with a discussion of the results obtained. Further potential improvements or simplification of the system safety, at the light of the Fukushima accident, including an indirect coupled cycle for the energy conversion and a self sustainable Decay Heat Removal loop will be mentioned. The main issues related to the necessary R&D programme accompanying the system development will be recalled (fuel and materials, helium coolant technology, components such as gas circulators, valves and heat exchangers, thermal barriers). (author)

  15. Current status on fast reactor program in Kazakhstan

    International Nuclear Information System (INIS)

    Spent Fuel Assemblies Drum (SFAD) was successfully drained into several tanks for storage, and the SFAD itself was dismantled. The Project on the utilization of liquid radioactive wastes (LRW), which were accumulated during the operation of BN-350 reactor, is going according to schedule. Special designed cesium traps have been manufactured. Works on cesium decontamination of BN-350 primary circuit sodium commenced in December 2002 and are ongoing. The main results obtained to date are: the amount of Cs-137 trapped by 5 traps is approximately 10,000 Ci, and the Cs concentration in the primary sodium decreased by two orders of magnitude 100 (from 10-3 to 10-5 Ci/kg). The plans for the Project on primary circuit sodium draining, including the usage of a drilling device have been developed and reviewed. All preparatory drilling tests have been completed. According to the plans, the primary sodium should be drained before end of 2003. The EAGLE project comprises the preparation and conduct of out-of-pile and in-pile experiments designed to address the key safety issues relevant to eliminating or mitigating the re-criticality potential during a postulated core-disruptive accident in future commercial sodium-cooled fast reactors. The work has been performed since 2000 under a contract between NNC RK and JNC (Japan Nuclear Cycle Development Institute), after some related feasibility studies for the experiments funded by JAPC (Japan Atomic Power Company)

  16. The Last Twenty Years of Experience with Fast Reactors in Japan

    International Nuclear Information System (INIS)

    Fast reactor development experience gained in Japan in the last twenty years is summarized in this paper. In this twenty years, the safety, reliability and economic goals of fast reactors have become more ambitious than in the past. However, twenty years of progress have shown that the domestic commercialized sodium cooled fast reactor (SFR) concept, the Japanese SFR, could achieve those targets discussed in the Feasibility Study on Commercialized Fast Reactor Cycle Systems (FS) and the Fast Reactor Cycle Technology Development (FaCT) projects. The Monju prototype fast breeder reactor is finally going to restart by the end of this Japanese fiscal year (March 2010) and will take on the role of a technology and human resource development centre from both a domestic and an international point of view. (author)

  17. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

  18. A Study of Reactor Neutrino Monitoring at Experimental Fast Reactor JOYO

    CERN Document Server

    Furuta, H; Hara, T; Haruna, T; Ishihara, N; Ishitsuka, M; Ito, C; Katsumata, M; Kawasaki, T; Konno, T; Kuze, M; Maeda, J; Matsubara, T; Miyata, H; Nagasaka, Y; Nitta, K; Sakamoto, Y; Suekane, F; Sumiyoshi, T; Tabata, H; Takamatsu, M; Tamura, N

    2011-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3m from the JOYO reactor core of 140MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11\\pm1.24(stat.)\\pm0.46(syst.)events/day. Although the statistical significance of the measurement was not enough, the background in such a compact detector at the ground level was studied in detail and MC simulation was found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

  19. Fast breeder reactors: experience and trends. V. 1

    International Nuclear Information System (INIS)

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium presentations were divided into sessions devoted to the following topics: Experience of LMFBR construction and operation and resultant development strategies (6 papers); LMFBR plant startup and commissioning tests and general behaviour (8 papers); Core performance experience for high burnup and core design trends (8 papers); Experience and trends in the LMFBR fuel cycle (4 papers); Core design and behaviour (3 papers); Fuels and materials (7 papers). A separate abstract was prepared for each of these papers

  20. Nuclear data requirements for fast and intermediate reactor calculations

    International Nuclear Information System (INIS)

    During 1960/61, some work has been done at Karlsruhe in the compilation of reliable nuclear data for fast- and intermediate-reactor calculations. Materials included thus far are He, O16, Na23, Cr, Fe, Ni, Mo, U235, U238 and Pu239. For fast- and intermediate-reactor optimization studies, reliable theoretical prediction of space-dependent neutron energy spectra, critical masses, breeding ratios, burn-up, Doppler coefficients and related subjects, the author was interested in much more detailed microscopic cross-section data concerning mainly neutron absorption and fission than are needed, for example, for the evaluation of few-group constants in the conventional multi-group diffusion theory. In spite of much progress in the experimental determination and theoretical interpretation of nuclear data, many inconsistencies in the experimental results of different authors and laboratories and large gaps in experimental nuclear-data work still remain. This paper discusses a great number of these gaps and inconsistencies in respect of the above-mentioned nuclei. (author)

  1. Scenarios for Fast Reactors Deployment with Plutonium Recycling

    International Nuclear Information System (INIS)

    Plutonium-management is crucial towards sustainability of nuclear energy and demands a progressive recycling policy starting in LWRs, by use of MOX-fuel, up to the multi-recycling of Pu in Fast Reactors, such as Sodium Fast Reactors (SFRs). This paper focuses on the transition where SFRs perform a multi-recycling of Pu. In order to sketch a few of the possible futures towards sustainability, illustrative scenarios for SFR deployment in progressive replacement of the French PWR fleet with different chronologies (from 2040 or 2080) were analyzed in the context of the French 2006 Law. The sensitivity of scenarios to the SFR core design is evaluated by considering a homogeneous core (SFR V2B) or a new heterogeneous core with a significant gain on sodium void effect (CFV). Scenarios of symbiotic fleet, in which a 60 GWe PWR fleet is maintained while SFRs are progressively introduced from 2040 depending on the plutonium produced by the PWR fleet, are also envisaged. The first evaluation focuses on the maximal achievable installed SFR power using plutonium from PWR spent fuel. Then, different key drivers to modulate SFR deployment are presented (PWR and SFR spent fuel cooling time, reprocessing capacities). These studies allow to identify the minimal installed SFR power in function of the Pu yearly produced by PWR fleet. (author)

  2. Fast Shutdown System tests in the Georgia Tech Research Reactor

    International Nuclear Information System (INIS)

    The Fast Shutdown System (FSS) is a new safety system design concept being considered for in installation in the Savannah River (SRS) production reactors. This system is expected to mitigate the consequences of a Design Basis Loss of Coolant Accident, and therefore allow higher operational power levels. A test of this system in the Georgia Tech Research Reactor is proposed to demonstrate the efficacy of this concept. Three tests will be conducted at full power (5MW) and one at low power (100kw). Two full power tests will be conducted with the FSS rod backfilled with one (1) atmosphere of He-4, and one with the rod evacuated. The low power conducted with the FSS rod evacuated. Neutron flux and pressure data will be collected with an independent data acquisition system (DAS). Safety issues associated with the performance of the Fast Shutdown System experiments are addressed in this report. The credible accident scenarios were analyzed using worst case scenarios to demonstrate that no significant nuclear or personnel safety hazards would result from the performance of the proposed experiments

  3. IRPhE-TAPIRO-ARCHIVE, Fast neutron source reactor primary documents, reactor physics experiments

    International Nuclear Information System (INIS)

    Description of program or function: The TAPIRO reactor, located in the ENEA Casaccia Centre near Rome, is a highly enriched uranium fast neutron facility. The nominal power is 5 kW (thermal) and the core centre neutron flux is 4. E12/cm2/s. The reactor has a cylindrical core (12.6 cm diameter and 10.9 cm height) made of 93.5 % enriched uranium metal in a uranium-molybdenum alloy which is totally reflected by copper. The copper reflector (cylindrical-shaped) is divided into two concentric zones: the inner zone, up to 17.4 cm radius, and the outer zone up to 40.0 cm. Radius. The height of the reflector is 72.0 cm. The reactor is surrounded by borate concrete shielding about 170 cm thick. The maximum depth available for the epithermal column is 160 cm, reserved for filter/moderator materials. The graphite column extends to the external reflector boundary where a sector of the outer copper reflector has been removed and then characterized by a very hard neutron spectrum. Along the column the spectrum gradually softens up to thermal values - Different materials can be interposed, such as U-nat, Pb, Fe, etc. to reproduce spectrum transition conditions at interface points between regions with different compositions. - Activation foils can be used for activation analysis with threshold energies in the fast, intermediate and epithermal regions. The archive contains reports characterising the reactor and describes experiments carried out, together with the corresponding data

  4. Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

    Science.gov (United States)

    Stacey, W. M.

    2009-09-01

    The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

  5. A review of fast reactor activities in Switzerland - March 1984

    International Nuclear Information System (INIS)

    As a result of the noncentralized government in Switzerland there is no clear national policy for the future application of nuclear energy. This is reflected in the lack of a generally agreed nuclear energy research policy in the country. Consequently, activities related to several advanced reactor concepts are funded simultaneously at similar, but relatively low levels. The total expenditure of 5.9 million Swiss Francs (approx. 1 SFr per capita) for fast reactor activities in 1983 must be judged in the light of this situation. The funds have been allocated to an LMFBR safety programme (52%) and a fuel development programme (48%). In the field of LMFBR safety analytical work is performed on hypothetical core disruptive accidents (HCDAs) and on the integrity of components under HCDA loadings with emphasis on the dynamic behaviour of the reactor cover. A considerable effort has recently been devoted to the preparations for the SONACO natural convection experiment. Another relatively new experimental activity, involving small-scale vapour explosions with freon and water, has produced evidence of interesting physical effects which are not in accord with the assumptions of current molten fuel-coolant interaction (MFCI) models. The fuel development programme has continued with the manufacture of spherepac mixed carbide fuel pins for an irradiation experiment in FFTF. However, the time scale of the experiment has suffered a set-back due to an accident in a glove box of the production line

  6. Nuclear data and multigroup methods in fast reactor calculations

    International Nuclear Information System (INIS)

    The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)

  7. Status of national programmes on fast reactors 1998/99. 32nd annual meeting. Working material

    International Nuclear Information System (INIS)

    Over the past 32 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1998/1999, as reported at the 32. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States

  8. Shape optimization of a sodium cooled fast reactor

    Science.gov (United States)

    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.

  9. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    In the power increase performance test of the experimental fast reactor ''Joyo'', which was in progress since April, the first stage of the rated thermal output of 50 MW has been accomplished on July 5. Thereafter, the continuous opeation test at 50 MW for 100 hours was performed for the verification of its overall operational performance from August 13 to 16. The safety evaluation for power increase up to 75 MW and 100 MW, which was under way since September, last year, was completed, and the power increase was licensed on September 20. Concerning the design of the prototype fast breeder reactor ''Monju'', the studies on the specifications of the Construction Preliminary Design (2) have been finished. In respect of the analysis and preparation of materials for the Safety Licensing by the Committee, the developments of the analytical codes for rupture propagation in the heat transfer tubes of steam generators and for decay heat have been conducted. In the construction site surveys, the third geological structure survey and beach deformation survey have all ended, while the meteorological and seismic observations, the prediction of the diffusion of drained warm water, the survey of river flow, etc. are now under way. A report on the survey conducted on the construction site in Shiraki was received by the Fukui prefectural government in July, and the copies of a report on the assessment of environmental effect were submitted in August to both the national government and the Fukui prefectural government. The situations of progress of the research and development works on reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structural materials, safety and steam generators are reported. (Nakai, Y.)

  10. RBEC lead-bismuth cooled fast reactor: review of conceptual decisions

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P.; Fomichenko, P.; Mikityuk, K.; Nevinitsa, V.; Shchepetina, T.; Subbotin, S.; Vasiliev, A. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    2001-07-01

    A concept of the RBEC lead-bismuth fast reactor-breeder is a synthesis, on one hand, of more than 40-year experience in development and operation of fast sodium power reactors and reactors with Pb-Bi coolant for nuclear submarines, and, on the other hand, of large R and D activities on development of the core concept for modified fast sodium reactor. The report briefly presents main parameters of the RBEC reactor, as a candidate for commercial exploitation in structure of the future nuclear power. (author)

  11. IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Presentations

    International Nuclear Information System (INIS)

    The fast reactor, which can generate electricity and breed additional fissile material for future fuel stocks, is a resource that will be needed when economic uranium supplies for the thermal reactors diminish. Further, the fast-fission fuel cycle in which material is recycled (a basic requirement to meet sustainability criteria) offers the flexibility needed to contribute decisively towards solving the problem of growing “spent” fuel inventories by greatly reducing the volume, the heat load and the radiotoxic inventory of high-level wastes that must be disposed of in long-term geological repositories. This is a waste management option that will play an increasingly important role in the future, and help to ensure that nuclear energy remains a sustainable long-term option in the world’s overall energy mix. In recognition of the fast reactor’s importance for the sustainability of the nuclear option, currently there is worldwide renewed interest in fast reactor technology development, as indicated, e.g., by the outcome of the Generation IV International Forum (GIF) technology review, which concluded with 3 out of 6 innovative systems to be fast reactors (gas cooled fast reactor, sodium cooled fast reactor, and heavy liquid metal cooled fast reactor), plus a potential fast core for a 4th concept, the super-critical water reactor. Currently, fast reactor construction projects are ongoing in India (PFBR) and Russian Federation (BN-800), whilst in China the first experimental fast reactor (CEFR) is in the commissioning phase. Fast reactor programs are also carried out in Europe (in particular in France), Japan, Republic of Korea and the USA. The most important challenges for fast reactors are in the areas of cost competitiveness with respect to LWRs and other energy sources, enhanced safety, non-proliferation, and public acceptance. With the exception of this latter, these translate into technology development challenges, i.e. the development of advanced reactor

  12. Plutonium breeding in liquid-metal fast breeder reactors and light water reactors

    International Nuclear Information System (INIS)

    The possibilities of breeding in liquid-metal fast breeder reactors (LMFBRs) and light water reactors (LWRs) are compared in two ways. The feasibility of breeding has been demonstrated in the Phenix reactor with a measured gain of 0.14. The gain in Superphenix will amount to about0.20. The studies show that while maintaining the performance of commercial reactors their breeding gain can be further increased either by the concept of heterogeneous cores or by using carbide or nitride fuel (breeding gain about0.35). Recently, the old idea of breeding in advanced pressurized water reactors (PWRs) has been taken up again with the objective of attaining a gain of 0.05. Unfortunately, these objectives had to be limited to a conversion ratio of 0.9 for safety reasons, and it is not certain whether operation will be rewarding economically. The strategy of substituting PWRs is examined using the French example. By gradually introducing LMFBRs, the cumulated uranium supplies in France can be kept within reasonable limits, which means that they attain three to four times the home resources. This is not possible with advanced LWRs, which can be considered only as a possible backup solution for plutonium recycling into PWRs

  13. Development of hydride absorber for fast reactor. Application of hafnium hydride to control rod of large fast reactor

    International Nuclear Information System (INIS)

    The application of hafnium hydride (Hf-hydride) to a control rod for a large fast reactor where the B4C control rod is originally employed is studied. Three types of Hf-hydride control rods are designed. The control rod worth and its change during the burnup are evaluated for different hydrogen-to-hafnium ratios and are compared with those of the original B4C control rod. The result indicates that the worths of the Hf-hydride and the 10B-enriched B4C control rods are approximately the same, and the lifetime of the Hf-hydride control rod is almost four times longer than that of the 10B-enriched B4C control rod. The core performances of the shutdown margin, sodium void reactivity, Doppler reactivity coefficient, and breeding ratio are analyzed. It is indicated that those for the Hf-hydride control rod are almost the same as those for the original B4C control rod. The behavior of neutrons moderated by the Hf-hydride control rod is analyzed. It is confirmed that the Hf-hydride control rod does not cause any thermal spike problems in the fast reactor core. (author)

  14. Irradiation Experiments on Plutonium Fuels for Fast Reactors

    International Nuclear Information System (INIS)

    An assessment carried out some years ago indicated that cermet fuels might provide the high burn-up and integrity required for fast reactors. An irradiation programme was started at Harwell on (U, Pu)O2 -SS cermet plates and rods, mainly In thermal neutron fluxes, to gain experience of dimensional stability at temperatures typical of modern sodium-cooled fast reactor designs (600-650°C). A subsequent assessment showed that cermets carried a large penalty as far as breeding was concerned and (U, Pu)C was chosen by Harwell for long-term study as an alternative, economic, fast reactor fuel. However, the results from the cermet experiments were of sufficient promise to proceed with parallel irradiation programmes on cermets and carbide. The studies of cermets showed that dimensional instability (swelling and cladding rupture) were caused by the pressures exerted on the steel matrix by the fuel particles, and that the initial density of the fuel particles was important in determining the burn-up at which failure occurred. Further, it was shown that cermets provided a useful vehicle for studying the changes occurring in oxide fuel particles with increasing burn-up. The disappearance of initial porosity and its replacement by fission gas bubbles and segregated solid fission products was studied in some detaiL No significant differences were observed between UO2 and(U,Pu)O2 particles. The initial studies of (U, Pu)C were concerned with the effect of varying composition and structure on swelling and fission gas release. A tantalum-lined nickel alloy cladding material was used to contain both pellet and powder specimens In an irradiation experiment in the core of the Dounreay fast reactor. This showed that the presence of a metal phase in the fuel led to a high swelling rate, that fission gas release was low up to ∼ 3% bum-up, and that a low density powder accommodated the swelling without excessive straining of the can. A subsequent experiment was conducted in a thermal

  15. Status of fast reactor and pyroprocess technology development in Korea

    International Nuclear Information System (INIS)

    For the time being, PWRs will remain as the major source of nuclear power in Korea. However, the storage of the spent fuels produced from those PWRs is a big issue. The on-site spent fuel storage capacity will reach its limit by 2016. Therefore, a decision-making process for spent fuel management is under way. It has been recognized that one of the most promising nuclear options for electricity generation is a fast reactor system which efficiently utilizes uranium resources and reduces radioactive wastes from nuclear power plants, thus contributing to sustainable development. In response to this recognition, the widespread concern about the management of spent fuels caused us to develop technologies for Sodium-cooled Fast Reactors (SFRs) as one of the most promising future types of reactors in Korea. The pyroprocessing technology capitalizes on the recovery of actinide elements from spent fuel for the recycling and fissioning in SFRs for the purpose of power generation. The overriding goal of this R and D plan for pyroprocessing technology combined with SFRs is to develop a closed nuclear fuel cycle that is economically viable, resistant to diversion of nuclear materials, and minimizes generation of waste products, thereby efficiently increasing the capacity of a final spent fuel repository by approximately 100 times. In this fuel cycle, plutonium remains with other isotopes and impurities throughout the processes and cannot be chemically separated in pure form, which reduces the risk of nuclear proliferation. Confining the final product in a hot cell also makes it far less open to misuse. In order to provide a consistent direction to long-term R and D activities the Korea Atomic Energy Commission (KAEC) approved a long-term development plan for future nuclear reactor systems which include SFR, pyroprocess and VHTR on December 22, 2008. This long-term plan will be implemented through nuclear R and D programs of the National Research Foundation, with funds from the

  16. Fast nuclear reactors. Associated international projects. State of the art and assessment of the concepts

    International Nuclear Information System (INIS)

