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Sample records for integral fast reactor

  1. The Integral Fast Reactor

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  2. Integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

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

  3. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

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

  4. Integral Fast Reactor concept

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  5. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-01-01

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

  6. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    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 ampersand D

  7. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two dramatic demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the Integral Fast Reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics and also makes possible a simplified closed fuel cycle and waste process improvements

  8. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the integral fast reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics also makes possible a simplified close fuel cycle and waste process improvements. The paper describes the IFR concept, the inherent safety, tests, and status of IFR development today

  9. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1985-01-01

    During the past two years, scientists from Argonne have developed an advanced breeder reactor with a closed self contained fuel cycle. The Integral Fast Reactor (IFR) is a new reactor concept, adaptable to a variety of designs, that is based on a fuel cycle radically different from the CRBR line of breeder development. The essential features of the IFR are metal fuel, pool layout, and pyro- and electro-reprocessing in a facility integral with the reactor plant. The IFR shows promise to provide an inexhaustible, safe, economic, environmentally acceptable, and diversion resistant source of nuclear power. It shows potential for major improvement in all of the areas that have led to concern about nuclear power

  10. The integral fast reactor concept

    International Nuclear Information System (INIS)

    Chang, Yoon I.; Marchaterre, J.F.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

  11. The Integral Fast Reactor concept

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  12. Integral fast reactor concept

    International Nuclear Information System (INIS)

    Chang, Y.I.; Marchaterre, J.F.; Sevy, R.H.

    1984-01-01

    Key features of the IFR consist of a pool-type plant arrangement, a metal fuel-based core design, and an integral fuel cycle with colocated fuel cycle facility. Both the basic concept and the technology base have been demonstrated through actual integral cycle operation in EBR-II. This paper discusses the inherent safety characteristics of the IFR concept

  13. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    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

  14. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    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: a pool-type primary system, and advanced ternary alloy metallic fuel, and 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 the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and 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

  15. Advances by the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  16. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1989-01-01

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

  17. Fission energy: The integral fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1989-01-01

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

  18. Integral data for fast reactors

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  19. Integral fast reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  20. Integral Fast Reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative 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. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  1. The Integral Fast Reactor (IFR) concept

    International Nuclear Information System (INIS)

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

    1989-01-01

    In addition to maintaining the viability of its present commercial nuclear technology, a principal challenge in the US in the 1990s and beyond will be to regain and maintain a position among the world leadership in advanced reactor research and development. In this paper we'll discuss factors which we believe should today provide the rationale and focus for advanced reactor R and D, and we will then review the status of the major US effort, the Integral Fast Reactor (IFR) program

  2. Integral fast reactor shows its mettle

    International Nuclear Information System (INIS)

    Chang, Ya.; Lajnberri, M.; Barris, L.; Uoters, L.

    1988-01-01

    The main aspects of the problem of developing a closed fuel cycle at a NPP built in the so-called integrated version when a fast reactor and the plant for spent fuel regeneration and fuel element production are located in the same site (IFR project), are considered. The technologies of U-Pu-Zr alloy fuel reprocessing and production based on high-temperature metallurgical process and the method of casting under pressure are described. The demonstration of practical feasibility of the fuel cycle on the basis of the IFR reactor is planned for 1990

  3. The integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1990-01-01

    The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management

  4. The integral fast reactor - an overview

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.; Hannum, W.H.

    1997-01-01

    The Integral Fast Reactor (IFR) is a system that consists of a fast-spectrum nuclear reactor that uses metallic fuel and liquid-metal (sodium) cooling, coupled with technology for high-temperature electrochemical recycling, and with processes for preparing wastes for disposition. The concept is based on decades of experience with fast reactors, adapted to priorities that have evolved markedly from those of the early days of nuclear power. It has four essential, distinguishing features: efficient use of natural resources, inherent safety characteristics, reduced burdens of nuclear waste, and unique proliferation resistance. These fundamental characteristics offer benefits in economics and environmental protection. The fuel cycle never involves separated plutonium, immediately simplifying the safeguarding task. Initiated in 1984 in response to proliferation concerns identified in the International Nuclear Fuel Cycle Evaluation (INFCE, 1980), the project has made substantial technical progress, with new potential applications coming to light as nuclear weapons stockpiles are reduced and concerns about waste disposal increase. A breakthrough technology, the IFR has the characteristics necessary for the next nuclear age. (author)

  5. Integral Fast Reactor fuel pin processor

    International Nuclear Information System (INIS)

    Levinskas, D.

    1993-01-01

    This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves

  6. Actinide behavior in the integral fast reactor

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1993-05-01

    Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory's site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core

  7. Actinide burning in the integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1993-01-01

    During the past few years, Argonne National Laboratory has been developing the integral fast reactor (IFR), an advanced liquid-metal reactor concept. In the IFR, the inherent properties of liquid-metal cooling are combined with a new metallic fuel and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics, and waste management. A key feature of the IFR concept is its unique pyroprocessing. Pyroprocessing has the potential to radically improve long-term waste management strategies by exploiting the following attributes: 1. Minor actinides accompany plutonium product stream; therefore, actinide recycling occurs naturally. Actinides, the primary source of long-term radiological toxicity, are removed from the waste stream and returned to the reactor for in situ burning, generating useful energy. 2. High-level waste volume from pyroprocessing call be reduced substantially as compared with direct disposal of spent fuel. 3. Decay heat loading in the repository can be reduced by a large factor, especially for the long-term burden. 4. Low-level waste generation is minimal. 5. Troublesome fission products, such as 99 Tc, 129 I, and 14 C, are contained and immobilized. Singly or in combination, the foregoing attributes provide important improvements in long-term waste management in terms of the ease in meeting technical performance requirements (perhaps even the feasibility of demonstrating that technical performance requirements can be met) and perhaps also in ultimate public acceptance. Actinide recycling, if successfully developed, could well help the current repository program by providing an opportunity to enhance capacity utilization and by deferring the need for future repositories. It also represents a viable technical backup option in the event unforeseen difficulties arise in the repository licensing process

  8. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-01-01

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  9. Looking to the future with the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.

    1985-01-01

    During the past two years, scientists from Argonne have developed a design for an advanced breeder reactor with a closed, self-contained fuel cycle. This Integral Fast Reactor (IFR) is a pool-type, sodium-cooled reactor. It uses a new metal-alloy fuel design which overcomes the problem of swelling. The possibility of unauthorised diversion of nuclear fuel, and the need to transport plutonium to and from the site, is overcome by using a pyrometallurgical fuel reprocessing technique in a compact facility that is an integral part of the reactor plant. (author)

  10. Integrating the fuel cycle at IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

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

    1992-01-01

    During the past few years Argonne National Laboratory has been developing the Integral Fast Reactor (IFR), an advanced liquid metal reactor. Much of the IFR technology stems from Argonne National Laboratory's experience with the Experimental Breeder Reactors, EBR 1 and 2. The unique aspect of EBR 2 is its success with high-burnup metallic fuel. Irradiation tests of the new U-Pu-Zr fuel for the IFR have now reached a burnup level of 20%. The results to date have demonstrated excellent performance characteristics of the metallic fuel in both steady-state and off-normal operating conditions. EBR 2 is now fully loaded with the IFR fuel alloys and fuel performance data are being generated. In turn, metallic fuel becomes the key factor in achieving a high degree of passive safety in the IFR. These characteristics were demonstrated dramatically by two landmark tests conducted at EBR 2 in 1986: loss of flow without scram; and loss of heat sink without scram. They demonstrated that the combination of high heat conductivity of metallic fuel and thermal inertia of the large sodium pool can shut the reactor down during potentially severe accidents without depending on human intervention or the operation of active engineered components. The IFR metallic fuel is also the key factor in compact pyroprocessing. Pyroprocessing uses high temperatures, molten salt and metal solvents to process metal fuels. The result is suitable for fabrication into new fuel elements. Feasibility studies are to be conducted into the recycling of actinides from light water reactor spent fuel in the IFR using the pyroprocessing approach to extract the actinides (author)

  11. The safety basis of the integral fast reactor program

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Seidel, B.R.

    1990-01-01

    The Integral Fast Reactor (IFR) and metallic fuel have emerged as the US Department of Energy reference reactor concept and fuel system for the development of an advanced liquid-metal reactor. This article addresses the basic elements of the IFR reactor concept and focuses on the safety advances achieved by the IFR Program in the areas of (1) fuel performance, (2) superior local faults tolerance, (3) transient fuel performance, (4) fuel-failure mechanisms, (5) performance in anticipated transients without scram, (6) core-melt mitigation, and (7) actinide recycle

  12. Waste management in IFR [Integral Fast Reactor] fuel cycle

    International Nuclear Information System (INIS)

    Johnson, T.R.; Battles, J.E.