    The recognition of the strategic importance of nuclear energy as a source of sustainable energy may be perceived in the continuous development, in many countries, of the technology of fast nuclear reactors with an associated closed fuel cycle, assuming that these Generation IV innovative systems will be required in the future. These reactors fulfill international requirements for safety and reliability, economic competitiveness, sustainability and proliferation resistance. They have the potential of using more efficiently the natural resources of Uranium and of reducing the volume and radiotoxicity of the nuclear waste by partitioning and transmutation of Minor Actinides. The national and international programs being carried out today are concentrated in the following concepts: Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), Gas Fast Reactor (GFR), Super Critical Water Reactor (SCWR) and Molten Salt Reactor (MSR). This article presents a short review of the technology of the mentioned concepts and details the current state of the main national and international related projects. (author)

  17. Selection of sodium coolant for fast reactors in the US, France and Japan

    International Nuclear Information System (INIS)

    Highlights: ► Trilateral study was conducted on coolant selection of fast reactor concept. ► Fast reactor concepts are vital for nuclear fuel cycle sustainability goals. ► Sodium, gas and lead cooled fast reactors are capable to achieve the goals. ► Sodium cooled fast reactor is the most matured technology. ► Gas and lead cooled fast reactor require long term development. - Abstract: The joint paper presents a common view of fast reactor specific missions in the development of nuclear energy and a cross-analysis of merits and demerits of several Fast Reactors concepts studied worldwide and especially in the Generation-IV International Forum (GIF) framework. The paper provides the context for fast reactors development in the United States, France and Japan and focuses on the comparison on Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), and Lead-cooled Fast Reactor (LFR), i.e. the three fast reactor concepts that have the potential to meet the nuclear fuel cycle sustainability goals. The information provided in the article permits the reader to understand each country's objectives to see that not only the objectives searched for but also the technical orientations are converging. The authors underline that SFR technology evaluation relies significantly on the substantial base technology development programs within each country which is without comparison for the other two fast reactor technologies, e.g., SFR technology has already been developed to commercial or near commercial scale in each country whereas the performance of LFR and GFR technology is still uncertain. The main GFR merits are the potential for high temperatures and the easier possibilities for inspections and repairs. The main challenges are the fuel (fabrication, in-pile behavior), materials for high temperatures, and the implementation of mitigation means to manage severe core degradation. The main LFR merit is the lack of chemical reactivity of the lead coolant with air

  18. Degrading the Plutonium Produced in Fast Breeder Reactor Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jor-Shan; Kuno, Yusuke [Tokyo University, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan)

    2009-06-15

    Plutonium quality, defined as the plutonium isotopic composition, is an important measure for proliferation-resistance (PR) of a nuclear energy system. The quality of the plutonium produced in the blanket assemblies of a fast breeder reactor could be as good as or better than the weapons-grade (WG). The presence of such good quality plutonium is a proliferation concern. There are various options to degrade the plutonium produced in the breeder blanket. The obvious one is to blend the blanket plutonium with those produced from the reactor core during reprocessing. Other options try to prevent the generation of good quality plutonium (Pu). The Protected Plutonium Production (P{sup 3}) Project proposed by Tokyo Institute of Technology (TIT)1,2,3 advocates the doping of certain amount of neptunium (Np), or americium (Am) in fresh blanket fuel for irradiation. The increased production of {sup 238}Pu, {sup 240}Pu and {sup 242}Pu by neutron capture in {sup 237}Np and Am would degrade the blanket plutonium. However, as {sup 237}Np is a controlled material according to IAEA, its use as doping material in fresh blanket fuel presents a concern for nuclear proliferation. In addition, the fabrication of fresh blanket fuel with inclusion of americium would be complicated due to the emission of intense low-energy gamma radiation from {sup 241}Am. Am is normally accompanied by Cm since the separation of those 2 elements is very difficult. Fuel containing both Am and Cm may make Safeguards measurement difficult. A variation would be doping the fresh blanket fuel with minor actinide (e.g., a group of neptunium, americium, and curium), or with separated reactor-grade (RG) plutonium. The drawback of such schemes would be the need for glove boxes in fresh blanket fuel fabrication. It is possible to fuel the breeder blankets with recycled (reprocessed) uranium oxide. The recycled uranium, recovered from reprocessing, contains {sup 236}U, which when irradiated in the blanket would

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

    Science.gov (United States)

    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.

  20. Designing a SCADA system simulator for fast breeder reactor

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

    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.

  2. Fast reactors and surplus weapon`s grade plutonium utilization

    Energy Technology Data Exchange (ETDEWEB)

    Antipov, S.A.; Astafiev, V.A.; Bibilashvily, Yu.K.; Reshetnikov, F.G. [All-Russia Research Inst. of Inorganic Materials, Moscow (Russian Federation)

    1997-12-31

    Liberated large quantities of surplus weapon`s grade Pu have lead to a more acute problem relevant to Pu utilization and storage. The Russian scientists acknowledge the use of Pu in fast reactors to be most efficient. In this case a new composition of fuel is to be designed that would much reduce or eliminate Pu breeding. With this aim in view two kinds of fuel are under study, namely, higher Pu content (up to 45-50%) fuel and the one containing an inert diluent. The study and optimization of their production processes are in progress. In the nearest future experimental fuel rods fueled with both the fuel types will be loaded to be in-pile tested. (author)

  3. In-reactor experiments in fast breeder test reactor for assessment of core structural materials

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India is a sodium cooled reactor with neutron flux level of the order of 1015 n/cm2/s and temperature of coolant in the range of 650-790K (380-520oC). This reactor is being used as a test bed for the development of fuel and structural materials required for Indian Fast Reactor Programme. FBTR is also used as a test facility to carry out accelerated irradiation tests on thermal reactor structural materials. In-reactor experiments on core structural materials are being carried out by subjecting prefabricated specimens to desired conditions of temperature and neutron fluence levels in FBTR. Non-instrumented irradiation capsules that can be loaded at any location of FBTR core are used for the experiments. Pressurised capsules of zirconium alloys have been developed and subjected to irradiation in FBTR to determine the irradiation creep rate of indigenously developed zirconium alloys (Zircaloy-2 and Zr-2.5%Nb alloy) for life assessment of pressure tubes of Indian Pressurised Heavy Water Reactors (PHWRs). Technology development of pressurised capsules was carried out at IGCAR. These pressurised capsules were filled with argon and a small fraction of helium at a high pressure (5.0-6.5 MPa at room temperature) in such a way that the target stresses were attained in the walls of the pressurised capsules at the desired temperature of irradiation in the reactor. FBTR was operated at a low power of 8 MWt during this irradiation campaign to have an inlet temperature of about 579 K (306oC) which was close to the temperature of pressure tubes at full power in PHWR. Irradiation of thirty pressurised capsules was carried out in FBTR using six irradiation capsules for different durations (upto 79 days). The fluence levels attained by the pressurised capsules were up to 1.1 x 1021 n/cm2 (E> 1 MeV) at temperatures of 579 to 592 K. Post-irradiation increase in diameter of the pressurised

  4. Feasibility study for fast reactor and related fuel cycle. Preliminary studies in 1998

    International Nuclear Information System (INIS)

    Prior to the feasibility study for fast reactors (FRs) starting from the 1999 fiscal year, planned in the medium and long-term program of JNC, preliminarily studies were performed on 'FR systems except sodium cooled MOX fueled reactors'. Small scale or module type reactors, heavy metal (Pb or Pb-Bi) cooled reactors, gas cooled reactors, light water cooled reactors, and molten salt reactors were studied on the basis of literature. They were evaluated from the viewpoint of the technical possibility (the structure integrity, earthquake resistance, safety, productivity, operability, maintenance repair, difficulty of the development), the long-term targets (market competitiveness as an energy system, utilization of uranium resources, reduction of radioactive waste, security of the non-proliferation), and developmental risk. As the result, the following concepts should be studied for future commercialized FRs. Small scale and module type reactor: Middle-sized reactor with an excellent economical efficiency. Small power reactor with a multipurpose design concept. Gas cooled reactor: CO2 gas cooled reactor, He gas cooled reactor. Heavy metal cooled reactor: Russian type lead cooled reactor. Light water cooled reactor: Light water cooled high converter reactor and super critical pressure light water cooled reactor. Molten salt reactor: Trichloride molten salt reactor which matches the U-Pu cycle. (author)

  5. Unsteady thermal analysis of gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    This thesis presents numerical analysis of transient heat transfer in an equivalent coolant-fuel rod cell of a typical gas cooled, fast nuclear reactor core. The transient performance is assumed to follow a complete sudden loss of coolant starting from steady state operation. Steady state conditions are obtained from solving a conduction problem in the fuel rod and a parabolic turbutent convection problem in the coolant section. The coupling between the two problems is accomplished by ensuring continuity of the thermal conditions at the interface between the fuel rod and the coolant. to model turbulence, the mixing tenght theory is used. Various fuel rod configurations have been tested for optimal transient performance. Actually, the loss of coolant accident occurs gradually at an exponential rate. Moreover, a time delay before shutting down the reactor by insertion of control rods usually exists. It is required to minimize maximum steady state cladding temperature so that the time required to reach its limiting value during transient state is maximum. This will prevent the escape of radioactive gases that endanger the environment and the public. However, the case considered here is a limiting case representing what could actually happen in the worst probable accident. So, the resutls in this thesis are very indicative regarding selection of the fuel rode configuration for better transient performance in case of accidents in which complete loss of collant occurs instantaneously

  6. Flow induced vibrations in liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Flow induced vibrations are well known phenomena in industry. Engineers have to estimate their destructive effects on structures. In the nuclear industry, flow induced vibrations are assessed early in the design process, and the results are incorporated in the design procedures. In many cases, model testing is used to supplement the design process to ensure that detrimental behaviour due to flow induced vibrations will not occur in the component in question. While these procedures attempt to minimize the probability of adverse performance of the various components, there is a problem in the extrapolation of analytical design techniques and/or model testing to actual plant operation. Therefore, sodium tests or vibrational measurements of components in the reactor system are used to provide additional assurance. This report is a general survey of experimental and calculational methods in this area of structural mechanics. The report is addressed to specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors. 92 refs, 90 figs, 8 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France

    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.

  8. Simulation of hydrocarbons pyrolysis in a fast-mixing reactor

    Institute of Scientific and Technical Information of China (English)

    MG Ktalkherman; IG Namyatov

    2015-01-01

    Currently, thermal decomposition of hydrocarbons for the production of basic petrochemicals (ethylene, propyl-ene) is carried out in steam-cracking processes. Aside from the conventional method, under consideration are alternative ways purposed for process intensification. In the context of these activities, the method of high-temperature pyrolysis of hydrocarbons in a heat-carrier flow is studied, which differs from previous ones and is based on the ability of an ultra-short time of feedstock/heat-carrier mixing. This enables to study the pyrolysis process at high temperature (up to 1500 K) at the reactor inlet. A set of model experiments is conducted on the lab scale facility. Liquefied petroleum gas (LPG) and naphtha are used as a feedstock. The detailed data are obtain-ed on temperature and product distributions within a wide range of the residence time. A theoretical model based on the detailed kinetics of the process is developed, too. The effect of governing parameters on the pyrolysis process is analyzed by the results of the simulation and experiments. In particular, the optimal temperature is detected which corresponds to the maximum ethylene yield. Product yields in our experiments are compared with the similar ones in the conventional pyrolysis method. In both cases (LPG and naphtha), ethylene selectivity in the fast-mixing reactor is substantial y higher than in current technology.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Manufacturing of prototype fast breeder reactor components: challenges and achievements

    International Nuclear Information System (INIS)

    In the presentation, three components of 500 MWe Prototype Fast Breeder Reactor (PFBR), viz. grid plate, roof slab and fuel handling systems, are focused, which have been responsible for the considerable delay of the project schedule. The manufacturing challenges of grid plate mainly originated from large number of sleeves resulting in higher self weight and hard facing of large diameter sleeves. Machining of large diameter plates and shell assembly to the required tight tolerances on dimensions, hard facing with nickel based cobalt free hard facing material on continuous, large diameter (6.7 m) annular tracks, heat treatment of large austenitic stainless steel parts at 1050℃ with controlled rates of cooling and heating together with control on temperature gradient across the parts, complex assembly of a large number of parts (∼14900) meeting the important requirements on verticality of sleeve assemblies (Ø0.1 mm) and delicate handling and transportation are truly challenging activities in the manufacturing technology. In case of roof slab, complex manufacturing process, especially welding between the shell and stiffeners caused lamellar tearing problems and extensive testing time. Inclined fuel transfer machine, multiple repairs, heavy weight and testing strategy resulted in long manufacturing and testing time. Some general lessons learnt are also brought out in this presentation. Technology development prior to start of construction is essential for long delivery components. Judicious choice of tolerances, number and location of welds and inspections has to be made. Robust criteria need to be applied for the acceptance of manufacturing deviations and material compositions. Indigenous materials should be used after qualifications of manufacturing process of direct relevance apart from routine standards. From the rich experience gained through the manufacture and erection of reactor assembly components of PFBR, important guidelines and approaches were derived

  11. Last twenty years experiences with fast reactors in Japan

    International Nuclear Information System (INIS)

    core. Joyo conducted pin-scaled irradiation tests of minor actinide (MA) bearing fuel. However, Joyo has been shut down since May 2007 after an obstacle at a part of fuel assembly discharging equipment was found, and it needs repairing. To demonstrate that loop-type SFR can be used as a power reactor with the experience of Joyo applied, PNC started construction 'Monju' in October 1985 and it attained first criticality in April 1994. However, a sodium-leak incident occurred in the secondary coolant system on December 8, 1995 and its operation has shut down since then. Presently, entire system function test is in progress for resumption of operation through an overhaul and reconstruction. The original mission of Monju is to demonstrate the reliability of FBR power plant through its operation and establish sodium handling technologies. Monju is now positioned as a core function in 'the energy R and D centralization plan' that Fukui prefecture is promoting, and will be used as one of the main fields for R and D toward FR commercialization. (1) Demonstration FR designed by Utilities. As for the design of demonstration FR in the next stage after Monju, Japan Atomic Power Co. (JAPC) became the center and promoted the design studies and innovative elemental technology development from 1985 to 1999. Developing the loop-type reactor technology to rationalize the design, top-entry loop-type SFR (approx. 660 MWe) was selected, and a construction cost was estimated around 1.3 times as much as that of a light-water reactor (equivalent to 1000MWe power plant). (2) JSFR (Japanese Sodium-cooled FR). In 1999, JAEA launched the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (FS)' with domestic partners, such as utilities, vendors,and universities. The study evaluated advanced concept of FBR cycle systems with various coolants such as sodium, heavy metal, gases and water, and suitable fuel types such as oxide, nitride and metal,with viewpoints of safety, economics

  12. Plutonium bearing oxide fuels for recycling in thermal reactors and fast breeder reactors

    International Nuclear Information System (INIS)

    Programs carried out in the past two decades have established the technical feasibility of using plutonium as a fuel material in both water-cooled power reactors and sodium-cooled fast breeder reactors. The problem facing the technical community is basically one of demonstrating plutonium fuel recycle under strict conditions of public safety, accountability, personnel exposure, waste management, transportation and diversion or theft which are still evolving. In this paper only technical and economic aspects of high volume production and the demonstration program required are discussed. This paper discusses the role of mixed oxide fuels in light water reactors and the objectives of the LMFBR required for continual growth of nuclear power during the next century. The results of studies showing the impact of using plutonium on uranium requirements, power costs, and the market share of nuclear power are presented. The influence of doubling time and the introduction date of LMFBRs on the benefits to be derived by its commercial use are discussed. Advanced fuel development programs scoped to meet future commerical LMFBR fuel requirements are described. Programs designed to provide the basic technology required for using plutonium fuels in a manner which will satisfy all requirements for public acceptance are described. Included are the high exposure plutonium fabrication development program centered around the High Performance Fuels Laboratory being built at the Hanford Engineering Development Laboratory and the program to confirm the technology required for the production of mixed oxide fuels for light water reactors which is being coordinated by Savannah River Laboratories

  13. Analysis of deficiencies in fast reactor blanket physics predictions

    International Nuclear Information System (INIS)

    This analysis addresses a deviation between experimental measurements and fast reactor blanket physics predictions. A review of worldwide results reveals that reaction rates in the blanket are underpredicted with the discrepancy increasing with penetration into the blanket. The analysis of this discrepancy involves two parts: quantifying possible error reductions using the most advanced methods and investigating deficiencies in current methodology. The source of these discrepancies was investigated by application of ''state-of-the-art'' group constant generation and flux prediction methodology to flux calculations for the Purdue University Fast Breeder Blanket Facility (FBBF). Refined group constant generation methods yielded a significant reduction in the blanket deviations; however, only about half of the discrepancy can be accounted for in this manner. Transport theory calculations were used to predict the blanket neutron transmission problem. The surprising result is that transport theory predictions utilizing diffusion theory group constants did not improve the blanket results. Transport theory predictions exhibited blanket underpredictions similar to the diffusion theory results. The residual blanket discrepancies not explained using advanced methods require a refinement of the theory. For this purpose an analysis of deficiencies in current methodology was performed

  14. Fast reactor development program in France in 1997

    International Nuclear Information System (INIS)

    The operating performance of the French nuclear plants in 1997 were again satisfactory, confirming the good 1996 results; 78% of the national electricity production was of nuclear origin (376/481 TWh); the nuclear safety indicators were even better with a slight decrease in the number of reported incidents; the availability factor was higher than 82,5%; the kWh production cost continue to decrease by about 1,7% per year; thanks to an ALARA approach, individual and collective doses are still decreasing; the liquid and gaseous releases are also still decreasing and the volume of nuclear wastes is equivalent to last year. The major event of 1997 was the confirmation for the start up of the most advanced nuclear plants, N4 type, with Chooz B1 reaching full power on May 9, 1997 and Chooz B2 following soon with first criticality on March 10 and full power on September 18. As a confirmation and enjoying the experience of Chooz B the first unit of Civaux was synchronised to the grid end of 1997 after fuel loading in Sept 1997. The status of fast reactors was deeply modified, with the governmental decision to abandon Superphenix, which led to decide to close also EFR studies in 1997, and the perspective of operating Phenix till 2004, mainly for studies on radioactive wastes incineration. Three meetings were held during the second half of the year by the > (GP - Permanent Committee) on Phenix; after reviewing all the safety aspects, they concluded to the possibility to resume operation according to the general following scheme: one cycle in 1998, about one year for works, then 6 cycles between 2000 and 2004. Of course, the main event concerning Superphenix is the decision of abandon announced in June 1997. The decree allowing operation had been cancelled in February. The works foreseen in the frame of the planned > shutdown had been conducted satisfactorily, during the first-half of the year. At the beginning of 1998, it was stated that the reactor would not operate again

  15. A review of the United Kingdom fast reactor programme - March 1984

    International Nuclear Information System (INIS)

    The UK programme in the field of fast reactors has continued successfully towards the following main objectives, details of which are contained in subsequent sections of this report: (a) Full power operation of the PFR. (b) Development work supporting the NNC designed CDFR. (c) Demonstration and development of the fast reactor fuel cycle

  16. European Fast Reactor IWGFR/FRCC-report. A review of the collaborative programme on the European Fast Reactor (EFR)

    International Nuclear Information System (INIS)

    The design work for the 1500 MWe European Fast Reactor EFR was started in 1988. Two years during phase 1 were devoted to the concept design; the subsequent concept validation phase 2 will last until March 1993. In autumn 1991 the 'concept design '91, CD91, was put forward; its major design features and the R and D support are described briefly together with the organisational structures. The European Fast Reactor Utilities Group 'EFRUG' presently comprises EdF (France), ENEL (Italy), Nuclear Electric (UK) and Bayernwerk, PreulsenElektra and RWE (Germany). For design and construction of EFR the group 'EFR Associates' is responsible, combining the companies Siemens (formerly Interatom, Germany), NNC Ltd. (UK) and Framatome/Div. Novatome (France). The necessary R and D support is given by CEA (France), UKAEA (UK) and KfK/Siemens (Germany). The R and D work is executed in the various national research centres ranging from Dounreay via Bensberg and Karlsruhe to Cadarache. The design work is done at Bensberg, Lyon and Risley. The present programme of design work extends to early 1993 and is aimed at producing a detailed consistent design for the nuclear part of the plant and a non site specific safety report. By that date the basic feasibility of the main design features will have been underwritten by the joint R and D programme and there will be an informal assessment of the general licensibility of the concept by the Ad Hoc Safety Club. In follow-up the utilities will then be in the position to decide whether to proceed with the next steps. The key issues of this phase will include the specification of the plant, the siting, the detailed engineering, licensing with possibly a public enquiry and the question of ownership and financing. In the international arena the collaboration with USSR is proceeding well on the basis of an USSR-Europe Agreement from January 8, 1991; it foresees review and specialists meetings in the field of fast breeder research. On the occasion of

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

  19. International Working Group on Fast Reactors Second Annual Meeting. Summary Report

    International Nuclear Information System (INIS)

    The Agenda of the Meeting was as follows: Opening of the meeting. 2. Appraisal of the IWGFB's activity for the period from the first annual meeting of the Group. 3. Comments on national programmes on fast breeder reactors. 4. Presentation of general findings and conclusions of national and regional meetings on fast reactor problems held in represented countries and international organisations last year. 5. Comments on the programmes of international meetings on fast reactors to be held in 1969. 6. Consideration of a schedule for meetings on fast reactors in 1970. 7. Suggestions for the topics and location of specialists' meetings in 1969-1970. 8. Suggestions for reviews and studies in the field of fast reactors. 9. The time and place of the third annual meeting of the IWGFR. 10. Closing of the meeting

  20. Status review of methods for the calculation of fast neutron nuclear data for structural materials of fast and fusion reactors

    International Nuclear Information System (INIS)

    The report contains the texts of the 9 invited papers delivered during the Second Research Co-ordination Meeting on ''Methods for the Calculation of Fast Neutron Nuclear Data for Structural Materials and Fast and Fusion Reactors'' held in Vienna during 15-17 February 1988. A separate abstract was prepared for each of these 9 papers. Refs, figs and tabs

  1. Operation and maintenance experience with control rod and their drive mechanisms of fast breeder test reactor

    International Nuclear Information System (INIS)

    This paper explains the functional and construction features of Control Rod Drive Mechanism (CRDM) and control rod used in Fast Breeder Test Reactor (FBTR) which is a 40 MWt loop type sodium cooled fast reactor. It discusses all safety related incidents and failures encountered during its service in reactor, the solutions evolved and modifications carried out to prevent recurrence. It also details the maintenance activities and periodical surveillance carried out. The results of a reliability analysis done are also discussed. (author)

  2. Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor

    International Nuclear Information System (INIS)

    Determination of thermal to fast neutron flux ratio (ffast) and fast neutron flux (φfast) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The ffast and subsequently φfast were determined using the absolute method. The ffast ranged from 48 to 155, and the φfast was found in the range 1.03x1010-4.89x1010 n cm-2 s-1. These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

  3. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  4. Progress in liquid metal fast reactor technology. Proceedings of the 28th meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    The key objectives and activities of Member State liquid metal fast reactor (LMFR) programmes are: Demonstration of effective designs; demonstration of system safety; demonstration of economic competitiveness with other power generation systems. The International Working Group on Fast Reactors (IWGFR) at its 1995 meeting observed that while some countries (as a result of static or falling power demand) are reducing the research and development programmes or delaying the commercial deployment of fast reactors, other countries are planning to introduce these reactors and are embarking on their own development programmes. In these circumstances the international exchange of information and experience is of increasing importance. These proceedings contain updated information from long standing members of the IWGFR and new information on the status of LMFR research and development from new members of the Group: Brazil, China, Republic of Kazakhstan and the Republic of Korea. Refs, figs, tabs

  5. Preliminary Study for Conceptual Design of Advanced Long Life Small Modular Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, T. K. [Argonne National Laboratory, Argonne (United States)

    2015-05-15

    As one of the non-water coolant Small-Modular Reactor (SMR) core concepts for use in the mid- to long-term, ANL has proposed a 100 MWe Advanced sodium-cooled Fast Reactor core concept (AFR-100) targeting a small grid, transportable from pre-licensed factories to the remote plant site for affordable supply. Various breed-and-burn core concepts have been proposed to extend the reactor cycle length, which includes CANDLE with a cigar-type depletion strategy, TerraPower reactors with fuel shuffling for effective breeding, et al. UNIST has also proposed an ultra-long cycle fast reactor (UCFR) core concept having the power rating of 1000 MWe. By adopting the breed-and-burn strategies, the UCFR core can maintain criticality for a targeting reactor lifetime of 60 years without refueling. The objective of this project is to develop an advanced long-life SMR core concept by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. A conceptual design of long life small modular fast reactor is under development by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. The feasibility of the long-life fast reactor concepts was reviewed to obtain the core design guidelines and the reactor design requirements of long life small modular fast reactor were proposed in this study.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma (Computational Engineering Analysis, Albuquerque, NM); Al Rashdan, Ahmad (Texas A& M University, College Station, TX); Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX); 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.

  7. Defect assessment procedure: A french approach for fast breeder reactors

    International Nuclear Information System (INIS)

    As a result of a collaborative effort between Commissariat a l'Energie Atomique, Electricite de France, and NOVATOME to produce and improve rules for fast breeder reactors, RCC-MR, an interim defect assessment procedure is now available in the first draft version (appendix A16). This procedure addresses defects detected during in-service inspection for reactor components operating at moderate or high temperature conditions. Three stages have been considered: initiation, propagation under cyclic loading with or without holdtime and crack instability by ductile and creep rupture. For each of these topics, procedures and rules based on fracture mechanics are proposed. Prediction of initiation is obtained by a simplified method named σd method which relies on the evaluation of the real stress-strain history on a small distance d (d = 0.05 mm for 316L(N) austenitic steel) close to the crack front and material characteristics (limiting stresses) that are available in nuclear codes. This method has been developed for fatigue, creep and creep-fatigue conditions. Defect growth assessment is performed for fatigue and creep-fatigue conditions. For creep-fatigue conditions, fatigue and creep crack growth per cycle are calculated separately and the total crack extension is taken as the sum of the two contributions. Extensive use of simplified method for estimating J (Js method) is made and developed when mechanical and thermal loadings are specified. On the final defect size, assessment may be made in order to avoid crack instability by ductile and creep rupture and collapse load on the remaining. The organization and contents of the present version of this appendix A16 is described. An overview of each specific rule is given

  8. Preapplication safety evaluation report for the Sodium Advanced Fast Reactor (SAFR) liquid-metal reactor

    International Nuclear Information System (INIS)

    This safety evaluation report (SER) presents the final results of a preapplication design review for the Sodium Advanced Fast Reactor (SAFR) liquid metal reactor (Project 673). The SAFR conceptual design was submitted by the US Department of Energy (DOE) in accordance with the US Nuclear Regulatory Commission (NRC) ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 FR 24643 which provides for the early Commission review and interaction). The standard SAFR plant design consists of four identical reactor modules, referred to as ''paks,'' each with a thermal output rating of 900 MWt, coupled with four steam turbine-generator sets. The total electrical output was held to be 1400 MWe. This SER represents the NRC staff's preliminary technical evaluation of the safety features in the SAFR design. It must be recognized that final conclusions in all matters discussed in this SER require approval by the Commission. During the NRC staff review of the SAFR conceptual design, DOE terminated work on this design in September 1988. This SER documents the work done to that date and no additional work is planned for the SAFR

  9. Development of high nitrogen electrodes for fast breeder reactor applications

    International Nuclear Information System (INIS)

    Austenitic stainless steels of AISI type 316 (316 SS) and its variants are used extensively as structural material for the components of fast reactors operating at temperature up to 823 K. SS 316LN has been chosen as the major structural material for the construction of Prototype Fast Breeder Reactor (PFBR) with a targeted service life of 40 years. To reduce the risk of sensitization in SS 316LN, the carbon content has been reduced to less than 0.03 wt%, and the nitrogen content has been specified as 0.08 wt% to compensate the loss in strength due to the reduced carbon content. An improved version of this alloy with nitrogen content of 0.14 wt% in a frilly austenite matrix has been developed for the future FBRs, to enhance the service life of the structural components up to 60 years. Indigenously developed modified E3 16-1 5 electrodes were used for the fabrication of PFBR components to enhance the structural reliability of the components. The modifications from AWS/ASME SFA 5.4 include stringent composition limits, narrow range of ferrite content, and impact toughness after aging at 1023K for 100h, tensile properties at elevated (service) temperatures and intergranular corrosion (IGC) test as per ASTM A262 Practice E. Since the improved version alloy is rich in nitrogen content than the existing alloy, it has become necessary to develop a welding consumable with reasonably good weldability that is suitable for the fabrication of future FBR components. At present there are no commercially available welding consumables to weld these steels and the development is under way. In this work, a matching consumable methodology was adopted to develop the welding consumable. However, as per specification targeting the chemistry, solidification mode and delta ferrite was challenging, since the solidification mode of the weld metal shifts to fully austenitic region due to dilution of nitrogen from the base metal, which may increase the risk of hot cracking susceptibility

  10. Axial distribution of absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    The coupled fast thermal system CFTS at the RB reactor is created for obtaining fast neutron fields. The axial distribution of fast neutron flux density in its second configuration (CFTS-2) is measured. The axial distribution of absorbed doses is computed on the basis of mentioned experimental results. At the end these experimental and computed results are given. (Author)

  11. Commercial US Vendors Focus on Reducing the Cost of Fast Reactors

    International Nuclear Information System (INIS)

    Fast reactor development was originally motivated by the perceived scarcity of uranium and fast reactors were designed to be integrated with fuel cycle reprocessing. Although these are still important considerations, several commercial companies in the United States of America are exploiting other key characteristics of the fast neutron spectrum to design reactors that offer significant capital and energy production cost reductions. Corporate operating philosophies, funding mechanisms, target markets, reactor fuels, coolants and designs from each of these companies are vastly different. Despite this, the companies are each focusing on one or more of the following fast neutron spectrum characteristics: compact designs, inherent safety characteristics, improved efficiencies due to high temperatures, extended core lifetimes and/or high fuel burnup. The challenge they perceive is to design reactors that customers will purchase primarily because they are the most cost effective for their energy needs. (author)

  12. Los Alamos experiments and their impacts on fast reactor safety

    International Nuclear Information System (INIS)

    Results of two sets of recent Los Alamos transition-phase experiments are reported herein. The two sets of experiments addressed two different behaviors of boiling pools of molten fuel, molten steel and steel vapor, in the transition phase of a core-disruptive accident (CDA) in a liquid-metal fast breeder reactor (LMFBR). The transient boilup experiments simulated the recriticality-induced motions of a boiling pool within a single subassembly during the subassembly-pool subphase of the transition phase. The melting wall experiments simulated the melting and entrainment of subassembly duct wall steel into a boiling pool during the same subphase. From the results of the transient boilup experiment we identified behaviors and phenomena that argue against an energetic disassembly from the subassembly-pool subphase. From the melting wall experiments we determined that a stable boiling pool is unlikely by showing that significant amounts of wall steel would likely be rapidly entrained and lead to pool collapse. 8 refs., 3 figs

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

  14. Dissolution of low burnup Fast Flux Test reactor fuel

    International Nuclear Information System (INIS)

    The first Fast-Flux Test Facility reactor fuel [mixed (U,Pu)O2 composition] has been used in dissolution tests for fuel reprocessing. The fuel tested here had a peak burnup of 0.22 at. %, with peak centerline temperatures of 19970C. Linear dissolution rates of 0.99 to 1.57 mm/h were determined for dissolver solution and fresh acid, respectively. Insoluble residues from dissolution at 950C ranged from 0.18 to 0.28% of the original fuel. From 2 to 37 wt % of the residue was recoverable plutonium. Dissolution at 290C yielded residues of 0.56 to 0.64% of the original fuel. The major elements present in the HF leached residue included Ru, Mo, and Rh. The recovered cladding from the 950C dissolution contained the equivalent of 198 mg of 239Pu per 100 g of hulls, while the cladding from the 290c experiments contained only 0.21 mg of 239Pu per 100 g of hulls. 9 references, 5 figures

  15. Development of vanadium fuel cladding for Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Vanadium alloys are promising material for some core components of the Sodium Fast Reactors, especially for fuel cladding applications. With good mechanical properties up to 800°C at least, good behavior under irradiation above 400°C and limited swelling, they also have the benefit from fusion program. In 2010, CEA launched the manufacturing of a V-4Cr-4Ti alloy, well documented in literature, to validate the uneasy fabrication process linked to interstitial element sensitivity and potential pollution in master alloys. 30kg of CEA-J57 alloy (7 mm-plates) were fabricated for the CEA by GfE Metalle und Materialien GmbH, Nuremberg, Germany. The program includes the investigation of recrystallization, resulting microstructure and DBTT values, high temperature mechanical properties such as tensile strength and creep resistance, chemical compatibility with both the oxide fuel and the coolant and assessment of tube fabrication, actually a triplex tube with inner and outer liners to protect vanadium from oxidation during the hot processing. (author)

  16. A Global Assessment of Fast Reactors in the Future

    International Nuclear Information System (INIS)

    Various criteria will be presented and used for assessing the future of sodium cooled fast reactors (SFRs) on a worldwide basis, including sustainability, economics, contribution to maintaining nuclear R&D excellence, long term acceptability of nuclear energy, leading position in nuclear energy industry for countries developing SFRs and diversification of the risks and insurance. One of the main concerns is public acceptance, which may vary over time for a number of reasons. If it is assumed that safety and non-proliferation concerns will be dealt with effectively, acceptance will most probably be obtained and the question will not be whether to launch SFRs on an industrial scale, but when and where. An assessment of the market will also be provided in this paper. The world market for industrial Gen IV SFRs is expected to be between 0 and 2 units (1500 MW(e)) per year based on an optimistic approach, before economic competitiveness is reached, and 10–15 later. Though there are large uncertainties on the exact period at which economic competitiveness will be reached, it is most probably likely to occur sometime during the second half of the century. In the future, the advantages of SFRs will likely grow significantly faster than any disadvantages. (author)

  17. UK fast reactor components. Sodium removal decontamination and requalification

    International Nuclear Information System (INIS)

    Extensive experience gained at the U.K.A.E.A. Dounreay Nuclear Power Development Establishment is being applied to form the basis of the plant to be provided for sodium removal, decontamination, and requalification of components in future commercial fast reactors. In the first part of a three part paper, the factors to be taken into account, showing the UK philosophy and approach to maintenance and repair operations are discussed. In the second part, PFR facilities for sodium removal and decontamination are described and some examples are given of cleaning components such as pumps, charge machine, cold trap baskets, and steam generator units. Similar facilities at DFR are briefly described. In the third part of the paper a short description is given of the Harwell mass transfer loop, currently used to study the deposition of activated stainless steel corrosion products. Decontamination method for pipework specimens cut from the loop are described and results of first screening tests of various chemical decontaminants are presented. (U.K.)

  18. Sodium Fast Reactor Safety and Licensing Research Plan

    International Nuclear Information System (INIS)

    This paper summarizes potential research priorities for the US Department of Energy (DOE) with the intent of improving the licensability of the sodium cooled fast reactor (SFR). In support of this project, five panels were tasked with identifying potential safety related gaps in the available information, data and models needed to support the licensing of an 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 paper to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for 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 access control 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. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

    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.

  20. Method of advancing research and development of fast breeder reactors

    International Nuclear Information System (INIS)

    In the long term plan of atomic energy development and utilization, fast breeder reactors are to be developed as the main of the future nuclear power generation in Japan, and when their development is advanced, it has been decided to positively aim at building up the plutonium utilization system using FBRs superior to the uranium utilization system using LWRs. Also it has been decided that the development of FBRs requires to exert incessant efforts for a considerable long period under the proper cooperation system of government and people, and as for its concrete development, hereafter the deliberation is to be carried out in succession by the expert subcommittee on FBR development projects of the Atomic Energy Commission. The subcommittee was founded in May, 1986, to deliberate on the long term promotion measures for FBR development, the measures for promoting the research and development, the examination of the basic specification of a demonstration FBR, the measures for promoting international cooperation, and other important matters. As the results of investigation, the situation around the development of FBRs, the fundamentals at the time of promoting the research and development, the subjects of the research and development and so on are reported. (Kako, I.)