    1991-01-01

    The fuel cycle of the Integral Fast Reactor (IFR) has important potential advantage for the management of high-level wastes. This sodium-cooled, fast reactor will use metal fuels that are reprocessed by pyrochemical methods to recover uranium, plutonium, and the minor actinides from spent core and blanket fuel. More than 99% of all transuranic (TRU) elements will be recovered and returned to the reactor, where they are efficiently burned. The pyrochemical processes being developed to treat the high-level process wastes are capable of producing waste forms with low TRU contents, which should be easier to dispose of. However, the IFR waste forms present new licensing issues because they will contain chloride salts and metal alloys rather than glass or ceramic. These fuel processing and waste treatment methods can also handle TRU-rich materials recovered from light-water reactors and offer the possibility of efficiently and productively consuming these fuel materials in future power reactors

  13. Actinide recycle potential in the integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. In the IFR pyroprocessing, minor actinides accompany plutonium product stream, and therefore, actinide recycle occurs naturally. The fast neutron spectrum of the IFR makes it an ideal actinide burner, as well. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and potential implications on long-term waste management

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

    International Nuclear Information System (INIS)

    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 ampersand D

  15. 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.

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

    International Nuclear Information System (INIS)

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

    1994-12-01

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

  17. 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.

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

    International Nuclear Information System (INIS)

    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

  19. Integral Fast Reactor: A future source of nuclear energy

    International Nuclear Information System (INIS)

    Southon, R.

    1993-01-01

    Argonne National Laboratory is developing a reactor concept that would be an important part of the worlds energy future. This report discusses the Integral Fast Reactor (IFR) concept which provides significant improvements over current generation reactors in reactor safety, plant complexity, nuclear proliferation, and waste generation. Two major facilities, a reactor and a fuel cycle facility, make up the IFR concept. The reactor uses fast neutrons and metal fuel in a sodium coolant at atmospheric pressure that relies on laws of physics to keep it safe. The fuel cycle facility is a hot cell using remote handling techniques for fabricating reactor fuel. The fuel feed stock includes spent fuel from the reactor, and potentially, spent light water reactor fuel and plutonium from weapons. This paper discusses the unique features of the IFR concept and the differences the quality assurance program has from current commercial practices. The IFR concept provides an opportunity to design a quality assurance program that makes use of the best contemporary ideas on management and quality

  20. Preparations for the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.

    1989-01-01

    Modifications to the Hot Fuel Examination Facility-South (HFEF/S) have been in progress since mid-1988 to ready the facility for demonstration of the unique Integral Fast Reactor (IFR) pyroprocess fuel cycle. This paper updates the last report on this subject to the American Nuclear Society and describes the progress made in the modifications to the facility and in fabrication of the new process equipment. The IFR is a breeder reactor, which is central to the capability of any reactor concept to contribute to mitigation of environmental impacts of fossil fuel combustion. As a fast breeder, fuel of course must be recycled in order to have any chance of an economical fuel cycle. The pyroprocess fuel cycle, relying on a metal alloy reactor fuel rather than oxide, has the potential to be economical even at small-scale deployment. Establishing this quantitatively is one important goal of the IFR fuel cycle demonstration

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

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1993-01-01

    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

  2. Integral measurement of fission products capture in fast breeder reactors

    International Nuclear Information System (INIS)

    Martin Deidier, Loick.

    1979-12-01

    For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set [fr

  3. Safety characteristics of the integral fast reactor concept

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Cahalan, J.E.; Sevy, R.H.; Wright, A.E.

    1985-01-01

    The Integral Fast Reactor (IFR) concept is an innovative approach to liquid metal reactor design which is being studied by Argonne National Laboratory. Two of the key features of the IFR design are a metal fuel core design, based on the fuel technology developed at EBR-II, and an integral fuel cycle with a colocated fuel cycle facility based on the compact and simplified process steps made possible by the use of metal fuel. The paper presents the safety characteristics of the IFR concept which derive from the use of metal fuel. Liquid metal reactors, because of the low pressure coolant operating far below its boiling point, the natural circulation capability, and high system heat capacities, possess a high degree of inherent safety. The use of metallic fuel allows the reactor designer to further enhance the system capability for passive accommodation of postulated accidents

  4. The design rationale of the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Wade, D.C.; Hill, R.N.

    1997-01-01

    The Integral Fast Reactor (IFR) concept has been developed over the last ten years to provide technical solutions to perceptual concerns associated with nuclear power. Beyond the traditional advanced reactor objectives of increased safety, improved economy and more efficient fuel utilization, the IFR is designed to simplify waste disposal and increase resistance to proliferation. Only a fast reactor with an efficient recycle technology can provide for total consumption of actinides. The basic physics governing reactor design dictates that, for efficient recycle, the fuel form should be limited in burnup only by radiation damage to fuel cladding. The recycle technology must recover essentially all actinides. In a fast reactor, not all fission products need to be removed from the recycled fuel, and there is no need to produce pure plutonium. Recovery, recycle, and ultimate consumption of all actinides resolves several waste-disposal concerns. The IFR can be configured to achieve safe passive response to any of the traditional postulated reactor accident initiators, and can be configured for a variety of power output levels. Passive heat removal is achieved by use of a large inventory sodium coolant and a physical configuration that emphasizes natural circulation. An IFR can be designed to consume excess fissile material, to produce a surplus, or to maintain inventory. It appears that commercial designs should be economically competitive with other available alternatives. (author)

  5. Technical feasibility of an Integral Fast Reactor (IFR) as a future option for fast reactor cycles. Integrate a small metal-fueled fast reactor and pyroprocessing facilities

    International Nuclear Information System (INIS)

    Tanaka, Nobuo

    2017-01-01

    Integral Fast Reactor that integrated fast reactor and pyrorocessing facilities developed by Argonne National Laboratory in the U.S. is an excellent nuclear fuel cycle system for passive safety, nuclear non-proliferation, and reduction in radioactive waste. In addition, this system can be considered as a technology applicable to the treatment of the fuel debris caused by the Fukushima Daiichi Nuclear Power Station accident. This study assessed the time required for debris processing, safety of the facilities, and construction cost when using this technology, and examined technological possibility including future technological issues. In a small metal-fueled reactor, it is important to design the core that achieves both of reduction in combustion reactivity and reduction in coolant reactivity. In system design, calorimetric analysis, structure soundness assessment, seismic feasibility establishment study, etc. are important. Regarding safety, research and testing are necessary on the capabilities of passive reactor shutdown and reactor core cooling as well as measures for avoiding re-criticality, even when emergency stop has failed. In dry reprocessing system, studies on electrolytic reduction and electrolytic refining process for treating the debris with compositions different from those of normal fuel are necessary. (A.O.)

  6. Safeguards operations in the integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Goff, K.M.; Benedict, R.W.; Brumbach, S.B.; Dickerman, C.E.; Tompot, R.W.

    1994-01-01

    Argonne National Laboratory is currently demonstrating the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that 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 IFR fuel cycle employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The safeguards aspects of the fuel cycle demonstration must be approved by the United States Department of Energy, but a further goal of the program is to develop a safeguards system that could gain acceptance from the Nuclear Regulatory Commission and International Atomic Energy Agency. This fuel cycle is described with emphasis on aspects that differ from aqueous reprocessing and on its improved safeguardability due to decreased attractiveness and diversion potential of all process streams, including the fuel product

  7. Proposed fuel cycle for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.; Walters, L.C.

    1985-01-01

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

  8. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. 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, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  9. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. 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, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  10. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. 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, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  11. Interim waste storage for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

    1991-01-01

    The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes that are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig

  12. 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.

  13. Actinide behavior in the Integral Fast Reactor. Final project report

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C.

    1994-11-01

    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 weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II 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.

  14. Actinide behavior in the Integral Fast Reactor. Final project report

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1994-11-01

    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 ( 237 Np, 240 Pu, 241 Am, and 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 weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II 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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

    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 ( 237 Np, 240 Pu, 241 Am, and 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

  16. Advanced reactor development: The LMR integral fast reactor program at Argonne

    International Nuclear Information System (INIS)

    Till, C.E.

    1990-01-01

    Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  18. Integral test of JENDL-3.3 on fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou; Hazama, Taira

    2003-05-01

    An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

  19. Progress and status of the integral fast reactor (IFR) development program

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    This paper discusses the Integral Fast Reactor (IFR) development program, in which the entire reactor system - reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are also presented

  20. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  1. Pyrometallurgical processing of Integral Fast Reactor metal fuels

    International Nuclear Information System (INIS)

    Battles, J.E.; Miller, W.E.; Gay, E.C.

    1991-01-01

    The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500 degrees C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way

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

    International Nuclear Information System (INIS)

    Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

    1991-01-01

    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

  3. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  4. Integrated fuel-cycle models for fast breeder reactors

    International Nuclear Information System (INIS)

    Ott, K.O.; Maudlin, P.J.

    1981-01-01

    Breeder-reactor fuel-cycle analysis can be divided into four different areas or categories. The first category concerns questions about the spatial variation of the fuel composition for single loading intervals. Questions of the variations in the fuel composition over several cycles represent a second category. Third, there is a need for a determination of the breeding capability of the reactor. The fourth category concerns the investigation of breeding and long-term fuel logistics. Two fuel-cycle models used to answer questions in the third and fourth area are presented. The space- and time-dependent actinide balance, coupled with criticality and fuel-management constraints, is the basis for both the Discontinuous Integrated Fuel-Cycle Model and the Continuous Integrated Fuel-Cycle Model. The results of the continuous model are compared with results obtained from detailed two-dimensional space and multigroup depletion calculations. The continuous model yields nearly the same results as the detailed calculation, and this is with a comparatively insignificant fraction of the computational effort needed for the detailed calculation. Thus, the integrated model presented is an accurate tool for answering questions concerning reactor breeding capability and long-term fuel logistics. (author)

  5. Nonproliferation and safeguards aspects of the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Hannum, W.H.; Wade, D.C.; McFarlane, H.F.; Hill, R.N.

    1997-01-01

    The National Academy of Sciences (NAS) has declared that the large and growing stocks of plutonium from weapons dismantlement in the U.S. and the former Soviet Union FSU are a ''clear and present danger'' to peace and security. Moreover, the opinion of some experts that plutonium of any isotopic blend is a proliferation threat has been well publicized, heightening the concern that plutonium produced in the civilian fuel cycle is itself a proliferation threat. Assuring that separated plutonium, from dismantled warheads as well as from civilian power programs, is under effective control has (again) become a high priority of U.S. diplomacy. One pole of the debate on how to manage this material is to declare it to be a waste, and to search for some way to dispose of it safely, securely, and permanently. The other pole is to view it as an energy resource and to safeguard it against diversion, putting it into active use in the civilian power program. The ultimate choice cannot be separated from the long-term strategy for use of peaceful nuclear power. Continued use of a once-through fuel cycle will lead to an ever-increasing quantity of excess plutonium-requiring safeguarding. Alternatively, recycling the world's stocks of plutonium in fast reactors, contrary to common misconception, will cap the world supply of plutonium and hold it in working inventories for generating power. Transition from the current-generation light water cooled reactors (LWRs) to a future fast-reactor-based nuclear energy supply under international safeguards would, henceforth, limit world plutonium inventories to the amount necessary and useful for power generation, with no further excess production. (author)

  6. The passive response of the Integral Fast Reactor concept to the chilled inlet accident

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1990-01-01

    Simple methods are described for bounding the passive response of a metal fueled liquid-metal cooled reactor to the chilled inlet accident. Calculation of these bounds for a prototype of the Integral Fast Reactor concept shows that failure limits --- eutectic melting, sodium boiling and fuel pin failure --- are not exceeded. 2 refs., 1 fig., 2 tabs

  7. Progress and status of the Integral Fast Reactor (IFR) development program

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1992-01-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The ALMR reactor plant design is being developed by an industrial team headed by General Electric and is presented in a companion paper. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are presented in the other two companion papers that follows this

  8. Progress and status of the Integral Fast Reactor (IFR) development program

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1992-04-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The ALMR reactor plant design is being developed by an industrial team headed by General Electric and is presented in a companion paper. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are presented in the other two companion papers that follows this.

  9. Progress and status of the Integral Fast Reactor (IFR) development program

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1992-01-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The ALMR reactor plant design is being developed by an industrial team headed by General Electric and is presented in a companion paper. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are presented in the other two companion papers that follows this.

  10. Integral test of JENDL dosimetry file using fast neutron field in the Experimental Fast Reactor JOYO

    International Nuclear Information System (INIS)

    Aoyama, Takafumi; Sekine, Takashi

    1999-09-01

    In order to evaluate the applicability of the JENDL dosimetry file, an integral test using a fast neutron spectrum field in the Experimental Fast Reactor JOYO Mark-II core was performed. The dosimeter set consisting of eight reactions of 46 Ti(n,p) 46 Sc, 54 Fe(n,p) 54 Mn, 58 Fe(n,γ) 59 Fe, 58 Ni(n,p) 58 Co, 59 Co(n,γ) 60 Co, 63 Cu(n,α) 60 Co, 238 U fission and 237 Np fission was irradiated for approximately 30 days near the core center of the JOYO Mk-II. Neutron flux at the dosimeter position was calculated using the two dimensional discrete ordinate transport code 'DORT'. The core configuration was modeled in XY geometry, and the 100 group cross section set of JSD-J2 / JFT-J2, which was processed from JENDL-2, was utilized. The absolute value of neutron flux was normalized so that the 235 U fission rate using the calculated neutron spectrum agreed with the measured reaction rate. The 103 group cross section data were processed by 'NJOY' code for nuclides to be used in the JOYO dosimetry. As the results of integral test for JENDL/D-99 (new file) and JENDL/D-91 (previous file), calculated values by JENDL/D-99 agreed well with the experimental values, and the C/E ratios ranged from 0.95 to 1.22. By comparing the results between JENDL/D-99 and JENDL/D-91, small differences exist, except for 58 Fe(n, γ) 59 Fe reaction, which was improved significantly in JENDL/D-99. (author)

  11. The integral fast reactor fuels reprocessing laboratory at Argonne National Laboratory, Illinois

    International Nuclear Information System (INIS)

    Wolson, R.D.; Tomczuk, Z.; Fischer, D.F.; Slawecki, M.A.; Miller, W.E.

    1986-09-01

    The processing of Integral Fast Reactor (IFR) metal fuel utilizes pyrochemical fuel reprocessing steps. These steps include separation of the fission products from uranium and plutonium by electrorefining in a fused salt, subsequent concentration of uranium and plutonium for reuse, removal, concentration, and packaging of the waste material. Approximately two years ago a facility became operational at Argonne National Laboratory-Illinois to establish the chemical feasibility of proposed reprocessing and consolidation processes. Sensitivity of the pyroprocessing melts to air oxidation necessitated operation in atmosphere-controlled enclosures. The Integral Fast Reactor Fuels Reprocessing Laboratory is described

  12. Sodium tests on an integrated purification prototype for a sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Abramson, R.

    1984-04-01

    This paper describes sodium tests performed on the integrated primary sodium purification prototype of the Creys Malville Super Phenix 1 fast breeder reactor. These tests comprised: - hydrostatic test, - thermal tests, - handling tests. They enabled a number of new technological arrangements to be qualified and provided the necessary information for the design and construction of the Super Phenix 1 purification units

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

    International Nuclear Information System (INIS)

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

    1993-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

    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. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-03-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-01-01

    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.

  19. Some aspects affecting fast reactor steam generator integrity considered from a utility viewpoint

    Energy Technology Data Exchange (ETDEWEB)

    Bolt, P R

    1975-07-01

    The important conditions affecting fast reactor steam generator integrity are discussed. In addition to the need for high integrity levels when the steam generator is first delivered to the power station site, the equally important aspect of demonstrating retention of continued high levels of integrity throughout the operating life of the station is described. The functional and related conditions that are believed important to the selection of a design type which can offer adequately high levels of integrity are given. Some of the data needs of a utility concerned with fast reactor S.G.U. design assessment are described, particular emphasis being given to areas believed to have a significant effect on steam generator reliability and integrity. (author)

  20. Fast reactors in nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkovskii, O

    1981-02-01

    The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

  1. Proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the integral fast reactor

    International Nuclear Information System (INIS)

    Burris, L.; Steindler, M.; Miller, W.

    1984-01-01

    The pool-type Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps: a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

  2. A proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.; Steindler, M.; Miller, W.

    1984-01-01

    The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps -- a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

  3. Analysis of unprotected overcooling events in the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1989-01-01

    Simple analytic models are developed for predicting the response of a metal fueled, liquid-metal cooled reactor to unprotected overcooling events in the balance of plant. All overcooling initiators are shown to fall into two categories. The first category contains these events for which there is no final equilibrium state of constant overcooling, as in the case for a large steam leak. These events are analyzed using a non-flow control mass approach. The second category contains those events which will eventually equilibrate, such as a loss of feedwater heaters. A steady flow control volume analysis shows that these latter events ultimately affect the plant through the feedwater inlet to the steam generator. The models developed for analyzing these two categories provide upper bounds for the reactor's passive response to overcooling accident initiators. Calculation of these bounds for a prototypic plant indicate that failure limits -- eutectic melting, sodium boiling, fuel pin failure -- are not exceeded in any overcooling event. 2 refs

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

    International Nuclear Information System (INIS)

    Johnson, T.R.