  1. Hydraulic Experiment for Simulative Assemblies of Blanket Assembly and Np Transmutation Assembly of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    CHENG; Dao-xi; QI; Xiao-guang; ZHAI; Wei-ming; YANG; Bing; ZHOU; Ping

    2013-01-01

    The out-of reactor hydraulic experiment of fast reactor assembly is one of the important experiments in the process of the development of the fast reactor assembly.In this experiment,the size of the throttling element in the foot of the assembly is decided which is fit for the flow division in the reactor and the

  2. Shape optimization of a Sodium Fast Reactor core

    Directory of Open Access Journals (Sweden)

    Dombre Emmanuel

    2013-01-01

    Full Text Available We apply in this paper a geometrical shape optimization method for the design of the core of a SFR (Sodium-cooled Fast Reactor in order to minimize a thermal counter-reaction known as the sodium void effect. In this kind of reactors, by increasing the temperature, the core may become liable to a strong increase of reactivity, a key-parameter governing the chain-reaction at quasi-static states. We first use the one group energy diffusion model and give the generalization to the two groups energy equation. We then give some numerical results in the case of the one group energy equation. Note that the application of our method leads to some designs whose interfaces can be parametrized by very smooth curves which can stand very far from realistic designs. We don’t explain here the method that it would be possible to use for recovering an operational design but there exists several penalization methods (see [2] that could be employed to this end. On applique dans cet article une méthode d’optimisation géométrique dans le cadre de la conception d’un cœur de réacteur SFR (Sodium-cooled Fast Reactor, i.e. réacteur à neutron rapide refroidi au sodium dans le but de minimiser une contre réaction thermique connue sous le nom d’effet de vidange sodium. Lorsqu’une augmentation de température survient, ce type de réacteur peut être sujet à une forte augmentation de réactivité, un paramètre clé dans le contrôle de la réaction en chaîne en régime quasi-statique. On a recours à l’équation de diffusion à un groupe puis on donne la généralisation du modèle d’optimisation pour l’équation de la diffusion à deux groupes d’énergie. On présente ensuite quelques résultats numériques obtenus dans le cas de l’équation à un groupe d’énergie. On note que l’application de cette méthode conduit à des designs de cœur présentant des interfaces très régulières qui sont loin d’un design de cœur faisable sur le

  3. Technological problems in the use of research fast reactors for radiotherapy of patients with malignant tumors

    International Nuclear Information System (INIS)

    The authors discuss the technological problems associated with the use of fast neutrons in radiotherapy of cancer patients and outline the approaches to the solution of these problems. The state of the art is assessed. Physical and radiobiologial prerequisites for the use of fast reactors for radiotherapy of patients with malignant tumors are analyzed. Results of clinic used of BR-10 reactor at the Medical Radiology Research Center, Russian Academy of Medical Sciences, are presented. Experimental and clinical findings indicate that the results of radiotherapy may be appreaciably improved if a novel perspective source of fast neutrons, a nuclear reactor, is used

  4. Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia); Sarmani, S.B. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, 43000 Kajang, Selangor (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia); Radir, M.H. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, 43000 Kajang, Selangor (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia)

    2011-05-15

    Determination of thermal to fast neutron flux ratio (f{sub fast}) and fast neutron flux ({phi}{sub fast}) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f{sub fast} and subsequently {phi}{sub fast} were determined using the absolute method. The f{sub fast} ranged from 48 to 155, and the {phi}{sub fast} was found in the range 1.03x10{sup 10}-4.89x10{sup 10} n cm{sup -2} s{sup -1}. These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

  5. Operational and decommissioning experience with fast reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    For three decades, several countries had large and vigorous fast breeder reactor development programmes. In most cases, fast reactor development programmes were at their peaks by 1980. From that time onward, fast reactor development in general began to decline. The effort essentially disappeared for fast breeder reactor development. Similarly, programmes in other nations were terminated or substantially reduced. In France, Superphenix was shut down at the end of 1998; SNR-300 in Germany was completed but not taken into operation, and KNK-II was permanently shut down in 1991 after 17 years of operation, and is scheduled to be dismantled by 2004; in the UK, PFR was shut down in 1994; BN-350 in Kazakhstan was shut down in 1998. It is difficult to argue that fast breeder reactors will be built in the near term when no commercial market exists and there is a plentiful supply of cheap uranium. Nevertheless, it is reasonable to assume that, were nuclear energy to remain an option as part of the long term world energy supply mix, meeting the sustainability requirements vis-a-vis natural resources and long lived radioactive waste management will require deploying systems involving several reactor types and fuel cycles operating in symbiosis. Apart from cost effectiveness, simplification, and safety considerations, a basic requirement to these reactor types and fuel cycles will be flexibility to accommodate changing objectives and boundary conditions. This flexibility can only be assured with the deployment of the fast neutron spectrum reactor technology, and reprocessing. At the same time that the interest in the fast reactor waned, the retirement of many of the developers of this technology reached its peak, between 1990 and 2000, and hiring diminished in parallel. Moreover, R and D programmes are being discontinued, and facilities falling in disuse. Under these circumstances, the loss of the fast reactor knowledge should be taken seriously. One particularly important

  6. A review of the Italian fast reactor programme

    International Nuclear Information System (INIS)

    In the frame of Italian nuclear program, this report deals with the current activities related to PEC reactor delay in construction and start-up, activities within the joint venture between Novatome, France and NIRA, Italy related to components for Super Phenix reactor, participation of NIRA in the Super Phenix studies covering technology of reactor components, reactor core, fuel, safety, fuel cycle technical and economical aspects, codes and standards

  7. Development of metallic fuels for Indian Fast Breeder Reactors

    International Nuclear Information System (INIS)

    The neutronic performance of metal fuel based on binary U-Pu alloy or ternary U-Pu-Zr alloys are better than conventional uranium plutonium mixed oxide or high density carbide ceramic fuel. The growing energy demand in India needs faster growth of nuclear power and warrants introduction of fast reactors based on metallic fuels in future. Physics calculation showed that fast reactor based on metallic fuels results in higher breeding ratio and lower doubling time compare to mixed oxide or carbide fuels. Moreover inclusion of pyro-processing of the fuel in the fuel cycle is expected to make metal fuel option more economical. As part of metal fuel development programme for future FBR's in India, capsule irradiation of metal fuel based on sodium bonded U-Pu-Zr alloy and metal (Zircaloy) bonded binary U-Pu (Pu ∼ 15 %) alloy are being actively pursued. For this purpose two design concepts have been proposed : one based on sodium bonded ternary alloy fuel of U-Pu-Zr (2-10 wt%) in modified T91 cladding material and the other is U-Pu binary alloy mechanically bonded to modified T91 cladding material with 'Zircaloy' as a liner between the fuel alloy and the clad. The Zircaloy liner act as a barrier in reducing the fuel clad chemical interaction. It also helps in transfer of heat from the fuel to the clad. The smear density of metal bonded pin will be between 70% - 85% and that for sodium bonded pin will be ∼ 70%. In metal bonded fuel pin design two/four semi-circular grooves of diameter ∼1.0 mm, will be provided in diametrically opposite directions in the fuel cross section to accommodate fuel swelling. A comparison has been made on the relative merits and demerits of these two fuel pin designs. The material for the axial blanket will be 'U' or U-Zr (Zr upto 10wt %) alloy based on the results of the out-of-pile thermal cycling behavior and irradiation performance. In the present investigation out-of-pile experiments have been carried out to address some of the issues of

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

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    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

  9. Status of national programmes on fast breeder reactors. Twenty-fifth annual meeting of the International Working Group on Fast Reactors. Summary report. Working material

    International Nuclear Information System (INIS)

    At present nuclear power accounts for approximately 17% of total electricity generation worldwide. Given continuing population growth and the needs of the third world and developing countries to improve their economic performance and standard of living, energy demand is expected to continue to grow through the 21st century. The proportion of energy supplied as electricity is also expected to continue to increase. Although fossil fuelled electricity generation is the option preferred by several countries for the short term, there are rising concerns over climatic consequences caused by extended burning of fossil fuels as a result of the demands of a fast expanding world population. In this situation nuclear electricity will become more and more important and the known reserves of uranium would be consumed quite quickly by thermal reactors. It would be possible to sustain a large nuclear programme only by introducing fast reactors. One can conclude that there are strategic reasons for pursuing the development of fast breeder reactors. It will become desirable essential, to have this technology available for introduction. The experience of the various prototypes presently in operation has confirmed the operability and benign characteristics of the LMFR and has given ground for confidence in the future. Current fast reactor designs offer very large margins of safety and by virtue of redundant and diverse safety systems the potential for an energetic core disruptive accident or for fast reactor core meltdown has been essentially eliminated. Several international forums reviewed the current trends in the fast reactor development. The view was reaffirmed that fast breeder reactors still remain the most practical tool for effective utilization of uranium resources for the future energy needs. Achievement of competitiveness with LMRs is still the first priority condition for the future deployment of this type of reactor. The recycling of plutonium into LMFBRs would allow

  10. Network Representation of Design Knowledge of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    A method of design knowledge representation was studied for the Japanese fast breeder reactor Monju, aiming at enhanced understanding of engineering considerations with mutual relations. Taking over design knowledge of Monju to next generation designers/engineers to be in charge of design of future FRs is by no means easy, in contrast with operation and maintenance knowledge which can be acquired in the real plant operation and maintenance. Specifications of the as-is Monju contains only a small part of the entire design knowledge, mainly by two reasons. Firstly, reasons for selecting the as-is specifications can not be understood until reaching proper knowledge source. Secondly, there are many rejected options on the design specifications. Design specifications are selected along with technical dependencies among a huge number and diversified specification items. Decisions design are made basically along with these dependencies which can hardly be traced in the currently available database or document libraries. Reasons for the rejections of options need to be profoundly understood, because those are not certainly due to technical inferiority. Some of rejected options can be worth reconsidering in the future, possibly by technical advances in materials, high-precision prediction software tools, rationalized standards/code, etc. The authors propose a new design knowledge representation approach based on networking of knowledge nodes along with the mutual dependencies. A prototype software has been developed and a basic performance test was made to visualize the dependency network. An additional function to enable design case studies on hypothetical adoptions of rejected options is now under consideration. (author)

  11. The present status of the fast breeder reactor industrialization in western Europe

    International Nuclear Information System (INIS)

    The development of the liquid metal fast breeder reactor in Europe started in the mid-fifties, after the successful operation of EBR-1 at ARCO, Idaho, in 1951. A more and more integrated development among the countries of the European Community culminated in 1986 with the beginning to power of the 1200 MWe SUPERPHENIX plant at Creys-Malville, France. The road is now open towards the full industrialization of the liquid metal fast breeder reactor at the moment, in 2005, when the first European thermal neutron power reactor station will have to be decommissioned and replaced. The European programme aims at providing the utilities at that time with a clear choice between thermal neutron reactors and fast breeder reactors, both economical but very different in their use of the limited natural resource that uranium is. (author)

  12. Present status of the fast breeder reactor industrialization in western Europe

    International Nuclear Information System (INIS)

    The development of the liquid metal fast breeder reactor in Europe started in the mid-fifties, after the successful operation of EBR-1 at ARCO, Idaho, in 1951. A more and more integrated development among the countries of the European Community culminated in 1986 with the startup of the 1200 MWe SUPERPHENIX plant at Creys-Malville, France. The road is now open towards the full industrialization of the liquid metal fast breeder reactor at the moment, in 2005, when the first European thermal neutron power reactor station will have to be decommissioned and replaced. The European programme aims at providing the utilities at that time with a clear choice between thermal neutron reactors and fast breeder reactors, both economical but very different in their use of the limited natural resources that uranium

  13. A review of the United Kingdom fast reactor programme: March 1987

    International Nuclear Information System (INIS)

    The UK fast reactor programme is reviewed under the following headings: Progress with PFR; Reprocessing; Commercial Design Studies: Structural Integrity; Engineering and Components; Materials; Sodium Chemistry; Core and Fuel; Safety; Plant Performance Studies. (U.K.)

  14. Fast Reactors and Related Fuel Cycles: Challenges and Opportunities (FR09). Proceedings of an International Conference

    International Nuclear Information System (INIS)

    This is the proceedings of an international conference on fast reactors and related fuel cycles convened to exchange experience and innovative ideas in order to achieve progress in this field. Fast reactor programmes are currently on an accelerated growth path in many countries of the world, and the last international fast reactor conference was held almost 20 years ago. The scope of discussion included key scientific and technological areas such as fuels and materials development, safety, advanced simulation, component and system design and coolant technology, in which innovation is pursued to ensure that next generations fast reactor fuel cycles will achieve their potential. The accompanying CD-ROM contains the contributed papers and posters, summaries of 150 oral presentations and the young generation event.

  15. On the major DYN3D developments for fast reactor design and transient analysis

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

    2013-07-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Honekamp, John R. (Downers Grove, IL); Fryer, Richard M. (Idaho Falls, ID)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Honekamp, J.R.; Fryer, R.M.

    1978-03-21

    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.

  18. Start-up dynamics of NPP with a dissociating coolant fast reactor

    International Nuclear Information System (INIS)

    The paper presents the method of computation of the coolant parameters at start-up of a MPP with dissociating coolant fast reactor based on the assumption of the quasi-stationary proceeding of the processes with slow variation of the coolant parameters. The results of the computation variation of the coolant basic parameters in the reactor upon warming-up are given. The mathematical model of the transient processes in the reactor core is given. The influence of the different effects on the reactor dynamics is shown. It is determined that in the range of parameters typical of the warming-up regimes, the reactor possesses good s lf-regulation characteristics

  19. Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Salt Fast Reactor (MSFR)

    International Nuclear Information System (INIS)

    Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor (MSFR) are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated. (authors)

  20. Quasi-static reactivity balance interpretations of inherent safety response in fast and thermal reactors

    International Nuclear Information System (INIS)

    The quasi-static reactivity balance provides a useful way to codify the inherent response of a reactor to unprotected accident initiators. This approach has been used to illuminate the underlying physics of passive reactivity shutdown in liquid-metal-cooled fast reactors (LMRs) and has facilitated the design effort to configure LMR reactor cores for favorable inherent safety features. The purpose of this paper is to extend the quasi-static reactivity balance methodology to thermal reactor types - and in particular to the modular high-temperature gas-cooled reactor (MHTGR) where, as in LMRs, passive reactivity shutdown is a design goal

  1. Neutron cross-section libraries in the AMPX master interface format for thermal and fast reactors

    International Nuclear Information System (INIS)

    Neutron cross-section libraries in the AMPX master interface format have been created for three reactor types. Included are an 84-group library for use with light-water reactors, a 27-group library for use with heavy-water CANDU reactors and a 126-group library for use with liquid metal fast breeder reactors. In general, ENDF/B data were used in the creation of these libraries, and the nuclides included in each library should be sufficient for most neutronic analyses of reactors of that type. Each library has been used successfully in fuel depletion calculations

  2. The Data Required for Fast Reactor Safety Assessment by Probability Methods

    International Nuclear Information System (INIS)

    Probability methods for the assessment of reactor safety have been developed in the UK to the point where it is possible to make some quantitative comparisons of the safety of differing reactor types measured against acceptable criteria for a given site. One particular method for rapid assessment of preliminary designs has been used to show those areas of fast reactor physics and kinetics where further work could materially assist the designer by allowing relaxation of the standards required of reactor shut-down and emergency systems without impairing reactor safety. (author)

  3. Overview of European Community (Activity 3) work on materials properties of fast reactor structural materials

    International Nuclear Information System (INIS)

    The Fast Reactor Coordinating Committee set up in 1974 the Working Group Codes and Standards, and organized its work into four main activities: Manufacturing standards, Structural analysis, Materials and Classification of components. The main purpose of materials activity is to compare and contrast existing national specifications and associated properties relevant to structural materials in fast reactors. Funds are available on a yearly basis for tasks to be carried out through Study Contracts. At present about four Study Contract Reports are prepared each year

  4. Fast reactor core management in Japan: twenty years of evolution at JOYO

    International Nuclear Information System (INIS)

    Twenty years of operations at the experimental fast reactor JOYO provide a wealth of experience with core and fuel management. This experience has been applied to several core modifications to upgrade JOYO's irradiation capability. Core physics tests and Post Irradiation Examination (PIE) results have been used to confirm the accuracy of neutron diffusion theory calculations. These experiences and accumulated data will be useful for the core design in future fast reactors in Japan's development. (author)

  5. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    International Nuclear Information System (INIS)

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour

  6. Commission of the European Communities - Activities in the field of fast reactors

    International Nuclear Information System (INIS)

    The Commission of the European Communities is performing activities in the field of fast reactor on two lines: a) activities aiming to prepare the commercialization of fast reactors by coordination and collaboration between national programmes. b) the execution of an own programme in the Joint Research Centre at Ispra (Italy) and Karlsruhe (Federal Republic of Germany) in the field of FBR safety and research on Pu-bearing fuel

  7. Fast breeder reactor reference system classification for the ENEA data bank

    International Nuclear Information System (INIS)

    This report contains the Reference System Classification (RSC) of fast breeder reactors: it provides a functional system breakdown of the reactor. For each system the following important characteristics are reported: the main function, the mode of operation, its location in the reactor, the main interface system, its main components and the component working environment (fluid and/or atmosphere type). The RSC represent a basic step in organizing the ENEA data bank for the registration and processing of reliability data on typical fast reactor components; it provides a functional component breakdown and represent a plant-unique identification in the process of omogenization of event-data coming from different reactors. In this report it was tried to take into account different generations of nuclear power plants, different plant layouts and solutions: in particular loop and pool reactors are separately treated

  8. Fast pyrolysis in a novel wire-mesh reactor: decomposition of pine wood and model compounds

    NARCIS (Netherlands)

    Hoekstra, E.; Swaaij, van W.P.M.; Kersten, S.R.A.; Hogendoorn, J.A.

    2012-01-01

    In fast pyrolysis, biomass decomposition processes are followed by vapor phase reactions. Experimental results were obtained in a unique wire-mesh reactor using pine wood, KCl impregnated pine wood and several model compounds (cellulose, xylan, lignin, levoglucosan, glucose). The wire-mesh reactor w

  9. Fourteenth annual meeting of the International Working Group on Fast Reactors. Summary report. Part II

    International Nuclear Information System (INIS)

    This report includes description of the state-of-the art in the field of fast reactor technology, research and development, in France, Belgium, India, Italy, USSR, USA, UK, Switzerland, and European Union. The emphasis in the majority of the reports is on the FBR safety issues, sodium cooling system, fuel elements development, reactor materials testing, risk assessment

  10. Count-to-count time interval distribution analysis in a fast reactor

    International Nuclear Information System (INIS)

    The most important kinetic parameters have been measured at the zero power fast reactor CORAL-I by means of the reactor noise analysis in the time domain, using measurements of the count-to-count time intervals. (Author) 69 refs

  11. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all H

  12. On the development of fast breeder reactors and the use of thorium in Brazil

    International Nuclear Information System (INIS)

    This work presents a discussion on the possibility of construction of fast breeder reactors in Brazil. It is specially concerned with the use of thorium which is abundant in our country. The main advantages of this projects are: develop fuel and reactor technology in Brazil, increase thorium research, demonstrate the safety of LMFBR and promote its public acceptance. (A.C.A.S.)

  13. Design and layout decision for refueling system of advanced fast neutron reactors

    International Nuclear Information System (INIS)

    Describes fast neutron reactor refueling features, BN-1200 power unit general data, its refueling system design concepts, individual refueling equipment purpose and designs, and required experimental studies to create it. Refueling equipment characteristics for BN-800 and BN-1200 reactors are compared. (author)

  14. Fabrication of fuel and recycling of minor actinides in fast reactors

    OpenAIRE

    Somers, Joseph

    2010-01-01

    Fuels for future fast reactors will not only produce energy, but they must also actively contribute to the minimisation of long lived wastes produced by these, and other reactor systems. The fuels must incorporate minor actinides (MA = Np, Am, Cm) for neutron transmutation into short lived isotopes. Within Europe oxide fuels are favoured. Transmutation can be considered in homogeneous or heterogeneous reactor recycle modes (i.e. in fuels or targets, respectively). Fabrication of such fuels...