    1986-01-01

    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

  5. Adjustement of multigroup cross sections using fast reactor integral data

    International Nuclear Information System (INIS)

    Renke, C.A.C.

    1982-01-01

    A methodology for the adjustment of multigroup cross section is presented, structured with aiming to compatibility the limitated number of measured values of integral parameters known and disponible, and the great number of cross sections to be adjusted the group of cross section used is that obtained from the Carnaval II calculation system, understanding as formular the sets of calculation methods and data bases. The adjustment is realized, using the INCOAJ computer code, developed in function of one statistical formulation, structural from the bayer considerations, taking in account the measurement processes of cross section and integral parameters defined on statistical bases. (E.G.) [pt

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

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

    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

  7. Integrity of Safety-Related Fast Reactor Structures

    International Nuclear Information System (INIS)

    Rose, R.T.; Tomkins, B.

    1981-01-01

    The LMFBR contains several structural items whose integrity must be safeguarded during the life of the plant. These items include the main core support structures (strongback, diagrid) and the primary tank to which these structures are attached. In order to demonstrate an acceptable level of structural integrity, the chosen design philosophy must be supported by both analytical and experimental evidence. This paper describes the current approaches in the UK to these requirements. Section 2 describes the materials mechanical properties tests performed to date on both fracture toughness and fatigue crack growth in Type 316 austenitic stainless steel plate and weldments. This data illustrates the problems in identifying the relevant materials fracture parameters for use in assessments. Section 3 shows the test programmes in hand to extend the materials programmes to tests on structural features (mainly welded wide plate tests) which incorporate the complexity of weldments in a structural context. This includes experimental evidence on the effects of local weld residual stresses on structural failure. Various routes are open for the integrity assessment of FR structures. These are discussed in Section 4 but in effect they reduce to a fracture mechanics approach using some technique to cope with elastic-plastic fracture. The main problems at present relate to our ability in analysis to cope with residual stresses and the post-initiation region of the fracture resistance curve. Also, there is the problem of initial defect sizing by current NDE techniques. Current conservative analytical assessments give acceptable defect sizes of order a few millimetres in irradiated weldments. Finally, Section 5 discusses the options open in design to cope with safety related structures under normal and abnormal loading conditions. It is clear that several options exist in design to satisfy the demand for high integrity

  8. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

    The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. Actinide recycle potential in the integral fast reactor (IFR) fuel cycle

    International Nuclear Information System (INIS)

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

    1991-01-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The use of metallic fuel in the IFR allows a radically improved fuel cycle technology. Based on the recent IFR process development, a preliminary assessment has been made to investigate the feasibility of further adapting pyrochemical processes to directly extract actinides from LWR spent fuel. The results of this assessment indicate very promising potential and two most promising flowsheet options have been identified for further research and development. This paper also summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs

  11. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  12. Progress and status of the integral fast reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

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

    1993-01-01

    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. The Integral Fast Reactor (IFR) fuel cycle, is based on the use of a metallic fuel alloy (U-Pu-Zr) that permits use of an innovative method for processing of spent fuel. This method, a combination of pyrometallurgical and electrochemical processes, has been termed pyroprocessing. It offers the advantages of a simple, compact processing system and limited volumes of stabilized high-level wastes. This translates to an economically viable system that is likely to receive favorable public response, particularly when combined with the other attributes of the Integral Fast Reactor. Substantial progress has been made in the development of the IFR pyroprocessing method. A comprehensive demonstration of the process will soon begin at the Argonne National Laboratory Idaho site, using spent fuel from the EBR-II reactor. An important advantage of the IFR is its ability to recycle fuel in the process of power generation, extending fuel resources by a considerable amount and assuring the continued viability of nuclear power stations by reducing dependence on external fuel supplies. Pyroprocessing is the means whereby the recycle process is accomplished. It can also be applied to the recovery of fuel constituents from spent fuel generated in the process of operation of conventional light water reactor power plants, offering the means to recover the valuable fuel resources remaining in that material

  13. The fast reactor

    International Nuclear Information System (INIS)

    1980-02-01

    The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

  14. 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.

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

    International Nuclear Information System (INIS)

    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

  16. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  17. Fast breeder reactor research

    International Nuclear Information System (INIS)

    1975-01-01

    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

  18. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in uranium-based metal alloys as a fuel for sodium-cooled fast reactors. In this paper we will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel. In the final section of this paper we extend the calculations to consider the failure of IFR ternary fuel under reactor accident conditions. (orig./GL)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  20. Fuel motion in overpower tests of metallic integral fast reactor fuel

    International Nuclear Information System (INIS)

    Rhodes, E.A.; Bauer, T.H.; Stanford, G.S.; Regis, J.P.; Dickerman, C.E.

    1992-01-01

    In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure

  1. Status of the Integral Fast Reactor fuel cycle demonstration and waste management practices

    International Nuclear Information System (INIS)

    Benedict, R.W.; Goff, K.M.; McFarlane, H.F.

    1994-01-01

    Over the past few years, Argonne National Laboratory has been preparing for the demonstration of the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety and operations, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle, which will be demonstrated at Argonne-West in Idaho, employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The required facility modifications and process equipment for the demonstration are nearing completion. Their status and the results from initial fuel fabrication work, including the waste management aspects, are presented. Additionally, estimated compositions of the various process waste streams have been made, and characterization and treatment methods are being developed. The status of advanced waste processing equipment being designed and fabricated is described

  2. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  3. Physics studies of weapons plutonium disposition in the Integral Fast Reactor closed fuel cycle

    International Nuclear Information System (INIS)

    Hill, R.N.; Wade, D.C.; Liaw, J.R.; Fujita, E.K.

    1995-01-01

    The core performance impact of weapons plutonium introduction into the Integral Fast Reactor (IFR) closed fuel cycle is investigated by comparing three disposition scenarios: a power production mode, a moderate destruction mode, and a maximum destruction mode, all at a constant heat rating of 840 MW(thermal). For each scenario, two fuel cycle models are evaluated: cores using weapons material as the sole source of transuranics in a once-through mode and recycle cores using weapons material only as required for a makeup feed. In addition, the impact of alternative feeds (recycled light water reactor or liquid-metal reactor transuranics) on burner core performance is assessed. Calculated results include mass flows, detailed isotopic distributions, neutronic performance characteristics, and reactivity feedback coefficients. In general, it is shown that weapons plutonium does not have an adverse effect on IFR core performance characteristics; also, favorable performance can be maintained for a wide variety of feed materials and fuel cycle strategies

  4. A status report on the integral fast reactor fuels and safety program

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Seidel, B.R.

    1990-01-01

    The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor (ALMR) concept being developed at Argonne National Laboratory. The IFR program is specifically responsible for the irradiation performance, advanced core design, safety analysis, and development of the fuel cycle for the US Department of Energy's ALMR program. The basic elements of the IFR concept are (a) metallic fuel, (b) liquid-sodium cooling, (c) modular, pool-type reactor configuration, (d) an integral fuel cycle based upon pyrometallurgical processing. The most significant safety aspects of the IFR program result from its unique fuel design, a ternary alloy of uranium, plutonium, and zirconium. This fuel is based on experience gained through > 25 yr operation of the Experimental Breeder Reactor II (EBR-II) with a uranium alloy metallic fuel. The ultimate criteria for fuel pin design is the overall integrity at the target burnup. The probability of core meltdown is remote; however, a theoretical possibility of core meltdown remains. The next major step in the IFR development program will be a full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. The IFR fuel cycle closure based on pyroprocessing will also have a dramatic impact on waste management options and on actinide recycling

  5. The fast breeder reactor

    International Nuclear Information System (INIS)

    Davis, D.A.; Baker, M.A.W.; Hall, R.S.

    1990-01-01

    Following submission of written evidence, the Energy Committee members asked questions of three witnesses from the Central Electricity Generating Board and Nuclear Electric (which will be the government owned company running nuclear power stations after privatisation). Both questions and answers are reported verbatim. The points raised include where the responsibility for the future fast reactor programme should lie, with government only or with private enterprise or both and the viability of fast breeder reactors in the future. The case for the fast reactor was stated as essentially strategic not economic. This raised the issue of nuclear cost which has both a construction and a decommissioning element. There was considerable discussion as to the cost of building a European Fast reactor and the cost of the electricity it would generate compared with PWR type reactors. The likely demand for fast reactors will not arrive for 20-30 years and the need to build a fast reactor now is questioned. (UK)

  6. Mass tracking and material accounting in the integral fast reactor (IFR)

    International Nuclear Information System (INIS)

    Orechwa, Y.; Adams, C.H.; White, A.M.

    1991-01-01

    This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations

  7. Mass tracking and material accounting in the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Orechwa, Y.; Adams, C.H.; White, A.M.

    1991-01-01

    The Integral Fast Reactor (IFR) is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory (ANL). There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure the compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities at ANL-West, utilizing Experimental Breeder Reactor 2 and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-Tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations. The components of the MTG System include: (1) an Oracle database manager with a Fortran interface, (2) a set of MTG ''Tasks'' which collect, manipulate and report data, (3) a set of MTG ''Terminal Sessions'' which provide some interactive control of the Tasks, and (4) a set of servers which manage the Tasks and which provide the communications link between the MTG System and Operator Control Stations, which control process equipment and monitoring devices within the FCF

  8. European R and D co-ordinate programme on structural integrity of fast breeder reactors

    International Nuclear Information System (INIS)

    Hoffmann, A.; Combescure, A.; Acker, D.; Corsi, F.; Martelli, A.; Vinzens, K.; Angerbauer, A.

    1989-01-01

    After a period of development of medium size prototype plants, PFR in United Kingdom, Phenix in France and SNR 300 in West Germany, and the build up of the large size prototype plant Super Phenix in France by the European Consortium, NERSA, Belgium, France, Italy, United Kingdom and West Germany decided to join their efforts in order to pursue the development of LMFBR nuclear power plants. This paper presents the European Research and Development coordinated program in the field of structural integrity of fast breeder reactors with its organization, its objectives, its programs and the resources allocated for development. (author)

  9. Fast reactors worldwide

    International Nuclear Information System (INIS)

    Hall, R.S.; Vignon, D.

    1985-01-01

    The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

  10. Actinide behavior in the integral fast reactor. Progress report, May 1, 1992--April 30, 1993

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C.

    1993-05-01

    Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory`s site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core.

  11. A risk characterization of safety research areas for integral fast reactor program planning

    International Nuclear Information System (INIS)

    Mueller, C.J.; Cahalan, J.E.; Hill, D.J.; Kramer, J.M.; Marchaterre, J.F.; Pedersen, D.R.; Sevy, R.H.; Tibbrook, R.W.; Wei, T.Y.; Wright, A.E.

    1988-01-01

    This paper characterizes the areas of integral fast reactor (IFR) safety research in terms of their importance in addressing the risk of core disruption sequences for innovative designs. Such sequences have traditionally been determined to constitute the primary risk to public health and safety. All core disruption sequences are folded into four fault categories: classic unprotected (unscrammed) events; loss of decay heat; local fault propagation; and failure to critical reactor structures. Event trees are used to describe these sequences and the areas in the IFR safety and related base technology research programs are discussed with respect to their relevance in addressing the key issues in preventing or delimiting core disruptive sequences. Thus a measure of potential for risk reduction is obtained for guidance in establishing research priorities

  12. A risk characterization of safety research areas for Integral Fast Reactor program planning

    International Nuclear Information System (INIS)

    Mueller, C.J.; Cahalan, J.E.; Hill, D.J.

    1988-01-01

    This paper characterizes the areas of Integral Fast Reactor (IFR) safety research in terms of their importance in addressing the risk of core disruption sequences for innovative designs. Such sequences have traditionally been determined to constitute the primary risk to public health and safety. All core disruption sequences are folded into four fault categories: classic unprotected (unscrammed) events; loss of decay heat; local fault propagation; and failure of critical reactor structures. Event trees are used to describe these sequences and the areas in the IFR Safety and related Base Technology research programs are discussed with respect to their relevance in addressing the key issues in preventing or delimiting core disruptive sequences. Thus a measure of potential for risk reduction is obtained for guidance in establishing research priorites

  13. Using the Integral Fast Reactor (IFR) to dispose of excess weapons plutonium

    International Nuclear Information System (INIS)

    Hannum, W.H.; Wade, D.C.

    1997-01-01

    Plutonium is a man-made radioactive element with a long half-life. The only way to dispose of plutonium permanently is by causing it to fission. The fission process is efficient only in a fast neutron spectrum, and multiple recycle with a minimal loss is required to approach complete destruction. To be consistent with nonproliferation objectives, the process should be compatible with rigorous safeguards, and should not involve handling separated plutonium; The Integral Fast Reactor (IFR) meets all of these requirements. In addition, several near-term denaturing options are available that are fully compatible with complete destruction. When coupled with electrical generation, ample revenues would be available to cover all handling, operating and safeguards costs, with a substantial residual net return on the investment. (author)

  14. 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...

  15. Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H.; Ackerman, J.P.

    1993-01-01

    This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled 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. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core

  16. Measurements and calculations of integral capture cross-sections of structural materials in fast reactor spectra

    International Nuclear Information System (INIS)

    Seth, S.; Brunson, G.; Gmuer, K.; Jermann, M.; McCombie, C.; Richmond, R.; Schmocker, U.

    1979-01-01

    This paper relates the checking of integral data of steel and iron in fast reactor lattices. The fully-rodded GCFR benchmark lattice of the zero-energy reactor PROTEUS was successively modified by replacing the PuO 2 -UO 2 fuel rods by steel-18/8 or steel-37 (iron) rods. The neutron spectra of the modified lattices in fact have median energies close to that of a typical LMFBR. The replacement of fuel by the structural material of interest was such that in each case the value of k(infinity) was reduced to near-unity. This allowed the measurement of the lattice-k(infinity) by the null-reactivity technique. In addition, the principal reaction rates (namely U238 capture and fission, relative to Pu239 fission) and the neutron spectrum were measured. These directly measured integral data which are particularly sensitive to the steel cross-sections can be used for the checking and systematic adjustment of data sets. The results may also be analysed so as to derive specific values for the integral capture cross-sections of steel and iron. Neutron balance equations were set-up for each of the lattices using the measured k(infinity) and reaction rates

  17. Study of the methodology for sensitivity calculations of fast reactors integral parameters

    International Nuclear Information System (INIS)

    Renke, C.A.C.

    1981-06-01

    A study of the methodology for sensitivity calculations of integral parameters of fast reactors for the adjustment of multigroup cross sections is presented. A description of several existent methods and theories is given, with special emphasis being regarded to variational perturbation theory, integrant of the sensitivity code VARI-1D used in this work. Two calculational systems are defined and a set of procedures and criteria is structured gathering the necessary conditions for the determination of the sensitivity coefficients. These coefficients are then computed by both the direct method and the variational perturbation theory. A reasonable number of sensitivity coefficients are computed and analyzed for three fast critical assemblies, covering a range of special interest of the spectrum. These coefficients are determined for severa integral parameters, for the capture and fission cross sections of the U-238 and Pu-239, covering all the energy up to 14.5 MeV. The nuclear data used were obtained the CARNAVAL II calculational system of the Instituto de Engenharia Nuclear. An optimization for sensitivity computations in a chainned sequence of procedures is made, yielding the sensitivities in the energy macrogroups as the final stage. (Author) [pt

  18. Actinide recycle potential in the Integral Fast Reactor (IFR) fuel cycle

    International Nuclear Information System (INIS)

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

    1990-01-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The use of metallic fuel in the IFR allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. In the electrorefining operation, uranium and plutonium are selectively transported from an anode to a cathode, leaving impurity elements, mainly fission products, either in the anode compartment or in a molten salt electrolyte. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management, because these actinides are automatically recycled back into the reactor for in-situ burning. Based on the recent IFR process development, a preliminary assessment has also been made to investigate the feasibility of further adapting the pyrochemical processes to directly extract actinides from LWR spent fuel. The results of this assessment indicate very promising potential and two most promising flowsheet options have been identified for further research and development. This paper also summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs. 5 refs., 4 figs., 4 tabs

  19. Lanthanide fission product separation from the transuranics in the integral fast reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Ackerman, J.P.