  15. Comparison of multigroup and few-group calculations of fast power reactor parameters

    International Nuclear Information System (INIS)

    The basic parameters of a fast breeder reactor in two-dimensional cylindrical geometry and in multi- and few-group diffusion approximation were calculated and compared. Two different types of reactor were considered, viz., homogeneous and heterogeneous. The results can serve as a quantitative aid for the choice of the proper number of groups for the calculations of various reactor parameters with required accuracy. (author)

  16. Selection of sodium coolant for fast reactors in the US, France and Japan

    Energy Technology Data Exchange (ETDEWEB)

    Sakamoto, Yoshihiko, E-mail: sakamoto.yoshihiko@jaea.go.jp [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan); Garnier, Jean-Claude; Rouault, Jacques [CEA, DEN, DER, Centre de Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Grandy, Christopher; Fanning, Thomas; Hill, Robert [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Chikazawa, Yoshitaka; Kotake, Shoji [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Trilateral study was conducted on coolant selection of fast reactor concept. Black-Right-Pointing-Pointer Fast reactor concepts are vital for nuclear fuel cycle sustainability goals. Black-Right-Pointing-Pointer Sodium, gas and lead cooled fast reactors are capable to achieve the goals. Black-Right-Pointing-Pointer Sodium cooled fast reactor is the most matured technology. Black-Right-Pointing-Pointer Gas and lead cooled fast reactor require long term development. - Abstract: The joint paper presents a common view of fast reactor specific missions in the development of nuclear energy and a cross-analysis of merits and demerits of several Fast Reactors concepts studied worldwide and especially in the Generation-IV International Forum (GIF) framework. The paper provides the context for fast reactors development in the United States, France and Japan and focuses on the comparison on Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), and Lead-cooled Fast Reactor (LFR), i.e. the three fast reactor concepts that have the potential to meet the nuclear fuel cycle sustainability goals. The information provided in the article permits the reader to understand each country's objectives to see that not only the objectives searched for but also the technical orientations are converging. The authors underline that SFR technology evaluation relies significantly on the substantial base technology development programs within each country which is without comparison for the other two fast reactor technologies, e.g., SFR technology has already been developed to commercial or near commercial scale in each country whereas the performance of LFR and GFR technology is still uncertain. The main GFR merits are the potential for high temperatures and the easier possibilities for inspections and repairs. The main challenges are the fuel (fabrication, in-pile behavior), materials for high temperatures, and the implementation of

  17. Technical meeting on 'Operational and decommissioning experience with fast reactors'. Working material

    International Nuclear Information System (INIS)

    For three decades, several countries had large and vigorous fast breeder reactor development programs. In most cases, fast reactor development programs were at their peaks by 1980. Fast test reactors [Rapsodie (France), KNK-II (Germany), FBTR (India), JOYO (Japan), DFR (UK), BR-10, BOR-60 (Russia), EBR-II, Fermi, FFTF (U.S.A.)] were operating in several countries, with commercial size prototype reactors [Phenix, Superphenix (France), SNR-300 (Germany), MONJU (Japan), PFR (UK), BN-350 (Kazakhstan), BN-600 (Russia)] just under construction or coming on line. From that time onward, fast reactor development in general began to decline. By 1994 in the USA, the Clinch River Breeder Reactor (CRBR) had been cancelled, and the two fast reactor test facilities, FFTF and EBR-II had been shutdown - with EBR-II permanently, and FFTF in a standby condition. Thus, effort essentially disappeared for fast breeder reactor development. Similarly, programs in other nations were terminated or substantially reduced. In France, Superphenix was shut down at the end of 1998; SNR-300 in Germany was completed but not taken into operation, and KNK-II was permanently shut down in 1991 after 17 years of operation, and is scheduled to be dismantled by 2004; in the UK, PFR was shut down in 1994; BN-350 in Kazakhstan was shut down in 1998. It is difficult to argue that fast breeder reactors will be built in the near term when no commercial market exists and there is a plentiful supply of cheap uranium. Nevertheless, it is reasonable to assume that, were nuclear energy to remain an option as part of the long-term world energy supply mix, meeting the sustainability requirements vis-a-vis natural resources and long-lived radioactive waste management will require deploying systems involving several reactor types and fuel cycles operating in symbiosis. Apart from cost effectiveness, simplification, and safety considerations, a basic requirement to these reactor types and fuel cycles will be flexibility

  18. Status of liquid metal fast reactor development. Proceedings of the 27. meeting of the International Working Group on Fast Reactors held in Vienna, 17-19 May 1994

    International Nuclear Information System (INIS)

    These proceedings contain updated and new information on the status of fast reactor development and on activities in the field of advanced nuclear power technology during 1993, as reported at the 27th meeting of the IWGFR held in Vienna, from 17 to 19 May 1994. Refs, figs and tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Composite nuclear fuel fabrication methodology for gas fast reactors

    Science.gov (United States)

    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

  1. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    In recent years, engineering oriented work, rather than basic research and development (R&D), has led to significant progress in improving the economics of innovative fast reactors and associated fuel cycle facilities, while maintaining and even enhancing the safety features of these systems. Optimization of plant size and layout, more compact designs, reduction of the amount of plant materials and the building volumes, higher operating temperatures to attain higher generating efficiencies, improvement of load factor, extended core lifetimes, high fuel burnup, etc. are good examples of achievements to date that have improved the economics of fast neutron systems. The IAEA, through its Technical Working Group on Fast Reactors (TWG-FR) and Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), devotes many of its initiatives to encouraging technical cooperation and promoting common research and technology development projects among Member States with fast reactor and advanced fuel cycle development programmes, with the general aim of catalysing and accelerating technology advances in these fields. In particular the theme of fast reactor deployment, scenarios and economics has been largely debated during the recent IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, held in Paris in March 2013. Several papers presented at this conference discussed the economics of fast reactors from different national and regional perspectives, including business cases, investment scenarios, funding mechanisms and design options that offer significant capital and energy production cost reductions. This Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics addresses Member States’ expressed need for information exchange in the field, with the aim of identifying the main open issues and launching possible initiatives to help and

  2. Proliferation resistant fuel cycle system for the transition from light water reactors to fast reactors

    International Nuclear Information System (INIS)

    Full text: Introduction of commercial fast reactors (FR) is predicted to start around 2050 in Japan. Effective utilization of plutonium in FR is important for the sustainable electricity generation by nuclear. Successive replacement of light water reactors (LWR) to FR will take more than 60-years and reasonable fuel cycle management is necessary during this period. The transition scenario has various unpredictable factors such as introduction speed and time of FR, and flexible fuel cycle system was proposed to respond to these factors. The system consists of LWR and FR spent fuels reprocessing for reduction of LWR spent fuel volume and FR fuel fabrication. LWR fuel reprocessing only carries out about 90% uranium removal from LWR spent fuel, then the composition of residual spent fuel called recycle material is about 50% uranium, 15% plutonium and 35% fission products + minor actinides. Recycle material is transferred to FR fuel reprocessing to recover plutonium and uranium followed by mixed oxide (MOX) fuel fabrication for FR with radioactive impurities. Depending on the introduction time of FR, recycle material (about 1/10 volume of original spent fuel) may be stored for future use. The system has some characteristics compared with ordinary system that consists of full reprocessing facilities for LWR and FR spent fuels to produce FR fresh fuels. The LWR reprocessing facility becomes much smaller due to no Pu-U recovery and fabrication. The recycle materials can supply higher content of plutonium to FR and be compactly stored in case of FR introduction delay. Plutonium always contains uranium and impurities (fission products and minor actinides), thus the system maintains high proliferation resistance. The plutonium balance was calculated under several conditions, which revealed that the system could supply enough and no excess plutonium to FR

  3. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor

    DEFF Research Database (Denmark)

    Leipold, Frank; Furtula, Vedran; Salewski, Mirko;

    2009-01-01

    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...

  4. Fast breeder reactors insertion in a D2O - natural U nuclear power plants park

    International Nuclear Information System (INIS)

    A model for the evolution of Argentine's installed nuclear power for the next 40 years is presented. The consequences of fast breeder reactors' introduction are studied in both autarchic Pu cycle and a limited reprocessing system. The passage of a reactor park like the national, of natural U - heavy water to one of fast breeder reactors, can only be obtained in a very long term due, fundamentally, to the need of Pu produced for those to feed the last ones. (M.E.L.)

  5. Experimental possibilities and fast neutron dose map of the fast neutron fields at the RB reactor facility

    International Nuclear Information System (INIS)

    The RB is an unshielded, zero power nuclear facility with natural and enriched uranium fuel (2% and 80%) and D2O as moderator. It is possible to create different configurations of non-reflected and partially reflected critical systems and to make experiments in the fields of thermal neutrons. The fields of fast neutrons with 'softened' fission spectrum are made by modifying the system: modified experimental fuel channel EFC, coupled fast-thermal system in two configurations CFTS-1 and CFTS-2, coupled fast-thermal core HERBE. The intermediate and fast neutron absorbed doses in fast neutron fields are given. In first configuration of RB reactor it was almost impossible to perform dosimetric and other experiments. By creating these fields, with in our circumstances available fuel elements, the possibilities for different experiments are greatly improved. Now we can irradiate food samples, soil samples, electronic devices, study material properties, perform various dosimetry experiments, etc. (1 tab.)

  6. Physics of Fast and Intermediate Reactors. Vol. II. Proceedings of the Seminar on the Physics of Fast and Intermediate Reactors. Vol. II

    International Nuclear Information System (INIS)

    It is generally agreed that the ultimate economic advantage of power produced by nuclear fission over that produced by conventional sources depends on the ability of a certain type of reactor to breed precious nuclear fuel out of the plentiful but not readily fissionable isotope of uranium. This fact is mainly responsible for the importance attached to the development of fast power reactors, but many other interesting properties of unmoderated or weakly moderated reactor systems have also been brought to light by reactor physicists. In August 1961 the Agency organized in Vienna a Seminar on the Physics of Fast and Intermediate Reactors, at which all the topics relating to this important branch of reactor science were discussed. The main feature of this meeting was extensive discussion of the 66 written contributions, which set the stage for a wide exchange of experience and ideas throughout 13 half- day sessions. The Seminar was attended by 132 scientists from 22 Member States and two international organizations. It is hoped that these Proceedings of the Seminar, which include both the papers presented and a record of the discussions, will be useful as a reference work both to research workers in the field and to newcomers to it for many years to come. The Agency's thanks are due to all the participating scientists for their written or oral contributions and especially to those among them who, as session chairmen, led the discussions and contributed greatly to the success of the meeting.

  7. Physics of Fast and Intermediate Reactors. V. III. Proceedings of the Seminar on the Physics of Fast and Intermediate Reactors. Vol. III

    International Nuclear Information System (INIS)

    It is generally agreed that the ultimate economic advantage of power produced by nuclear fission over that produced by conventional sources depends on the ability of a certain type of reactor to breed precious nuclear fuel out of the plentiful but not readily fissionable isotope of uranium. This fact is mainly responsible for the importance attached to the development of fast power reactors, but many other interesting properties of unmoderated or weakly moderated reactor systems have also been brought to light by reactor physicists. In August 1961 the Agency organized in Vienna a Seminar on the Physics of Fast and Intermediate Reactors, at which all the topics relating to this important branch of reactor science were discussed. The main feature of this meeting was extensive discussion of the 66 written contributions, which set the stage for a wide exchange of experience and ideas throughout 13 half-day sessions. The Seminar was attended by 132 scientists from 22 Member States and two international organizations. It is hoped that these Proceedings of the Seminar, which include both the papers presented and a record of the discussions, will be useful as a reference work both to research workers in the field and to newcomers to it for many years to come.

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

    Science.gov (United States)

    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.

  9. Network representation of design knowledge of prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    A method of design knowledge representation was studied for the Japanese fast breeder reactor Monju, aiming at enhanced understanding of engineering considerations with mutual relations. Taking over design knowledge of Monju to next generation designers/engineers to be in charge of design of future FRs is by no means easy, in contrast with operation and maintenance knowledge which can be acquired in the real plant operation and maintenance. Specifications of the as-is Monju contains only a small part of the entire design knowledge, mainly by two reasons. Firstly, reasons for selecting the as-is specifications can not be understood until reaching proper knowledge source. Secondly, there are many passed-over options on the design specifications. Reasons for passing-over these options are not always technical inferiority. A large part of the current specifications are selected because the worst possible technical value can be foreseeable or guaranteed to be acceptable within limited R and D period and resource, not because the expected value is estimated to be the lower. In other words, in the future where new materials with improved properties, faster and more accurate analysis/prediction methods, rationalized technical standards or regulatory requirements, and/or some other environment for thorough comparison among specification options are available, these passed-over options are likely to be worth reconsidering. There are a huge number of technical documents on diversified engineering studies, such as calculation of maximum possible temperature gradient of important structures, necessary sodium flow rate in particular sub-assemblies, etc. for validation of each decision making in design. A large part of these documents are scanned and stored in a data base with each catalogue data for electronic browse. The authors propose a network representation of these items of design decision making, where the items are mutually connected by directed arcs, where nodes stand

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

    Science.gov (United States)

    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.

  11. Status of Phenix operation and of sodium fast reactors in the world

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Martin, L. [Phenix plant, 30 - Bagnols sur Ceze (France); Courtois, C. [CEA Marcoule 30 (France)

    2007-07-01

    The French fast breeder reactor (FBR) Phenix restarted in 2003 after 6 years of safety reevaluation procedures. The goal of the experiments performed at Phenix is, first, to demonstrate the technical feasibility of transmutation of minor actinides and long-life products in a fast reactor and secondly, to acquire knowledge on structure materials for future energy systems and on innovative nuclear fuel concepts. After several years of Generation IV discussions, many countries have announced or confirmed their priority for the fast sodium reactor as a reference design. These countries today include Japan, China, Korea, India and Russia (simultaneously with lead reactors). The United States have announced a project for a waste-burning reactor. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision of building a prototype scheduled for operation in 2020. These declarations are all sustained in a very practical manner by ongoing events in this field. Following the excellent results obtained by the BN-600 (600 MWe), Russia has re-launched the BN-800 project. China is currently in the process of building a 75 MWt research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU (250 MWe) for divergence in 2008. In India, a 1200 MWt power reactor is under construction, scheduled for divergence in 2010, the first of 3 planned sodium reactors.

  12. Development of a three dimension multi-physics code for molten salt fast reactor

    International Nuclear Information System (INIS)

    Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)

  13. Research and development studies carried out for the seismic verification of the Italian PEC Fast Reactor

    International Nuclear Information System (INIS)

    This paper presents the main features and results of the numerical and experimental studies that were carried out by ENEA, in co-operation with ANSALDO and ISMES for the seismic verification of the Italian PEC Fast Reactor. More precisely, the paper focuses on the wide-ranging research and development programme that was performed on the reactor building, the reactor-block, the main vessel, the core and the shutdown system. The needs of these detailed studies are stressed and the general validity of the analyses in the framework of research and development activities for nuclear reactors are pointed out. The adopted design criteria and methods are presented in a separate paper, together with the effects of seismic conditions on PEC design, and comparisons with the other fast reactors of the European Community countries. (author)

  14. Inherent safety of advanced nuclear engineering based on BN-800 - type fast reactors

    International Nuclear Information System (INIS)

    Considerations based on the prolonged experience of fast reactor operations exhibiting outlook application of reactors on a basis of BN-800 with sodium coolant are given. Reliability and safety of the block are supported by the probability analysis of safety in the content of engineering project. Conversion on the reactor core with nitride fuel will significantly raise a possibility to conform to safety and nonproliferation of fission materials needs. The suggested optimum variant for reactor core on a basis of nitride fuel is advanced

  15. Exploding the myths about the fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, S.

    1979-01-01

    This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

  16. Analysis of sodium-cooled fast reactor operations world-wide and consequences for future reactor design and operation

    International Nuclear Information System (INIS)

    In 2007, the sodium-cooled fast reactors which are currently operating or have operated throughout the world will have accumulated collective experience of 379 years of operation for 18 different reactors. This paper summarizes the various incidents and problems which have impacted these reactors' operations, ranks them by function and lists the solutions which were brought. This paper solely looks at the operating problems which occurred in the 'sodium' part of the fast reactors. Problems encountered in the classic water/steam part are not described herein review is made for all these reactors on: The water/sodium reactions occurred in the steam generators; -The technical difficulties on the primary components (pumps and exchangers) and the repair operations; -The incidents in handling operations; - The spurious leaks or transfers of sodium and their consequences; - The intakes of air or impurities; -The experience from fuel and clad failures; - The neutronic operations and control; -The material behaviour problems (as 321 or SPX1 drum); - The difficulties due to sodium aerosols. And all the specific difficulties due to sodium related technological problems. This paper also describes how the experience gained has been taken into account in the safe operation of these plants and also in the design of future reactors. n conclusion it appears that a significant experience has been accumulated that allow today good availability for the remaining operating plants and also good design possibilities for the future. (author)

  17. Restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    The experimental fast reactor Joyo is the first sodium cooled fast reactor in Japan. Joyo attained initial criticality as a breeder core in April 1977 and has operated as a high performance irradiation test bed since 2003. The 15th periodic inspection of Joyo commenced in May 2007 with the Fuel Handling Machine (FHM) being set up on the Rotating Plug (R/P) for refueling in June. When the R/P was taken down, measuring the load of the Hold-Down Shaft (HDS) revealed an abnormal decrease above the in-vessel storage rack (IVS). The HDS is a cylindrical FMH device that holds down the 6 surrounding subassemblies (S/As) which are adjacent to a withdrawn S/A. In order to investigate the cause of this, an in-vessel observation was conducted using a radiation-resistant fiber scope (RRF). As a result of the observations, it was discovered that the top of the irradiation test S/A 'MARICO-2' (the material testing rig with temperature control) had bent onto the IVS as an obstacle, and had damaged the Upper Core Structure (UCS). During the investigation of this incident, the in-vessel observations using RRF etc. took place at (1) the top of the S/As and the IVS for foreign material, (2) the bottom face of the UCS for damage under the condition with the level of sodium at -50 mm below the top of the S/As. In-vessel observation techniques for a Sodium cooled Fast Reactor (SFR) are important in confirming its safety and integrity. Since an in-vessel observation for an SFR has to be conducted under severe conditions that include high temperatures (∼ 200 deg-C) and high radiation doses (∼ 400 Gy/h), and the primary sodium coolant has to be retained in the Reactor Vessel (R/V) to remove the decay heat, an in-vessel observation equipment has to be designed to not only tolerate the severe conditions but also be capable of being inserted into the sealed R/V through the fixed holes built in to the R/P and gain access to the observation areas. The in-vessel observations were successfully

  18. Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

    International Nuclear Information System (INIS)

    The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

  19. Impact of safety design considerations on fast reactor operations

    International Nuclear Information System (INIS)

    With careful choice of design features, large pool type LMFBR's can combine highly reliable engineered safeguards with favorable inherent characteristics. This paper discusses a number of areas where safety design considerations can influence reactor operation and reactor load factor. Despite efforts to reduce capital costs of LMFBRs, it is likely that there will be a significant capital cost penalty relative to thermal reactors for the foreseeable future, and thus achieving a high load factor is of major importance in achieving acceptable generation costs. Reactor load factor is affected by shut-down time required for fuel handling, maintenance, in-service inspection and shut-downs due to unavailability of safety related systems required to provide a given degree of reliability. The paper considers the influence on reactor operation of the requirement to demonstrate continuing structural integrity and adequate reliability of decay heat removal, flexibility of refuelling and timing of removal of failed fuel, and the influence of containment on access to the plant

  20. Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

    Energy Technology Data Exchange (ETDEWEB)

    Philbin, J.S.; Nelson, W.E.; Rosenstroch, B.

    1978-10-01

    The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.

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

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States); 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.

  2. Progress on reactor system technology in the FaCT project toward the commercialization of fast reactor cycle system

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) is now carrying out the 'Fast Reactor Cycle Technology Development (FaCT)' project toward the commercialization of fast reactor (FR) cycle system. The design targets including 'Safety and Reliability', 'Sustainability', 'Economic Competitiveness' and 'Nuclear Non-proliferation' have been established as the principle of specifications for FR cycle technology at the deployment stage around 2050, to contribute to the global needs which the 21st century has encountered more than ever before, such as the environmental protection and the remarkable increase of energy demand foreseen especially in developing countries. In accordance with those design targets, the design study and the related research and development (R and D) on innovative technologies for Japan Sodium-cooled Fast Reactor (JSFR) have been in progress aiming at the completion of the conceptual design stage by 2015. The demonstration reactor is planned to operate around 2025. An interim report is ready for issue in June 2009. The report will show the design specifications considered to be feasible at present to meet the requirements for the commercialization and the R and D results to support the feasibility, as well as the investigation on optional measures to take for some of the innovative technologies which may have several high technical hurdles to be realized. (author)

  3. In situ tests on the PEC fast reactor building

    International Nuclear Information System (INIS)

    This paper describes forced excitation tests carried out at the PEC reactor building, to determine seismic motion amplifications produced in the building itself. Experimental results are used to gauge numerical methodologies capable of assessing the margins existing in the design analysis. (orig./HP)

  4. The development of fast neutron reactors in France

    International Nuclear Information System (INIS)

    The French strategy is based on a coherent and carefully defined development program launched in 1950, each new stage of which is decided upon after the results of the preceding stage are analyzed. Rapsodie, the forst experimental reactor began operating in 1967 after more than 10 years of full scale test of its components. The in-pile fuel and component behaviour experience gained was put to immediate use for the design and out-of-pile tests of components for Phenix. Phenix is a prototype power generating demonstration reactor which has been operating since 1974. Rapsodie was developed for rigorous and statistical test of fuel behaviour. Super Phenix, the 1200 MWe reactor of the Creys Malville plant was ordered in 1977 and benefited from the 3 years operating experience gained with Phenix and the 10 years of operating experience acquired with Rapsodie. In 1986, after only one year of experience with Super Phenix, it is expected that all the parties involved in the financial and technical aspects of Super Phenix will be in the position to suggest the next stage in the development of large commercial plants to the government. The next reactor in the series, Super Phenix 2, is currently being studied

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

    Energy Technology Data Exchange (ETDEWEB)

    Widiawati, Nina, E-mail: nina-widiawati28@yahoo.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung (Indonesia)

    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.