    1993-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed by Argonne National Laboratory. This reactor uses liquid-metal cooling and metallic fuel. Its spent fuel will be reprocessed using a pyrochemical method employing molten salts and liquid metals in an electrofining operation. The lanthanide fission products are a concern during reprocessing because of heating and fuel performance issues, so they must be removed periodically from the system to lessen their impact. The actinides must first be removed form the system before the lanthanides are removed as a waste stream. This operation requires a relatively good lanthanide-actinide separation to minimize both the amount of transuranic material lost in the waste stream and the amount of lanthanides collected when the actinides are first removed. A computer code, PYRO, that models these operations using thermodynamic and empirical data was developed at Argonne and has been used to model the removal of the lanthanides from the electrorefiner after a normal operating campaign. Data from this model are presented. The results demonstrate that greater that 75% of the lanthanides can be separated from the actinides at the end of the first fuel reprocessing campaign using only the electrorefiner vessel

  20. Study on integrated TRU multi-recycling in sodium cooled fast reactor CDFR

    International Nuclear Information System (INIS)

    Hu Yun; Xu Mi; Wang Kan

    2010-01-01

    In view of recently proposed closed fuel cycle strategy which would recycle the integrated transuranics (TRU) from PWR spent fuel in the fast reactors, the neutronics characteristics of TRU recycled in China Demonstration Fast Reactor (CDFR) are studied in this paper. The results show that loading integrated TRU to substitute pure Pu as driver fuel will mainly make the influence on sodium void worth and negligible effects on other parameters, and hence TRU recycling in CDFR is feasible from viewpoint of core neutronics. If TRU is multi-recycled, the variation of TRU composition depends on fuel types and the ratio of TRU and U when recycling. It is indicated that, when TRU is multi-recycled in CDFR with MOX fuel, the minor actinides (MA) fraction in TRU will firstly decrease to ∼7.24% (minimum) within 8 TRU recycle times and then slowly increase to ∼7.7% after 20 TRU recycle times; while when TRU is multi-recycled in CDFR with metal fuel (TRU-U-10Zr), the MA fraction in TRU will gradually approach to an equilibrium state with the MA fraction of ∼3.8%, demonstrating better MA transmutation effect in metal fuel core. No matter 7.7 or 3.8%, they are both lower than ∼10% in PWR spent fuel with burnup of 45 GWd/tU, which presents satisfying effect of MA amount controlling for TRU multi-recycling strategy. On the other hand, the corresponding recycling parameters such as TRU heat release and neutron emission rate are also much lower in metal fuel than those in MOX fuel. Moreover, TRU recycled in metal fuel will bring greater fissile Pu isotopes equilibrium fraction due to better breeding capability of metal fuel. Finally, it could be summarized that integrated TRU multi-recycling in fast reactor can make contributions to both breeding and transmutation, and such strategy is a prospective closed fuel cycle manner to achieve the object of effective control of cumulated MA amount and sustainable development of nuclear energy.

  1. A description of the demonstration Integral Fast Reactor fuel cycle facility

    International Nuclear Information System (INIS)

    Courtney, J.C.; Carnes, M.D.; Dwight, C.C.; Forrester, R.J.

    1991-01-01

    A fuel examination facility at the Idaho National Engineering Laboratory is being converted into a facility that will electrochemically process spent fuel. This is an important step in the demonstration of the Integral Fast Reactor concept being developed by Argonne National Laboratory. Renovations are designed to bring the facility up to current health and safety and environmental standards and to support its new mission. Improvements include the addition of high-reliability earthquake hardened off-gas and electrical power systems, the upgrading of radiological instrumentation, and the incorporation of advances in contamination control. A major task is the construction of a new equipment repair and decontamination facility in the basement of the building to support operations

  2. Fuel cycle facility control system for the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    Benedict, R.W.; Tate, D.A.

    1993-01-01

    As part of the Integral Fast Reactor (IFR) Fuel Demonstration, a new distributed control system designed, implemented and installed. The Fuel processes are a combination of chemical and machining processes operated remotely. To meet this special requirement, the new control system provides complete sequential logic control motion and positioning control and continuous PID loop control. Also, a centralized computer system provides near-real time nuclear material tracking, product quality control data archiving and a centralized reporting function. The control system was configured to use programmable logic controllers, small logic controllers, personal computers with touch screens, engineering work stations and interconnecting networks. By following a structured software development method the operator interface was standardized. The system has been installed and is presently being tested for operations

  3. Proliferation resistance of the fuel cycle for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.

    1993-01-01

    Argonne National Laboratory has developed an electrorefining pyrochemical process for recovery and recycle of metal fuel discharged from the Integral Fast Reactor (FR). This inherently low decontamination process has an overall decontamination factor of only about 100 for the plutonium metal product. As a result, all of the fuel cycle operations must be conducted in heavily shielded cells containing a high-purity argon atmosphere. The FR fuel cycle possesses high resistance to clandestine diversion or overt, state- supported removal of plutonium for nuclear weapons production because of two main factors: the highly radioactive product, which is also contaminated with heat- and neutron-producing isotopes of plutonium and other actinide elements, and the difficulty of removing material from the FR facility through the limited number of cell transfer locks without detection

  4. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. The integration of fast reactor to the fuel cycle in Slovakia

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Necas, V.

    2009-01-01

    A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of WWER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual WWER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the WWER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in WWER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from WWER-440 in the fission products. The next step is multi recycling of Pu in the fission products to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (Authors)

  7. The integration of fast reactor to the fuel cycle in Slovakia

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Necas, V.

    2009-01-01

    A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of VVER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual VVER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the VVER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in VVER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from VVER-440 in the FR. The next step is multirecycling of Pu in the FR to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (authors)

  8. SIMIFR: A code to simulate material movement in the Integral Fast Reactor

    International Nuclear Information System (INIS)

    White, A.M.; Orechwa, Yuri.

    1991-01-01

    The SIMIFR code has been written to simulate the movement of material through a process. This code can be used to investigate inventory differences in material balances, assist in process design, and to produce operational scheduling. The particular process that is of concern to the authors is that centered around Argonne National Laboratory's Integral Fast Reactor. This is a process which involves the irradiation of fissile material for power production, and the recycling of the irradiated reactor fuel pins into fresh fuel elements. To adequately simulate this process it is necessary to allow for locations which can contain either discrete items or homogeneous mixtures. It is also necessary to allow for a very flexible process control algorithm. Further, the code must have the capability of transmuting isotopic compositions and computing internally the fraction of material from a process ending up in a given location. The SIMIFR code has been developed to perform all of these tasks. In addition to simulating the process, the code is capable of generating random measurement values and sampling errors for all locations, and of producing a restart deck so that terminated problems may be continued. In this paper the authors first familiarize the reader with the IFR fuel cycle. The different capabilities of the SIMIFR code are described. Finally, the simulation of the IFR fuel cycle using the SIMIFR code is discussed. 4 figs

  9. Advanced control system for the Integral Fast Reactor fuel pin processor

    International Nuclear Information System (INIS)

    Lau, L.D.; Randall, P.F.; Benedict, R.W.; Levinskas, D.

    1993-01-01

    A computerized control system has been developed for the remotely-operated fuel pin processor used in the Integral Fast Reactor Program, Fuel Cycle Facility (FCF). The pin processor remotely shears cast EBR- reactor fuel pins to length, inspects them for diameter, straightness, length, and weight, and then inserts acceptable pins into new sodium-loaded stainless-steel fuel element jackets. Two main components comprise the control system: (1) a programmable logic controller (PLC), together with various input/output modules and associated relay ladder-logic associated computer software. The PLC system controls the remote operation of the machine as directed by the OCS, and also monitors the machine operation to make operational data available to the OCS. The OCS allows operator control of the machine, provides nearly real-time viewing of the operational data, allows on-line changes of machine operational parameters, and records the collected data for each acceptable pin on a central data archiving computer. The two main components of the control system provide the operator with various levels of control ranging from manual operation to completely automatic operation by means of a graphic touch screen interface

  10. Measurement control design and performance assessment in the Integral Fast Reactor fuel cycle

    International Nuclear Information System (INIS)

    Orechwa, Y.; Bucher, R.G.

    1994-01-01

    The Integral Fast Reactor (IFR)--consisting of a metal fueled and liquid metal cooled reactor together with an attendant fuel cycle facility (FCF)--is currently undergoing a phased demonstration of the closed fuel cycle at Argonne National Laboratory. The recycle technology is pyrometalurgical based with incomplete fission product separation and all transuranics following plutonium for recycle. The equipment operates in batch mode at 500 to 1,300 C. The materials are highly radioactive and pyrophoric, thus the FCF requires remote operation. Central to the material control and accounting system for the FCF are the balances for mass measurements. The remote operation of the balances limits direct adjustment. The radiation environment requires that removal and replacement of the balances be minimized. The uniqueness of the facility precludes historical data for design and performance assessment. To assure efficient operation of the facility, the design of the measurement control system has called for procedures which assess the performance of the balances in great detail and will support capabilities for the correction of systematic changes in the performance of the balances through software

  11. Fast breeder reactors

    International Nuclear Information System (INIS)

    Waltar, A.E.; Reynolds, A.B.