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

    Science.gov (United States)

    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.

  7. Evaluation of fluid effects on the dynamic response of a fast reactor core

    International Nuclear Information System (INIS)

    The results of dynamic experiments on shaking tables, carried out in water (simulating sodium) on both single and coupled core element prototypes and core simplified mock-up configurations of the Italian PEC fast reactor test facility, with excitation gradually increasing up to above Safe Shutdown Earthquake, have been analysed by use of the one-dimensional computer program CORALIE and the two-dimensional program CLASH. The study confirmed the conservative nature of the PEC core design calculations, provided the natural frequency and damping values to be used in the calculations for the Final Safety Report, and allowed the fluid-structure interaction model to be assessed for the PEC core seismic analysis. It also contributed to the validation of the above-mentioned computer codes for their general use for the fast reactor core analysis as well as to a better understanding of fluid-structure interaction problems concerning the fast reactor core

  8. IAEA note on multi-national fuel cycle centres as related to fast breeder reactors

    International Nuclear Information System (INIS)

    The significant aspects of associating fast breeder reactor fuel cycles with the concept of regional fuel cycle centres, as studied earlier by the IAEA, are identified. The results of the RFCC Study Project are presented, and how in particular non-proliferation and safeguards, radioactive waste management and economic considerations would be effected by inclusion of fast breeder reactor fuel cycle facilities and possibly fast breeder reactors as well in such centres, are discussed. The current effort of the IAEA to develop a computer programme which models the material flows in the nuclear fuel cycle which could be applied to the analysis of alternative siting strategies for FBR and its fuel cycle facilities is discussed

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

    Science.gov (United States)

    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.

  10. Twelfth annual meeting of the International Working Group on Fast Reactors. Summary report. Part II

    International Nuclear Information System (INIS)

    Examining several alternative nuclear power scenarios through the long term it showed the comparative needs of advanced reactors for uranium and for supporting services, thereby establishing the basis for further development of uranium resources and specific reactor systems. Even with dramatic increases in known resources, nuclear power would be able to play only a temporary role in satisfying world energy needs. The use of advanced fast breeders can do much to reduce the total rate of depletion of uranium resources. Breeder reactors would provide a virtually inexhaustible source of energy supply within foreseeable extensions of known uranium resources. This document includes status reports on activities related to research, development, construction, operation, experimental data, safety issues of fast breeder reactors in Germany, Italy, European Union, USSR, OECD, Japan, USA, UK, France

  11. Plans for the development of the IFR [Integral Fast Reactor] fuel cycle

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) is a concept for a self-contained facility in which several sodium-cooled fast reactors of moderate size are located at the same site along with complete fuel-recycle and waste-treatment facilities. After the initial core loading with enriched uranium or plutonium, only natural or depleted uranium is shipped to the plant, and only wastes in final disposal forms are shipped out. The reactors have driver and blanket fuels of uranium-plutonium-zirconium alloys in stainless steel cladding. The use of metal alloy fuels is central to the IFR concept, contributing to the inherent safety of the reactor, the ease of reprocessing, and the relatively low capital and operating costs. Discharged fuels are recovered in a pyrochemical process that consists of two basic steps: an electrolytic process to separate fission products from actinides, and halide slagging to separate plutonium from uranium

  12. The European Lead Fast Reactor Strategy and the Roadmap for the Demonstrator ALFRED

    International Nuclear Information System (INIS)

    The development of the technology for a new reactor must follow a chain of gradual and progressive steps to reach maturity. This paper presents the strategy for the development of a nuclear fast reactor with closed fuel cycle and based on lead coolant technology for economical electricity production: the European Lead Fast Reactor (ELFR). This reactor has the potential to multiply by two orders of magnitude the energy output from a given amount of uranium while improving the safety of the plant and the management of high level radioactive waste (HLRW) through the transmutation of minor actinides. The roadmap of ELFR is based on the progressive up-scaling from a zero power facility to the commercial deployment of the first-of-a kind plant. The intermediate steps are a Technological Pilot Plant (MYRRHA), a Demonstrator (ALFRED) and a Prototype (PROLFR). The roadmap of the demonstrator ALFRED is also presented. (author)

  13. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    Directory of Open Access Journals (Sweden)

    Sungjoo Lee

    2016-09-01

    Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

  14. Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights

    International Nuclear Information System (INIS)

    In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)

  15. Status of National Programmes on Fast Breeder Reactors. International Working Group on Fast Reactors, Twentieth Annual Meeting, Vienna, 24-27 March 1987

    International Nuclear Information System (INIS)

    The Agenda of the meeting was as follows: 1. Approval of the Agenda. 2. Approval of the minutes of the 19th meeting of the IWGFR. 3. Report of the Scientific Secretary regarding the WD activities of the Working Group. 4. Presentations and discussions on national programmes on fast breeder reactors. 5. Consideration of conferences on fast breeder reactors. a. ANS-ENS International Conference on Fast Breeder Systems Experience Gained and Path to Economical Power Generation, Richland, Washington, USA, 13-17 September 1987. b. International Conference on Liquid Metal Engineering and Technology, Avignon, France, 17-20 October 1988. c. Other meetings of interest to IWGFR members. 6. Consideration of major recommendations of some of the WD IWGFR Specialists' Meetings. 7. Consideration of arrangements for Specialists' Meetings in 1987. a. Specialists' Meeting on Fission and Corrosion Products Behaviour in Primary Circuits of LMFBRs, Karlsruhe, Fed. Rep. of Germany, May 1987. b. Specialists' Meeting on LMFBR Reactor Block Antiseismic Design and Verification, Bologna, Italy, October 1987. 8. Selection of topics for Specialists' Meetings to be held in 1988 and suggestions of the IWGFR on other Specialists' Meetings and their justifications. 9. Consideration of joint research activities: a. Coordinated Research Programme on a Comparative Assessment of Processing Techniques for Analysis of Sodium Boiling Noise Detection Data. b. Coordinated Research Programme on Intercomparison of LMFBR Core Mechanics Codes. c. New Topics of CRP. d. Other Activities. 10. Updating of ''LMFBR Plant Parameters''. 11. Informal discussion on ''Safety Criteria for Fast Reactors in IWGFR Countries''. 12. The date and place of the 21th Annual Meeting of the IWGFR

  16. Design, Manufacturing and Irradiation Behaviour of Fast Reactor Fuel. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    Fast reactors are vital for ensuring the sustainability of nuclear energy in the long term. They offer vastly more efficient use of uranium resources and the ability to burn actinides, which are otherwise the long-lived component of high level nuclear waste. These reactors require development, qualification, testing and deployment of improved and innovative nuclear fuel and structural materials having very high radiation resistance, corrosion/erosion and other key operational properties. Several IAEA Member States have made efforts to advance the design and manufacture of technologies of fast reactor fuels, as well as to investigate their irradiation behaviour. Due to the acute shortage of fast neutron testing and post-irradiation examination facilities and the insufficient understanding of high dose radiation effects, there is a need for international exchange of knowledge and experience, generation of currently missing basic data, identification of relevant mechanisms of materials degradation and development of appropriate models. Considering the important role of nuclear fuels in fast reactor operation, the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) proposed a Technical Meeting (TM) on 'Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels', which was hosted by the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russian Federation, from 30 May to 3 June 2011. The TM included a technical visit to the fuel production plant MSZ in Elektrostal. The purpose of the meeting was to provide a forum to share knowledge, practical experience and information on the improvement and innovation of fuels for fast reactors through scientific presentations and brainstorming discussions. The meeting brought together 34 specialists from national nuclear agencies, R and D and design institutes, fuel vendors and utilities from 10 countries. The presentations were structured into four sections: R and D Programmes on FR Fuel

  17. Present status and future program of YAYOI as a fast pulse reactor

    International Nuclear Information System (INIS)

    Fast neutron source reactor YAYOI was constructed in 1971 and has been operated by the Faculty of Engineering of the University of Tokyo. The reactor is a development of AFSR and HARMONIE and is air cooled, modified to enhance flexibility for research and training, using 93% enriched uranium metal fuel. The YAYOI is principally used for LMFBR development work. The new features of YAYOI include pulsation with or without an electron linac. (author)

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

    International Nuclear Information System (INIS)

    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.

  19. A Consistent Comparative Study of Advanced Sodium-cooled Fast Burner Cores loaded with Thorium and Uranium-based Metallic Fuels

    International Nuclear Information System (INIS)

    We considered uranium-based metallic fuel of TRU-U-10Zr for driver fuel and thorium was considered as blanket because thorium blanket produces less amount of TRU than uranium blanket and use of thorium blanket leads to smaller sodium void worth than the use of uranium blanket due to the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. However, these cores using thorium blanket still have a large amount of TRU production from the driver fuels because the driver fuels contain a large amount of depleted uranium which leads to the production of TRU through neutron capture. The objective of this work is to consistently compare the neutronic performances of advanced sodium cooled fast reactor cores loaded with thorium and uraniumbased metallic fuels as driver fuel for TRU burning. Our main emphasis is given on the analyses of the differences in the core performance parameters. For consistent comparison, we used the same core configuration and all the same design parameters except for the fact that depleted uranium in uraniumbased fuel is replaced with thorium. We considered the cores having no thorium blanket and the cores having thorium blanket that were designed in our previous works

  20. Operational experience and upgrading program of the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Twenty years of successful operations at the experimental fast reactor JOYO provide a wealth of experience covering core management, chemical analysis of sodium and cover gas for impurity control, natural convection tests, upgrade of fuel failure detection system, corrosion product measurement, development of operation and maintenance support system, and replacement of major components in the cooling systems. Some of the data obtained is stored in a database to preserve the related knowledge. This experience and accumulated data will be useful for the design of future fast reactors. (author)

  1. Possibilities to optimize sodium cooled fast reactors with respect to Actinide burning

    International Nuclear Information System (INIS)

    The optimization of fast reactor cores with respect to the destruction of Minor Actinides starts from the question, whether the neutron spectrum can be hardened by changing the core geometry, core size or fuel type. Spectrum hardening would increase the destruction (fission) of actinides compared to their further production due to capture processes. Reference point of this study is the European Fast Reactor EFR. It is shown that a flat core geometry with metal fuel would offer an optimum and that small power units, i. e. smaller cores are most favourable

  2. Neutron Age Determination in Fast Reactor Materials using the Group Method

    Directory of Open Access Journals (Sweden)

    Kabanova Marina F.

    2016-01-01

    Full Text Available The article deals with the methods of identifying fast neutron age in sodium (Na and uranium-238 (238U; describes the model of advanced and effective fast neutron nuclear reactors (FN, where Na is a coolant while 238U is involved in the fuel cycle in large quantities; justifies the choice of the group method for calculating the neutron age value in the substances mentioned above that can show the accuracy of the used constants for Na and estimate various versions of multilevel description of neutron moderation in 238U – the most powerful resonance absorber of the neutron reactor active zone.

  3. Strengthening the R and D on fast reactor technology, and promoting its industrialization

    International Nuclear Information System (INIS)

    Based on the strategic thoughts of energy development in China expounded by Jiang Zemin in the article entitled 'Reflections on Energy Issues in China', the author points out in this paper that R and Ds on fast reactor technology shall be carried out timely in China by taking full advantage of international scientific resources, and overall planning in this regard shall be made as well. The point of view of strengthening fast reactor technology R and D and promoting its industrialization is also put forward in the paper. (authors)

  4. Creep-fatigue damage rules for advanced fast reactor design. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The IAEA, following the recommendations of the International Working Group on Fast Reactors, convened a Technical Committee Meeting on Creep-Fatigue Damage Rules to be used in Fast Reactor Design. The objective of the meeting was to review developments in design rules for creep-fatigue conditions and to identify any areas in which further work would be desirable. The meeting was hosted by AEA Technology, Risley, and held in Manchester, United Kingdom, 11-13 June 1996. It was attended by experts from the European Commission, France, India, Japan, the Republic of Korea, the Russian Federation and the United Kingdom. Refs, figs, tabs

  5. Fast-quench reactor for hydrogen and elemental carbon production from natural gas and other hydrocarbons

    Science.gov (United States)

    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.

  6. Application of COREMELT-3D code at analysis of severe fast reactor accidents

    International Nuclear Information System (INIS)

    The code COREMELT for calculations of initial and transition stages of severe accident is considered. It is used to conduct connected calculations of nonstationary neutronic and thermohydraulic processes in sodium fast reactor core. The code has some versions depending on dimensions of solving problem and consists of thermohydraulic module COREMELT and neutronic module RADAR. Using the code COREMELT-3D connected calculations of core disassembly accidents of ULOF and UTOP type have been conducted for sodium fast reactors safety analysis. The main problem of code COREMELT-3D use is duration of calculation, speeding of the code is possible when calculating algorithms are parallelized

  7. On an optimized neutron shielding for an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    The molten salt reactor technology has gained renewed interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner core vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all core internal structures. On the basis of this new geometry a model for neutron physics calculation is presented and applied for a shielding optimization. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system has to be significantly increased and will finally be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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.

  9. Neutronic/Thermal-hydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    International Nuclear Information System (INIS)

    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.

  10. Commercial U.S. Vendors Focus on Reducing the Cost of Fast Reactors

    International Nuclear Information System (INIS)

    A US commercial perspective: • Focus on economic benefits of fast reactors (Market Determined): • Compact design, high energy density (modularity); • High temperature output; • Improved energy conversion efficiency; • Process heat application market; • Inherent safety capabilities. • Focus on burn or “breed and burn”; • Helps avoid some non-proliferation challenges. • Focus on reactor alone, not on a re-processing plant; • Reactor is the first place for economic payback; • Happy to burn reprocessed fuel if available

  11. Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

    International Nuclear Information System (INIS)

    Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G. [N.A. Dollezhal Institute ' NIKIET' , PO Box 788, Moscow, 101000 (Russian Federation)

    2006-07-01

    This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with {beta}{sub eff}, low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

  14. Teaching sodium fast reactor technology and operation for the present and future generations of SFR users

    International Nuclear Information System (INIS)

    This paper provides a description of the education and training activities related to sodium fast reactors, carried out respectively in the French Sodium and Liquid Metal School (ESML) created in 1975 and located in France (at the CEA Cadarache Research Centre), in the Fast Reactor Operation and Safety School (FROSS) created in 2005 at the Phenix plant, and in the Institut National des Sciences et Techniques Nucleaires (INSTN). It presents their recent developments and the current collaborations throughout the world with some other nuclear organizations and industrial companies. Owing to these three entities, CEA provides education and training sessions for students, researchers, and operators involved in the operation or development of sodium fast reactors and related experimental facilities. The sum of courses provided by CEA through its sodium school, FROSS, and INSTN organizations is a unique valuable amount of knowledge on sodium fast reactor design, technology, safety and operation experience, decommissioning aspects and practical exercises. It is provided for the national demand and, since the last ten years, it is extensively opened to foreign countries. Over more than 35 years, the ESML, FROSS, and INSTN have demonstrated their flexibility in adapting their courses to the changing demand in the sodium fast reactor field, operation of PHENIX and SUPERPHENIX plants, and decommissioning and dismantling operations. The results of this ambitious and constant strategy are first sharing of knowledge obtained from experimental studies carried out in research laboratories and operational feedback from reactors, secondly standardized information on safety, and finally the creation of a 'sodium community' that debates, shares the knowledge, and suggests new tracks for a better definition of design and operating rules. (author)

  15. A review of fast reactor program in Japan (April 2001 - March 2002)

    International Nuclear Information System (INIS)

    This report describes the research and development activities on fast reactors in Japan thru April 2001 to March 2002. In December 2001, the Cabinet decided the Plan for Reorganization of Government-funded Corporations including the merger of JNC and the Japan Atomic Energy Research Institute (JAERI). A law to set up a new entity is supposed to be submitted to the National Diet by the Japanese Fiscal Year (JFY) 2004. In the Experimental Fast Reactor Joyo, thirty-five duty cycle operations and thirteen special tests with the MK-II core were completed by June 2000 without any fuel pin failures or serious plant trouble. The reactor is currently being upgraded to the MK-III core. Though a fire broke out in the maintenance building of Joyo in October 2001, the Mk-III construction work was restarted in February 2002. In the Prototype Fast Breeder Reactor Monju, countermeasures against sodium leakage have already been drawn up based on Monju comprehensive safety review. The safety licensing examination for the plant modification of Monju is undergoing. As for the Feasibility Study on Commercialized Fast Reactor Cycle Systems, JFY2001 was the first year of its second phase. A three-year period from JFY2001 to 2003 is the initial term of this phase. During this term, research activities are being focused on the design of the candidate concepts and fundamental tests of key technologies. An interim summary of these activities will be checked and reviewed, and based on the results; the research for JFY 2004 to 2005 will be conducted in order to narrow down the number of alternatives for the fast reactor cycle. (author)

  16. Status of fast breeder reactor development in India

    International Nuclear Information System (INIS)

    The energy scenario and economic conditions in India are presented. India needs considerable energy for its rapid industrialisation with the liberal economic policy. Nuclear energy with FBR is the only large scale energy resource other than coal, available in the country. The present economic constraints have delayed the construction of new NPPs. The performance of operating reactors has improved considerably during the year. Operating experience of FBTR has been detailed particularly the reactivity incident and its investigations. Updated design of 500 MWe PFBR is presented. Various R and D works in support of FBR in the engineering, metallurgy, chemistry, reprocessing, safety etc. are detailed. (author)

  17. Seismic analysis of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    This report is a general survey of the recent methods to predict the seismic structural behaviour of LMFBRs. It shall put into evidence the impact of seismic analysis on the design of the different structures of the reactor. This report is addressed to specialists and institutions of governmental organizations in industrialized and developing countries responsible for the design and operation of LMFBRs. The information presented should enable specialists in the R and D institutions and industries likely to be involved, to establish the correct course of the design and operation of LMFBRs. Also, the safety aspect of seismic risk are emphasized in the report. Refs and figs

  18. The status of fast reactors programme in France in 2005

    International Nuclear Information System (INIS)

    From an overall point of view, four events must be pointed out: The new French law on energy: After the national debate organised in 2003 on energy policies for the next 30 years to come, the French government issued a White book on Energy, which contributed to the debate and gathered many contributions from French national representative organisations. In 2005, the French government proposed a new bill, adopted by French parliament on June 3, 2005. This law identified four major ideas for the future development of energy in France: - The economising of energy for a mastery of energy consumption, - The diversification of French energy resources with three activities involved: - Keeping the nuclear option open and building an EPR reactor as an industrial prototype before launching a series. - Supporting the development of renewable energies such as solar energy, biomass (including wood and bio-fuels), hydraulics and wind energy with the objective of reaching 21% of the electricity production in 2010. - Ensuring the safety of supply in the fields of oil, gas and coal. - The development of research in the energy sector, particularly an improvement of competitiveness for renewable energies, support for the nuclear energy option, the development of future reactor technologies, the hydrogen option. - The transport and distribution of energy that contributes considerably to economic and social development. This law, governing the orientations of the new energy policy also provides for the continued support of the nuclear option in France, which is a decisive option for three of the objectives in terms of the new energy policy defined within the texts of the law, namely 1) safeguard the independence of the French national energy resources and safety in their procurement, 2) fight against the Green House Effect and 3), ensure regular and competitive prices for electricity. It is therefore of the greatest importance to keep all options open for the replacement of present