    1981-01-01

    This book describes the major design features of fast breeder reactors and the methods used for their design and analysis. The foremost objective of this book is to fulfill the need for a textbook on Fast Breeder Reactor (FBR) technology at the graduate level or the advanced undergraduate level. It is assumed that the reader has an introductory understanding of reactor theory, heat transfer, and fluid mechanics. The book is expected to be used most widely for a one-semester general course on fast breeder reactors, with the extent of material covered to vary according to the interest of the instructor. The book could also be used effectively for a two-quarter or a two-semester course. In addition, the book could serve as a text for a course on fast reactor safety since many topics other than those appearing in the safety chapters relate to FBR safety. Methodology in fast reactor design and analysis, together with physical descriptions of systems, is emphasized in this text more than numerical results. Analytical and design results continue to change with the ongoing evolution of FBR design whereas many design methods have remained fundamentally unchanged for a considerable time

  12. The fast breeder reactor

    International Nuclear Information System (INIS)

    Patterson, W.

    1990-01-01

    The author criticises the United Kingdom Atomic Energy Authority's fast breeder reactor programme in his evidence to the House of Commons Select Committee on Energy in January 1990. He argues for power generation by renewable means and greater efficiency in the use rather than in the generation of electricity. He refutes the arguments for nuclear power on the basis of reduced global warming as he claims support technology produces significant amounts of carbon dioxide in any case. Serious doubts are raised about the costs of a fast breeder reactor programme compared to, say, generation by pressurised water reactors. The idea of a uranium scarcity in several decades is also refuted. The reliability of fast breeder reactor technology is called into question. He argues against reprocessing plutonium for economic, health and safety reasons. (UK)

  13. Treatment of wastes in the Integral Fast Reactor (IFR) fuel cycle

    International Nuclear Information System (INIS)

    Ackerman, J.P.; Johnson, T.R.; Chow, L.S.H.; Carls, E.L.; Hannum, W.H.; Laidler, J.J.

    1997-01-01

    In both the reactor portion and the fuel-cycle portion of the Integral Fast Reactor (IFR), handling, treatment and disposal of wastes are simpler than in current fuel cycles. The vast majority (> 99.9%) of the very-long-lived radioactive TRU elements are not sent to the repository; rather, they are recycled. High-level waste volume from the IFR process (called ''the pyroprocess'') is lower than that from either the direct disposal of spent fuel or from conventional PUREX-type reprocessing. The quantity of low-level waste is very low. In the pyroprocess, the actinides are recovered and separated from the bulk of the fission products by an electrorefining step wherein the actinides are electrotransported from chopped fuel elements and deposited at cathodes. The volatile fission products xenon, krypton, and tritium are collected for long-term storage and decay. Zirconium and the ''noble metal'' fission products (those that are less easily oxidized than zirconium) remain in the anode compartment, to be removed with the fuel cladding fragments and made into a metal waste form. The remaining fission products collect in the salt as chlorides. A process has been developed to periodically remove the contaminated salt from the electrorefiner, separate most of the fission products, and return the purified salt in a form that is ready for continuing use. To clean up the electrorefiner salt, the fission products are removed by ion exchange onto a column of Zeolite A. After the purification step, the column material and the contained fission products are converted to a mineral waste form for disposal. The processes and equipment for waste isolation and conversion to suitable disposal forms are described in this paper. (author)

  14. The fast breeder reactor

    International Nuclear Information System (INIS)

    Keck, O.

    1984-01-01

    Nowadays the fast-breeder reactor is a negative symbol of advanced technology which is getting out of control and, due to its complexity, is incomprehensible for politicians and therefore by-passes the established order. The author lists the most important decisions over state aid to the fast-breeder-reactors up until the mid-seventies and uses documents from the appropriate advisory bodies as reference. He was also aided by interviews with those directly involved with the project. The empirical facts forces us to discard our traditional view of the relationship between state and industry with regard to advanced technology. The author explains that it is impossible to find any economic value in the fast-breeder reactor. The insight gained through this project allows him to draw conclusions which apply to all aspects of state aid to advanced technology. (orig.) [de

  15. Recent research on structural integrity in support of the european fast reactor

    International Nuclear Information System (INIS)

    Acker, D.; Townley, C.H.A.

    1991-01-01

    For the European Fast Reactor, which is being developed jointly in France, Germany and the U.K., qualified material controlled fabrication processes and validated design rules are essential to achieve cost reduction in components, and at the same time, maintain the required standards of safety and reliability. In the framework of an inter-governmental agreement, the R and D organisations: CEA in France, KfK and INTERATOM in Germany and AEA Technology in the UK, have co-ordinated their work to support the EFR project. In the Structural Integrity area, this co-ordination is undertaken by the AGT 9B group, which: The Purpose of the present paper is to review the progress which has been made up to the end of 1990, as recorded by the 35 AGT9B Milestone Reports which had been issued by this date. Only a brief outline of the main results will be given. Detailed descriptions of the work will be found in the papers presented to SMIRT Conferences in 1987, 1989 and 1991 and referenced in chapter 7. The results fall into five areas - General design guidance - Analysis of structures submitted to repeated loading - Fracture mechanics - Buckling - Design rules for specific components

  16. Waste removal in pyrochemical fuel processing for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Ackerman, J.P.; Johnson, T.R.; Laidler, J.J.

    1994-01-01

    Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix

  17. The Mass Tracking System for the Integral Fast Reactor fuel cycle

    International Nuclear Information System (INIS)

    Adams, C.H.; Beitel, J.C.; Birgersson, G.; Bucher, R.G.; Carrico, C.B.; Daly, T.A.; Keyes, R.W.

    1994-01-01

    As part of the Fuel Cycle Facility (FCF) of Argonne National Laboratory's Integral Fast Reactor (IFR) demonstration, a computer-based Mass-Tracking (MTG) System has been developed. The MTG System collects, stores, retrieves and processes data on all operations which directly affect the flow of process material through FCF and supports such activities as process modeling, compliance with operating limits (e.g., criticality safety), material control and accountability and operational information services. Its architecture is client/server, with input and output connections to operator's equipment-control stations on the floor of FCF as well as to terminal sessions. Its heterogeneous database includes a relational-database manager as well as both binary and ASCII data files. The design of the database, and the software that supports it, is based on a model of discrete accountable items distributed in space and time and constitutes a complete historical record of the material processed in FCF. Although still under development, much of the MTG System has been qualified and is in production use

  18. Integrated leak rate testing of the fast flux test facility reactor containment building

    International Nuclear Information System (INIS)

    James, E.B.; Farabee, O.A.; Bliss, R.J.

    1978-01-01

    The initial Integrated Leak Rate Test (ILRT) of the Fast Flux Test Facility containment building was performed from May 27 to June 2, 1978. The test was conducted in air with systems vented and with the containment recirculating coolers in operation. 10 psig and 5 psig tests were run using the absolute pressure test method. The measured leakage rates were .033% Vol/24 hr. and -.0015% Vol/24 hrs. respectively. Subsequent verification tests at both 10 psig and 5 psig proved that the test equipment was operating properly and it was sensitive enough to detect leaks at low pressures. This ILRT was performed at a lower pressure than any previous ILRT on a reactor containment structure in the United States. While the initial design requirements for ice condenser containments called for a part pressure test at 6 psig, the tests were waived due to the apparent statistical problems of data analysis and the repeatability of the data itself at such low pressure. In contrast to this belief, both the 5 and 10 psig ILRT's were performed in a successful manner at FFTF

  19. Management of transuranics using the Integral Fast Reactor (IFR) fuel cycle

    International Nuclear Information System (INIS)

    Wade, D.C.

    1994-01-01

    The 50 years of activities following the discovery of self-sustaining fission chains have given rise to a buildup of roughly 900 tons of manmade transuranics. Of the total, about 260 tons of Pu 239 were generated for use in weapons while the remainder were generated as a byproduct of electrical power produced worldwide by the commercial thermal nuclear power industry. What is to be done with these actinides? The options for disposition include interminable storage, burial, or recycle for use. The pros and cons of each option are being vigorously debated regarding the impact upon the issues of human and ecological risk -- both current and future; weapons proliferation potential -- both current and future; and total life cycle benefits and costs. As to the options for utilization, commercial uses for actinides (uranium and transuranics) are of limited diversity. The actinides have in the past and will in the future find application in large scale mostly by virtue of their ability to release energy through fission, and here their utility is unmatched -- whether the application be in commercial electricity generation or in armaments. The integral Fast Reactor (IFR) fuel cycle offers a number of features for management of the current and future burden of manmade transuranic materials and for capturing the energy content of the U 238 . These features are discussed here

  20. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1982-01-01

    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.)