  19. Commission of the European Communities: Review of fast reactor activities performed during 1991

    International Nuclear Information System (INIS)

    The Commission of the European Community continued its fast reactor activities on the same lines as in previous years, performing harmonisation and research activities. The Fast Reactor Coordinating Committee (FRCC) pursued the exchange of information and the discussions of specific problems, e.g. the influence of LMFBR design characteristics on decommissioning. These discussions were based on the results of a study performed under the CEC Decommissioning Programme. It was concluded that decommissioning of fast reactors, in spite of some specific problems, is not more difficult than that of thermal reactors. In December 1991 the Council of Ministers approved a Reactor Safety programme which will be performed as a concerted action. The programme is mainly centred on the confinement of radioactivity under severe accidents in thermal reactors, but the inclusion of some fast reactor safety aspects could be envisaged. The FRCC has asked the Safety Working Group (SWG) discussed possible topics. So far the development of the CONTAIN code for fast reactors was found most suitable. The group had regular reports on highlights from national research and development programmes and discussed various safety aspects, e.g. the safety options applied to EFR, the safety of metal fuelled or lead cooled fast reactors. The group supported the idea launched by the IWGFR to produce a safety R and D index jointly with IAEA, if all Non-Community countries having a fast reactor programme are participating. The WAC group continued the exchange of information on code developments for severe accident analysis in national laboratories and at the EC Joint Research Centre and discussed specific modelling aspects, e.g. early transition phase phenomena. The group also performed further comparative calculations. An initiative of the IPPE Institute at Obninsk (Russia) to perform comparative calculations for hypothetical accidents with laboratories in Western Europe was supported. The activities of

  20. Uncertainties in analysis of innovative lead-cooled fast reactors

    International Nuclear Information System (INIS)

    There are numerous uncertainties in the prediction of innovative reactor design, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes the problems encountered in the analysis of the Encapsulated Nuclear Heat Source (ENHS) core benchmark and the new cross section libraries developed to overcome these problems. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and i designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tools benchmarked include MOCUP-a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on ENDF/B-VI evaluations; and KWO2-a coupled KENO-V.a and ORIGEN2.1 code with ENDFB-V.2 based 238 groups library. Uncertainties in the cross sections of lead were found particularly large and deserve careful evaluation. (author)

  1. Fast-mixed spectrum reactor. Progress report for 1980

    International Nuclear Information System (INIS)

    Reactor physics, fuel cycle, thermal-hydraulics and fuel cycle cost studies have been performed for this concept and are reported. The most serious drawback of previous FMSR designs, namely the level of irradiation damage to the stainless steel of the cladding and duct materials, has been greatly reduced by the new design. The peak fuel burnup level is also reduced. Work continued on earlier FMSR designs, and in particular, the centrally-moderated FMSR. Emphasis was placed on defining the first core and then the total sequence of core histories over the 30-year life of the reactor. It was found possible to define a two-year fuel cycle with limited reactivity swing over the cycle. Fuel cycle cost studies were begun. The results indicate a modest fuel cycle cost advantage for the FMSR, but the basic cost assumptions must be improved for metal fuel. Improved thermal-hydraulic analysis capabilities have greatly improved the understanding of heat transfer behavior, particularly where thter catalyst were synthesized, and seven were tested in the gas phase. In Task 3, the distributor plate foen conversion can be controlled by adjusting external variables of the coss two- fatigue stren obtained

  2. Status of national programmes on fast breeder reactors. Eighteenth annual meeting, Vienna, Austria, 16-19 April 1985

    International Nuclear Information System (INIS)

    The Eighteenth Annual Meeting on the Status of National Programmes in Member States of the IAEA on Fast Breeder Reactors had been held in April 1985. The representatives of the Member States and international organizations reported status and activities in the field of fast breeder reactors development and operation. A separate abstract was prepared for each of the 12 presentations of the meeting

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

    Science.gov (United States)

    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.

  4. Technical meeting on decommissioning of fast reactors after sodium draining. Working material

    International Nuclear Information System (INIS)

    The objective of the technical meeting was to provide a forum for in-depth scientific and technical exchange on topics related to the decommissioning experience with fast reactors, in particular with regard to the decommissioning of components after sodium draining. Accordingly, the scope of the meeting covers the review and analyses of the experience gained from the decommissioning of both active sodium loops and sodium cooled fast reactors (e.g., KNK II, Superphenix, RAPSODIE, EBR-II, FERMI, BN-350, BR-10). It is expected that the outcome of the meeting will contribute to the Agency initiative to preserve fast reactor data and knowledge. The main focus of the technical meeting was given on the decommissioning of both active loop and reactor components (e.g., the primary vessel of a sodium-cooled reactor) that have been drained of sodium, but that still conserve some residual amounts of sodium (e.g., films covering the entire surface of the component, or particular sodium heels that cannot be drained)

  5. Overview of pool hydraulic design of Indian prototype fast breeder reactor

    Indian Academy of Sciences (India)

    K Velusamy; P Chellapandi; S C Chetal; Baldev Raj

    2010-04-01

    Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multiphysics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

  6. Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D.; Herring, J.S.; Macdonald, P.E. [Idaho National Engineering and Environment Lab., Advanced Nuclear Energy, Idaho (United States)

    2001-07-01

    Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic 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. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  9. Improving Nuclear Safety of Fast Reactors by Slowing Down Fission Chain Reaction

    Directory of Open Access Journals (Sweden)

    G. G. Kulikov

    2014-01-01

    Full Text Available Light materials with small atomic mass (light or heavy water, graphite, and so on are usually used as a neutron reflector and moderator. The present paper proposes using a new, heavy element as neutron moderator and reflector, namely, “radiogenic lead” with dominant content of isotope 208Pb. Radiogenic lead is a stable natural lead. This isotope is characterized by extremely low micro cross-section of radiative neutron capture (~0.23 mb for thermal neutrons, which is smaller than graphite and deuterium cross-sections. The reflector-converter for a fast reactor core is the structure capable of transforming some part of prompt neutrons leaked from the core into the reflected neutrons with properties similar to those of delayed neutrons, that is, sufficiently large contribution to reactivity at the level of effective fraction of delayed neutrons and relatively long lifetime, comparable with lifetimes of radionuclides-emitters of delayed neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in the fast reactor. This can improve the fast reactor safety and reduce some requirements to the technologies used to fabricate fuel for the fast reactor.

  10. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  11. Toward a sustainable energy supply with reduced environmental burden. Development of metal fuel fast reactor cycle

    International Nuclear Information System (INIS)

    CRIEPI has been studying the metal fuel fast reactor cycle as an outstanding alternative for the future energy sources. In this paper, development of the metal fuel cycle is reviewed in the view point of technological feasibility and material balance. Preliminary estimation of reduction of the waste burden due to introduction of the metal fuel cycle technology is also reported. (author)

  12. SACRD: a data base for fast reactor safety computer codes, general description

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Forsberg, V.M.; Raiford, G.B.; Arwood, J.W.; Simpson, D.B.; Flanagan, G.F.

    1979-01-01

    SACRD is a data base of material properties and other handbook data needed in computer codes used for fast reactor safety studies. Data are available in the thermodynamics, heat transfer, fluid mechanics, structural mechanics, aerosol transport, meteorology, neutronics, and dosimetry areas. Tabular, graphical and parameterized data are provided in many cases. A general description of the SACRD system is presented in the report.

  13. The IAEA initiative on fast reactor data retrieval and knowledge preservation

    International Nuclear Information System (INIS)

    The paper introduces the IAEA initiative on fast reactor data retrieval and knowledge preservation. Apart from giving the rationale for the initiative and describing it, the paper outlines the road map of the initiative and explains IAEA's role in its implementation. The status of the initiative is summarised, and a concrete work plan for the future activities proposed. (author)

  14. Fast Reactor Development for a Sustainable Nuclear Energy Supply in China

    International Nuclear Information System (INIS)

    Nuclear energy is a new member of the energy supply family in China. Satisfactory operating records of all 11 nuclear power plants in China encourage its stepwise and large scale use and the PWR-FBR route matched with a closed nuclear fuel cycle forms a basic strategy. The sufficient utilization of nuclear resources and the treatment of highly radioactive waste by transmutation in fast reactors are the key issues for a sustainable development of nuclear energy. As the first step in FBR engineering development, the 65 MW(th) China Experimental Fast Reactor is approaching startup, the conceptual design of the 600-900 MW(e) China Demonstration Fast Reactor (CDFR) has been started and the 1000-1500 MW(e) China Demonstration Fast Breeder Reactor is under consideration. Three FBR development strategy targets are as follows: (1) To start realizing CDFR type commercial utilization in small batches by 2030; (2) To increase nuclear capacity to 240-250 GW(e), representing about 16%, mainly through FBRs by 2050; (3) To replace coal fired plants by nuclear power on a large scale in the period 2050-2100. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Commission of the European Communities review of fast reactor activities - March 1986

    International Nuclear Information System (INIS)

    As had been said at earlier meetings, the fast reactor activities of the Commission of the European Communities (CEC) fall into two categories, i.e. coordination and harmonisation and R and D activities. The main achievements made during the last year are briefly outlined

  17. Challenges in the design of Waste Management Plant (WMP) for Fast Reactor Fuel Cycle Facility (FRFCF)

    International Nuclear Information System (INIS)

    A 500 MW, Prototype Fast Breeder Reactor (PFBR) is being constructed by DAE at Kalpakkam, Tamilnadu. Fast Reactor Fuel Cycle Facility (FRFCF) is also being planned as an integrated facility in close proximity to PFBR to cater to the need of closing the fast reactor fuel cycle. The design of waste management plant of FRFCF is one of the important and challenging tasks in FRFCF. The high burn up, short cooling period and high fissile content of spent fuel input from fast Reactors to FRFCF are the main reasons for the difficulties / challenges posed in the management of wastes from FRFCF. Separation of actinides and Platinum Group Metals (PGM) from high level liquid wastes, higher waste oxide loading in vitrified waste product from high level liquid waste management, development of better sorbent and chemical treatment method for achieving higher decontamination factors in the treatment of low and intermediate level liquid wastes and separation of plutonium from spent / de-graded solvent waste are some of the challenges which are addressed in this paper. Solid waste management is also associated with major challenges because of high volume of alpha contaminated wastes. Proper segregation, assaying, conditioning and treatment for volume reduction are planned. Management of significant amount of solid wastes generated as fuel hardware components (alpha and non-alpha) through proper storage and retrieval for future conditioning / disposal are also addressed in this paper. This paper describes the challenges faced during the design and solutions provided for each system. (author)

  18. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs

  19. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  1. A concept of self-completed fuel cycle based on lead-cooled nitride-fuel fast reactors

    International Nuclear Information System (INIS)

    A concept of nuclear energy total system was studied based on the nitride fuel cycle and inherent safety lead-cooled fast reactors. In the nitride fuel reprocessing, a new concept for pyrochemical method was proposed due to reducing fuel cycle cost. The present designed lead-cooled fast reactors have higher safety, economics and minor actinide transmutation efficiency than those of MOX-fuel fast reactors. The construction of 1500 MWt plant is feasible as a result for technology studies for aseismic, steam-generator and reactor configuration systems. (author)

  2. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  4. Temperature-gradient induced circulation in liquid metal-fueled fast reactor systems

    International Nuclear Information System (INIS)

    This paper introduces a concept for a liquid metal-fueled fast reactor plant. The liquid metal fuel is a low-volume fraction plutonium-magnesium alloy (melting point ∼650degC). The reactor is formed around a large pool or vessel holding the liquid fuel. The fuel is cooled with heat exchangers placed at the perimeter of a reactor vessel. The molten fuel mixture undergoes circulation due to the temperature gradients in the reactor and heat exchangers. Such a reactor should have greater safety than present reactor types. Other potential benefits could include the retention of selected fission products somewhere in the reactor for irradiation and transmutation. The discussion of this concept begins with a brief review of liquid-fueled reactors, followed by a description and results of a very simplified analysis of the proposed concept. Materials aspects are addressed, and one group diffusion theory is used to estimate the critical radius of a reflected spherical system as a function of plutonium content. A simple model is developed to estimate the thermal-hydraulic behavior for a cylindrical geometry, Safety aspects and other factors are also discussed. (author)

  5. Status of Fast Reactor and Pyroprocess Technology Development in the Republic of Korea

    International Nuclear Information System (INIS)

    A fast reactor system with pyroprocess technology is one of the most promising options for electricity generation, with an efficient utilization of uranium resources and a reduction of radioactive wastes. On the experience gained during the development of the conceptual designs for KALIMER reactors, Korea Atomic Energy Research Institute (KAERI) is currently developing advanced sodium cooled fast reactor (SFR) design concepts that can better meet the Generation IV technology goals. The long term SFR development plan will be carried out with the aim of constructing an advanced SFR demonstration plant by 2028. For the development of pyroprocess technology, KAERI is currently establishing a pyroprocess integrated inactive demonstration facility (PRIDE), a mock-up facility for pyroprocessing, to produce the engineering data to be incorporated into the design of an engineering scale pyrochemical process facility, which is scheduled to be constructed by 2016. (author)

  6. Liquid Metal Fast Breeder Reactor Program: Argonne facilities

    Energy Technology Data Exchange (ETDEWEB)

    Stephens, S. V. [comp.

    1976-09-01

    The objective of the document is to present in one volume an overview of the Argonne National Laboratory test facilities involved in the conduct of the national LMFBR research and development program. Existing facilities and those under construction or authorized as of September 1976 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. The volume is divided into two sections: Argonne-East and Argonne-West. Introductory material for each section includes site and facility maps. The profiles are arranged alphabetically by title according to their respective locations at Argonne-East or Argonne-West. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc., involved in the LMFBR program is appended.

  7. Comparison of transient analysis of LBE-cooled fast reactor and ADS under loss of heat sink accident

    International Nuclear Information System (INIS)

    Highlights: • A natural circulation LBE-cooled research reactor is developed by Institute of Nuclear Energy Safety Technology. • The safety advantages for 10 MWth fast and ADS reactors under loss of heat sink were compared. • A neutronics and thermal-hydraulics coupled simulation code NTC was employed in this study. - Abstract: A natural circulation LBE-cooled research reactor developed by Institute of Nuclear Energy Safety Technology (INEST) is proposed and designed to be 10 MWth. The reactor has two operation modes, which are LBE-cooled critical fast reactor mode and LBE-cooled accelerator-driven sub-critical system (ADS) reactor mode. In this paper, protected loss of heat sink (PLOHS) and unprotected loss of heat sink (ULOHS) transient accidents were simulated for both the critical and ADS reactors above by NTC-2D code, aiming at the investigation of the safety advantages for the two reactor modes under this typical transient condition (loss of heat sink, LOHS). The results showed that during PLOHS for both the two modes, all the key parameters (core power, fuel, cladding and coolant temperatures in the hottest channel) decreased to very small values after the reactor scrammed, which meant the reactors under the two modes were both safe. For ULOHS, the fuel, cladding and coolant temperatures of the fast reactor increased smaller than those of the sub-critical one, which means the fast reactor had a better safety advantage under LOHS transient

  8. Aspects of the physics and chemistry of water radiolysis by fast neutrons and fast electrons in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCracken, D.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Tsang, K.T. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Laughton, P.J

    1998-09-01

    Detailed radiation physics calculations of energy deposition have been done for the coolant of CANDU reactors and Pressurized Water Reactors (PWRs). The geometry of the CANDU fuel channel was modelled in detail. Fluxes and energy-deposition rates for neutrons, recoil ions, photons, and fast electrons have been calculated using MCNP4B, WIMS-AECL, and specifically derived energy-transfer factors. These factors generate the energy/flux spectra of recoil ions from fast-neutron energy/flux spectra. The energy spectrum was divided into 89 discrete ranges (energy bins).The production of oxidizing species and net coolant radiolysis can be suppressed by the addition of hydrogen to the coolant of nuclear reactors. It is argued that the net dissociation of coolant by gamma rays is suppressed by lower levels of excess hydrogen than when dissociation is by ion recoils. This has consequences for the modelling of coolant radiolysis by homogeneous kinetics. More added hydrogen is required to stop water radiolysis by recoil ions acting alone than if recoil ions and gamma rays acted concurrently in space and time. Homogeneous kinetic models and experimental data suggest that track overlap is very inefficient in providing radicals from gamma-ray tracks to recombine molecular products in ion-recoil tracks. An inhomogeneous chemical model is needed that incorporates ionizing-particle track structure and track overlap. Such a model does not yet exist, but a number of limiting cases using homogeneous kinetics are discussed. There are sufficient uncertainties and contradictions in the data relevant to the radiolysis of reactor coolant that the relatively high CHC's (critical hydrogen concentration) observed in NRU reactor experiments (compared to model predictions) may be explainable by errors in fundamental data and understanding of water radiolysis under reactor conditions. The radiation chemistry program at CRL has been focused to generate quantitative water-radiolysis data in a

  9. Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

    2014-10-15

    Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

  10. Modern passive safety system for the advanced fast reactors with sodium cooling

    International Nuclear Information System (INIS)

    In fast reactors with sodium coolant it is possible to avoid serious damages of a core even at the heaviest scripts of development of accidents if to provide influence on reactance with the help of various type of Passive Safety System (PSS). With input of the PSS the reactor gets an additional negative feedback on the reactance, working at an output of the basic operational parameters (temperature, the coolant flow rate, power) for maximum permissible sizes. The scientific-technical and patent sources analysis has shown, that now it is already offered more than two hundred the various devices, capable to carry out functions of the fast reactors PSS. Comparison of various types of PSS is carried out under 9 generalized characteristics including: passivity, thresholdness, generation of efforts, inertia, multi-channels, stability to operational factors, refusal safety, simplicity and presentation, development conditions. For quantitative comparison of the device ''the perfection degree'' (K≤1) was defined as average size under 9 generalized characteristics. From the considered types of fast reactors PSS the most perfect now are fusible Lyophobic devices, basically meeting the requirements on all characteristics. Results of Lyophobic Passive Safety System development for the advanced fast reactors with sodium cooling are considered. Serviceability of the offered designs is proved experimentally at various operation temperatures on breadboard models sylphon devices and devices of type the sylphon-container with various lyophobic liquids: alloy Wuds (Tmt=80,0 deg C), an alloy lead-bismuth (Tmt=123,5 deg C), cadmium (Ttm=320,0 deg C), aluminium (Tym=660,0 deg C), developed Lyophobic Fusible Passive Safety System on excess of temperature are of interest for nuclear power installations of various type, first of all, as passive devices scram reactor and protection of the process equipment. (author)

  11. Design study and R and D progress on Japan sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    This paper describes the progress of the design study and research and development (R and D) for the Japan Sodium-cooled Fast Reactor (JSFR) implemented in the 'Fast Reactor Cycle Technology Development (FaCT)' project. A sodium-cooled fast reactor with an electric power of 1,500 MWe is targeted for commercialization at around 2050, and a demonstration reactor assuming a power output from 500 to 750 MWe is planned to start operation at around 2025. R and D on innovative technologies to achieve economic competitiveness and enhance reliability and safety is carried out for the commercialization. A compact reactor vessel without a vessel wall cooling system is pursued in consideration of the wall thickness enough to resist the severest seismic condition. A two-loop cooling system with shortened high-chromium steel piping is a crucial feature, and studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being carried out. A double-walled straight tube steam generator is investigated to enhance the reliability against sodium/water reaction, and developmental works are progressing, including the thermal-hydraulic design and trial manufacturing for components. Self-Actuated Shutdown System (SASS) is being developed with safety analysis of the applicability for JSFR and experimental demonstration in the experimental fast reactor JOYO. An advanced fuel handling system is pursued to enhance economic performance. In parallel with considering the necessity of studies on alternative technologies, discussion on whether the innovative technologies can be adopted for JSFR is in progress to be finalized in 2010. (author)

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, Anton [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Kim, Taek K. [Argonne National Lab. (ANL), Argonne, IL (United States); Middleton, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    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.

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

  17. Actinide recycle potential in the IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    Rising concern about the greenhouse effect reinforces the need to reexamine the question of a next-generation reactor concept that can contribute significantly toward substitution for fossil-based energy generation. Even with only the nuclear capacity on-line today, world-wide reasonably assured uranium resources would last for only about 50 years. If nuclear is to make a significant contribution, breeding is a fundamental requirement. Uranium resources can then be extended by two orders of magnitude, making nuclear essentially a renewable energy source. The key technical elements of the IFR concept are metallic fuel and fuel cycle technology based on pyroprocessing. Pyroprocessing is radically different from the conventional PUREX reprocessing developed for the LWR oxide fuel. Chemical feasibility of pyroprocessing has been demonstrated. The next major step in the IFR development program will be the full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. IFR fuel cycle closure based on pyroprocessing can also have a dramatic impact on the waste management options, and in particular on the actinide recycling. 6 figs

  18. Comparison study on cell calculation method of fast reactor

    International Nuclear Information System (INIS)

    Effective cross sections obtained by cell calculations are used in core calculations in current deterministic methods. Therefore, it is important to calculate the effective cross sections accurately and several methods have been proposed. In this study, some of the methods are compared to each other using a continuous energy Monte Carlo method as a reference. The result shows that the table look-up method used in Japan Nuclear Cycle Development Institute (JNC) sometimes has a difference over 10% in effective microscopic cross sections and be inferior to the sub-group method. The problem was overcome by introducing a new nuclear constant system developed in JNC, in which the ultra free energy group library is used. The system can also deal with resonance interaction effects between nuclides which are not able to be considered by other methods. In addition, a new method was proposed to calculate effective cross section accurately for power reactor fuel subassembly where the new nuclear constant system cannot be applied. This method uses the sub-group method and the ultra fine energy group collision probability method. The microscopic effective cross sections obtained by this method agree with the reference values within 5% difference. (author)

  19. Research and development studies on the seismic behaviour of the PEC fast reactor

    International Nuclear Information System (INIS)

    As introduction to the meeting, this paper provides an overview on the extensive research and development studies performed by ENEA, in co-operation with ANSALDO and ISMES, in the framework of the seismic verification of the Italian PEC fast reactor. The purpose is also to stress the reasons why a wide-ranging experimental programme and detailed numerical analysis, validated on the test results, have been performed for the PEC reactor building and the main vessel. Thus, after some notes on the high levels of the design earthquakes adopted for PEC and the important features of fast reactors in general and PEC as a specific case (making it particularly sensitive to seismic excitations), the paper presents the studies performed for the reactor-block, the core and the shutdown system, summarizing their main features and showing some of the main results. Furthermore, the non-negligible feed-backs of the seismic studies on the reactor-block design are recalled, and the needs of checking seismic design analysis of the main vessel and the reactor building are explained. The on-site experimental programme and the related numerical analysis concerning the main vessel and the reactor building are also shortly described: however, specific papers will present more details on these studies, and will also stress the usefulness of the on-site tests performed on the reactor building for the optimization of the PEC seismic monitoring system. Finally, the Italian lecture invited to this meeting will provide an overview on the state-of-the-art on on-site testing and seismic monitoring in Italy, stressing the perspective of adopting methodologies similar to those used for PEC, for nuclear power plants in general. (author)

  20. Fast pyrolysis of sugarcane and cassava residues in a free-fall reactor

    International Nuclear Information System (INIS)

    Fast pyrolysis of agricultural residues from sugarcane and cassava plantations was carried out in a laboratory-scale free-fall reactor unit. The objectives of this work were to investigate the effects of biomass types and pyroysis conditions, such as reactor temperature, condensation temperature, nitrogen flow rate and run duration, on pyrolysis product distribution, as well as to study the basic properties of the products. The results showed that all of the parameters affected the product distribution. The optimum reactor temperatures for maximising bio-oil yield were in the range of 350–450 °C. About 70 wt% of bio-oil yield could be obtained by pyrolysis of cassava stalk at a reactor temperature of 450 °C and a primary condensation temperature of 10 °C. It was also found that the minimum flow rate of nitrogen for obtaining high bio-oil yield was 1.5 l/min. The product characterisation showed that the bio-oil and char produced from the agricultural residues with the free-fall reactor unit were to a certain extent similar to those produced from different types of biomass with different types of pyrolysis reactor configurations. -- Highlights: ► Fast pyrolysis of sugarcane and cassava agricultural residues was carried out in a laboratory-scale free-fall reactor unit. ► The effects of process parameters on product yields were investigated. ► The process parameters included reactor temperature, condensation temperature, nitrogen flow rate and run duration. Basic properties of pyrolysis products were examined.

  1. Modern Approaches to Safety Assurance of a New Generation of Sodium Fast Reactors

    International Nuclear Information System (INIS)

    In the stage of designing sodium cooled fast reactors (SFRs) of a new generation there is a task to improve their inherent safety up to the level higher than that of the previous SFR designs. Modern safety requirements to the SFR of the fourth generation are described. Through the example of the BN-1200 reactor, approaches to safety assurance are demonstrated using development of inherent safety properties up to a brand new level compared to that of the earlier commercial reactor designs (BN-600 and BN-800). Also, passive safety devices and systems applied in the BN-1200 design are described. The goal is to eliminate the necessity for evacuation of residents under conditions of any possible realistic accidents. The paper presents properties of inherent self-protection of the BN-1200 reactor and estimation of their effectiveness in terms of safety assurance. The basic design approaches concerning safety are considered, including additional measures as applied to the BN-1200 reactor. These include measures aimed at the elimination or minimization of sodium leaks; design approaches to the passive shutdown systems (PSS) using various operation principles, namely: hydraulically suspended absorber rods operating in case of coolant flow rate decrease (PSS-H) and absorber rods operating in the case of increase of the core outlet coolant temperature above a certain value (PSS-T); passive decay heat removal system; sodium plenum above the core; gastight compartment above the reactor; core catcher made of refractory metal; and reactor guard vessel. (author)

  2. The Last Twenty Years of Experience with Fast Breeder Reactors: Lessons Learnt and Perspectives

    International Nuclear Information System (INIS)

    India has made significant achievements in the design and development of sodium cooled fast breeder reactors over the last twenty years. Attaining a maximum burnup of 165 GW.d/t for the plutonium-rich carbide fuel without any cladding failure, coupled with excellent performance of sodium components, including primary pumps, heat exchangers and steam generators over the last 24 years, reprocessing of carbide fuel with a burnup of 150 GW.d/t and engineering tests performed for validating the plant dynamics computer codes, are the achievements from the successful operation of the 40 MW(th) capacity loop type fast breeder test reactor. Indigenous design of the 500 MW(e) Prototype Fast Breeder Reactor (PFBR), executing high quality multidisciplinary R and D and successful manufacturing and erection of large dimensioned thin walled shell structures are the achievements in PFBR development. These achievements, apart from providing confidence in the PFBR project, are instrumental for the design of innovative future reactors. National and international collaboration established with R and D establishments and academic institutions would go a long way towards helping India to attain world leadership by 2020. (author)

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

    Science.gov (United States)

    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.

  4. NEA Activities in Preserving, Evaluating and Applying Data from Fast Reactor Experiments

    International Nuclear Information System (INIS)

    The goal of the OECD Nuclear Energy Agency (NEA) in the area of nuclear science is to help member countries identify, collate, develop and disseminate the basic scientific and technical knowledge required to ensure safe and reliable operation of current nuclear systems and to develop next generation technologies. Within these general goals, the current nuclear science programme has three key objectives: (i) to help advance the existing scientific knowledge needed to enhance the performance and safety of current nuclear systems, (ii) to contribute to building a solid scientific and technical basis for the development of future generation nuclear systems and (iii) to support the preservation of essential knowledge in the field of nuclear science. As part of the second and third of these objectives, an extensive programme of work to preserve and evaluate data from integral experiments has been established, including reactor physics, shielding and criticality safety experiments on fast reactor systems. Data from experimental facilities are reviewed and, if necessary, archives of information are made safe. This may typically involve the indexing and scanning of key documents and archiving of logbooks, for example. Selected experiments go through a detailed evaluation process and where deemed appropriate, a benchmark description is created in a standardized format for inclusion in one of the NEA Data Bank international databases. This information is used extensively by the international nuclear science community to validate their modelling and simulation tools. The process can be viewed as part of a broader knowledge management function, where information is gathered, evaluated, linked and made accessible to a wide range of users. The presentation gives details of the main databases maintained and developed by the NEA, focusing on those related to fast reactor applications. The status of recent preservation activities for fast reactor archives in the United Kingdom is

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. An Advanced Option for Sodium Cooled TRU Burner Loaded with Uranium-Free Fuels

    Energy Technology Data Exchange (ETDEWEB)

    You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    The sodium cooled fast reactors of this kind that are called burners are designed to have low conversion ratio by reducing fuel volume fraction or reducing neutron leakage or increasing neutron absorption. However, the typical SFR burners have a limited ability of TRU burning rate due to the fact that they use metallic or oxide fuels containing fertile nuclides such as {sup 238}U and {sup 232}Th and these fertile nuclides generate fissile nuclides through neutron capture even if they are designed to have low conversion ratio (e.g., 0.6). To further enhance the TRU burning rate, the removal of the fertile nuclides from the initial fuels is required and it will accelerate the reduction of TRUs that are accumulated in storages of LWR spent fuels. However, it has been well-known 4 that the removals of the fertile nuclides from the fuel degrade the inherent safety of the SFR burner cores through the significant decrease of the fuel Doppler effect, the increase of sodium void reactivity worth, and reduction of delayed neutron fraction. In this work, new option for the sodium cooled fast TRU burner cores loaded with fertile-free metallic fuels was proposed and the new cores were designed by using the suggested option. The cores were designed to enhance the inherent safety characteristics by using axially central absorber region and 6 or 12 ZrH1.8 moderator rods per fuel assembly. For each option, we considered two different types of fertile-free ternary metallic fuel (i.e., TRU-W-10Zr and TRU-Ni-10Zr). Also, we performed the BOR (Balance of Reactivity) analyses to show the self-controllability under ATWS as a measure of inherent safety. The core performance analysis showed that the new cores using axially central absorber region substantially improve the core performance parameters such as burnup reactivity swing and sodium void reactivity worth.

  7. Analysis of the transmutation of actinides minority in a sodium cooled fast reactor; Analisis de la transmutacion de actinidos minoritarios en un reactor rapido refrigerado por sodio

    Energy Technology Data Exchange (ETDEWEB)

    Ochoa Valero, R.

    2011-07-01

    Fast reactors represent a highly sustainable source of energy due to the use of a closed fuel cycle, which makes better use of natural resource and reducing the volume and heat load of high level radioactive waste.

  8. Decay heat removal in pool type fast reactor using passive systems

    Energy Technology Data Exchange (ETDEWEB)

    Parthasarathy, U. [Thermal Hydraulics Section, Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Sundararajan, T. [Department of Mechanical Engineering, IIT-Madras, Chennai 600 036 (India); Balaji, C., E-mail: balaji@iitm.ac.in [Department of Mechanical Engineering, IIT-Madras, Chennai 600 036 (India); Velusamy, K.; Chellapandi, P.; Chetal, S.C. [Thermal Hydraulics Section, Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Three dimensional thermal hydraulic analysis of decay heat system in a fast reactor model predictions compared with experimental results from PHENIX. Black-Right-Pointing-Pointer Calculations confirm adequacy of natural convection in decay heat removal. Black-Right-Pointing-Pointer Inter-wrapper flow found to reduce peak temperatures by 50 K in the blanket zone. - Abstract: Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the

  9. Decay heat removal in pool type fast reactor using passive systems

    International Nuclear Information System (INIS)

    Highlights: ► Three dimensional thermal hydraulic analysis of decay heat system in a fast reactor model predictions compared with experimental results from PHENIX. ► Calculations confirm adequacy of natural convection in decay heat removal. ► Inter-wrapper flow found to reduce peak temperatures by 50 K in the blanket zone. - Abstract: Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the results, it is concluded that the delay in initiation of SGDHRS, replacement

  10. Status of national programmes on fast reactors 1995-1996. Working material

    International Nuclear Information System (INIS)

    At present nuclear power accounts for approximately 17% of total electricity generation worldwide. Given continuing population growth and the needs of the third world and developing countries to improve their economic performance and standard of living, energy demand is expected to continue to grow through the 21st century. The proportion of energy supplied as electricity is also expected to continue to increase. Although fossil-fuelled electricity generation is the option preferred by several countries for the short term, there are rising concerns over climatic consequences caused by extended burning of fossil fuels as a result of the demands of a fast expanding world population. In this situation nuclear electricity will become more and more important and the known reserves of uranium would be consumed quite quickly by thermal reactors. It would be possible to sustain a large nuclear programme only by introducing fast reactors. One can conclude that there are strategic reasons for pursuing the development of fast breeder reactors. It will become desirably essential to have this technology available for introduction. The recycling of plutonium into LMFRs would allow 'burning' of the associated extremely long-life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. This additional important mission for the LMFR is gaining worldwide interest. In the framework of disarmament of nuclear weapons and the utilization of the nuclear material or peaceful purposes a role for fast reactors can be also considered. Over the past 29 years, the IAEA has actively encouraged and advocated international co-operation in Fast Breeder Reactor Technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1995, as reported at the 29th Annual

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

    Science.gov (United States)

    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.

  12. Uncertainty analysis for fuel flux calculations of fast reactors with external fuel cycle

    International Nuclear Information System (INIS)

    The paper focuses on the results of uncertainty analysis when calculating nuclide composition in fuel of fast reactors and on uncertainties of determining nuclide composition in the external fuel cycle. As demonstrated, the main contributions to the uncertainty of nuclide composition are due to: - uncertainties in operation of the reactor and in the fuel-cycle time; - uncertainties in nuclide clean-up factors at the Closed Nuclear Fuel Cycle (CNFC) stages when reprocessing spent nuclear fuel; - uncertainties in isotopic-kinetics cross-sections; - uncertainties in nuclide decay data. (author)

  13. Studies of the restructuring of fast breeder test reactor fuel by out-of-pile simulation

    International Nuclear Information System (INIS)

    The fast breeder test reactor (FBTR) at Kalpakkam, India, currently employs a mixed carbide of uranium and plutonium with a Pu/(Pu + U) ratio of 0.70 as fuel. The behavior of this fuel in a thermal gradient is investigated. An out-of-pile simulation facility is designed, set up, and commissioned. Experiments are conducted on FBTR fuel pellets to study the restructuring of the fuel at various levels of linear power and its cracking behavior in a thermal gradient. The results are discussed in terms of their significance for reactor operation

  14. Device for cooling the main vessel of a fast fission nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Debru, M.

    1984-10-16

    The annular space delimited by the main vessel and an internal shell is in communication with the zone of the reactor vessel, in which the cold primary liquid is located. The annular space delimited by the shell and by an internal shell is in communication with the lower part of the core via tubes. Thus, the cold primary liquid is injected into the space where it circulates from bottom to top, and flows into the space, where it circulates from top to bottom while at the same time cooling the main vessel. The invention applies, in particular, to fast fission nuclear reactors cooled by liquid sodium.

  15. The use of modern control to improve fast reactor safety and availability

    International Nuclear Information System (INIS)

    The development of metal fuel for fast reactors has created a new opportunity for innovation in plant control. By exploiting the superior neutronics properties of metal fuel, the reactivity burnup swing can be made zero, obviating the need for control rod reactivity addition over life. This result can be used to simplify elements of the plant control and protection system, the goal being improved plant safety and availability. These simplifications, however, depend on the development of methods for improved reactor inlet temperature control and reactivity feedback monitoring. The use of modern control theory to develop these methods is described. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Sodium fast reactor power monitoring using {sup 20}F tagging agent

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.; Normand, S. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, Centre de Saclay, 91191 Gif sur Yvette Cedex (France); Ban, G. [ENSICAEN, F-14050 Caen (France); Dumarcher, V.; Brau, H. P.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A. M.; Montagu, T.; Dautremer, T.; Barat, E.

    2009-07-01

    This work deals with the use of gamma spectrometry to monitor the fourth generation sodium fast reactor (SFR) power. Simulation part has shown that power monitoring in short response time and with high accuracy is possible measuring delayed gamma emitters produced in the liquid sodium. An experimental test is under preparation at French SFR Phenix experimental reactor to validate simulation studies. Physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as sodium velocity, atomic densities, neutron spectra and incident neutron cross-sections of fission reactions, and also sodium activation reactions producing gamma emitters. Then, a thermal hydraulic transfer function was used for taking into account primary sodium flow in our calculations. Gamma spectra were then determined by Monte-Carlo simulations. The experiment will be set during the reactor 'ultimate testing'. The Delayed Neutron Detection (DND) system cell has been chosen as the best available primary sodium sample for gamma power monitoring on Phenix reactor due to short sodium transit time from reactor core to measurement sample and homogenized sampling in the reactor hot pool. The main gamma spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The signal is then processed by a digital signal processing system (called Adonis) which always gives optimum answer even for high count rate and various time activity measurements. For power monitoring problematic, use of a short decay period gamma emitter as the {sup 20}F will allow to obtain a very fast response system without cumulative and flow distortion effects. These works introduce advantages and performances of this new power monitoring system for future SFR. (authors)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Sodium void reactivity effect influence on the prospective fast neutron reactor safety and concept approaches

    International Nuclear Information System (INIS)

    Full text: After the Chernobyl NPP accident,Soviet Union regulations have been provided by zero integral sodium void reactivity effect (SVRE) requirement for the BN-800 fast neutron reactor which was under final design stage at this time period. To comply with the requirement, a new established reactor core concept was developed. The concept quintessence is provision of an above-core sodium cavity. Sodium density reduction or sodium non-availability at the cavity results in negative reactivity. It was stated out that sodium cavity specifications could be selected in such a manner that zero integral void reactivity will be guaranteed after boiling-off of whole fuel assembly sodium inventory. At next study stages, the designers started to implement the requirement for next nuclear reactor generations. The prospective high-power sodium-cooled fast neutron reactors were designed to implement the reactor cores featured by zero sodium void reactivity effect. Nevertheless, the idea resulted in reducing the reactor core height and some hindrances in optimizing the high-power reactor core because of considerable size increase of delivery plate, rotating plug, decrease of the reactor control rod efficiency etc. The SVRE problem is persistent for the beyond the design basis accidents when considerable amount of the reactor sodium inventory could be boiled out. Despite of many study runs on the design basis accidents of the fast neutron reactors (including the sodium-cavity-provided reactors as well), the designers have not acquired some enough evident requirements for specifying the positive SVRE value. Besides, beyond the design basis accident probability resulting in large-scale in-reactor sodium boiling off is very small one. Accident sequences which could result in the scenario for extant and future reactors, they should include more than ten failures for passive/active actuation type items. The issue produces steadily some discussions on practicability of implementing