  1. Sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hokkyo, N; Inoue, K; Maeda, H

    1968-11-21

    In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

  2. The role of structural integrity in liquid metal fast breeder reactor safety

    International Nuclear Information System (INIS)

    Holmes, J.A.G.

    1982-01-01

    Extensive studies have demonstrated the favourable safety characteristics of liquid metal fast breeder reactors, which are attributable to both their inherent features and the engineered safeguards which are included. This requires demonstration that there is no risk of sudden catastrophic failure of the core support system allowing the core to drop off the control rods to give a prompt critical reactivity excursion. An important part of our work in support of the safety case for the U.K. Commercial Demonstration Fast Reactor is to demonstrate that such a failure is virtually incredible. This covers design features, study of the fracture behaviour of stainless steel structures, and inspection and monitoring during fabrication and service. The paper gives a broad description of the relevant design features and supporting work programme

  3. Behavior of low-burnup metallic fuels for the integral fast reactor at elevated temperatures in ex-reactor tests

    International Nuclear Information System (INIS)

    Tsai, Hanchung; Liu, Yung Y.; Wang, Da-Yung; Kramer, J.M.

    1991-07-01

    A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing design correlation, which provides a conservative margin to cladding failure. In a test which enveloped a wide range of postulated reactor transient events, a substantial temporal cladding integrity margin was demonstrated for an intact, whole fuel pin. The cause of the eventual pin breach was reaction-induced cladding thinning combined with fission-gas pressure loading. The behavior of the breached pin was benign. 7 refs., 7 figs., 1 tab

  4. Dounreay fast reactor

    International Nuclear Information System (INIS)

    Maclennan, R.; Eggar, T.; Skeet, T.

    1992-01-01

    The short debate which followed a private notice question asking for a statement on Government policy on the future of the European fast breeder nuclear research programme is reported verbatim. In response to the question, the Minister for Energy said that the Government had decided in 1988 that the Dounreay prototype fast reactor would close in 1994. That decision had been confirmed. Funding of fast breeder research and development beyond 1993 is not a priority as commercialization is not expected until well into the next century. Dounreay will be supported financially until 1994 and then for its subsequent decommissioning and reprocessing of spent fuel. The debate raised issues such as Britain losing its lead in fast breeder research, loss of jobs and the Government's nuclear policy in general. However, the Government's position was that the research had reached a stage where it could be left and returned to in the future. (UK)

  5. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    1983-02-01

    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.)

  6. The integral fast reactor (IFR) concept: Physics of operation and safety

    International Nuclear Information System (INIS)

    Wade, D.C.; Chang, Y.I.

    1987-01-01

    The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

  7. The integral fast reactor (IFR) concept: physics of operation and safety

    International Nuclear Information System (INIS)

    Wade, D.C.; Chang, Y.I.

    1987-01-01

    The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

  8. Fast reactor database

    International Nuclear Information System (INIS)

    1996-02-01

    This publication contains detailed data on liquid metal cooled fast reactors (LMFRs), specifically plant parameters and design details. Each LMFR power plant is characterized by about 400 parameters, by design data and by relevant materials. The report provides general and detailed design characteristics including structural materials, data on experimental, demonstration, prototype and commercial size LMFRs. The focus is on practical issues that are useful to engineers, scientists, managers and university students and professors. The report includes updated information contained in IAEA previous publications on LMFR plant parameters: IWGRF/51 (1985) and IWGFR/80 (1991) and reflects experience gained from two consultants meetings held in Vienna (1993,1994). This compilation of data was produced by members of the IAEA International Working Group on Fast Reactors (IWGFR)

  9. Knowledge management in fast reactors

    International Nuclear Information System (INIS)

    Kuriakose, K.K.; Satya Murty, S.A.V.; Swaminathan, P.; Raj, Baldev

    2010-01-01

    This paper highlights the work that is being carried out in Knowledge Management of Fast Reactors at Indira Gandhi Centre for Atomic Research (IGCAR) including a few examples of how the knowledge acquired because of various incidents in the initial years has been utilized for the successful operation of Fast Breeder Test Reactor. It also briefly refers to the features of the IAEA initiative on the preservation of Knowledge in the area of Fast Reactors in the form of 'Fast Reactor Knowledge Organization System' (FR-KOS), which is based on a taxonomy for storage and mining of Fast Reactor Knowledge. (author)

  10. The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

    International Nuclear Information System (INIS)

    Smith, R.R.

    1986-01-01

    This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies

  11. Design codes for fast reactor steam generators

    International Nuclear Information System (INIS)

    Townley, C.H.A.

    1978-01-01

    The paper reviews the design methods and design criteria which are available for fast reactor structures, and discusses the materials data which are required to demonstrate the integrity of the plant components. (author)

  12. Fast breeder reactors

    International Nuclear Information System (INIS)

    Ollier, J.L.

    1987-01-01

    The first industrial-scale fast breeder reactor (FBR) is the Superphenix I at Crays-Melville. It was designed and built by Novatome, a French company, and Ansaldo, an Italian company. The advantages of FBRs are summarized. The status of Superphenix and the testing schedule is given. The stages in its power escalation in 1986 are given. The article is optimistic about the future for FBRs and expects FBRs to take over from PWRs at the beginning of the 21st Century. To achieve economic viability, European financial cooperation for the research and development programme is advocated. (UK)

  13. Use of integral data in the development of design methods for fast reactors

    International Nuclear Information System (INIS)

    Doncals, R.A.; Lake, J.A.; Paik, N.C.

    1978-01-01

    The paper describes an LMFBR nuclear design methodology which has been strongly influenced by the availability of integral data, by the expansion of the differential nuclear data base, by improvements in large nuclear design computer codes, and by the specific reactor under consideration. The accuracy of the nuclear data base has been improved as the result of detailed differential measurements as well as extensive integral testing in the ZPR and ZPPR criticals. Due to the increased interest in radial parfait designs, the applicability of the design data and methods to the analysis of heterogeneous LMFBR systems have been explored. The ability of the design data and methods to predict integral parameters in ZPPR is also discussed

  14. The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

    Science.gov (United States)

    Mosunova, N. A.

    2018-05-01

    The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

  15. Experimental determination of neutron capture cross sections of fast reactor structure materials integrated in intermediate energy spectra. Vol. 2: description of experimental structure

    International Nuclear Information System (INIS)

    Tassan, S.

    1978-01-01

    A selection of technical documents is given concerning the experimental determination of the neutron capture cross-sections of fast reactor structural materials (Fe, Cr, Ni...) integrated over the intermediate energy spectra. The experimental structure project and modifications of the reactor RB2 for this experiment, together with criticality and safety calculations, are presented

  16. New fast reactor installation concept

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The large size and complexity of fast reactor installations are emphasised and these difficulties will be increased with the advent of fast reactors of higher power. In this connection a new concept of fast reactor installation is described with a view to reducing the size of the installation and enabling most components, including even the primary vessel, to be constructed within the confines of a workshop. Full constructional details are given. (U.K.)

  17. Fast reactors and nonproliferation

    International Nuclear Information System (INIS)

    Orlov, V.V.

    1997-01-01

    1.Three aspects of nonproliferation relevant to nuclear power are: Pu buildup in NPP spent fuel cooling ponds (∼ 104 t in case of consumption of ∼ 107 t cheap uranium). Danger of illegal radiochemical extraction of Pu for weapons production; Pu extraction from NPP fuel at the plants available in nuclear countries, its burning along with weapon-grade Pu in NPP reactors or in special-purpose burners; increased hazard of nuclear weapons sprawl with breeders and closed fuel cycle technology spreading all over the world. 2.The latter is one of major obstacles to creation of large-scale nuclear power. 3.Nuclear power of the first stage using 235 U will be able to meet the demands of certain fuel-deficient countries and regions, replacing ∼ 5-10% of conventional fuels in the global consumption for a number of decades. 4.Fast reactors of the first generation and the currently employed fuel technology are far from exhausting their potential for solving economic problems and meeting the challenges of safety, radioactive waste and nonproliferation. Development of large-scale nuclear power will become an option accepted by society for solving energy problems in the following century, provided a breeder technology is elaborated and demonstrated in the next 15-20 years, which would comply with the totality of the following requirement: full internal Pu breeding deterministic elimination of severe accidents involving fuel damage and high radioactivity releases: fast runaway, loss of coolant, fires, steam and hydrogen explosions, etc.; reaching a balance between radioactive wastes disposed of and uranium mined in terms of radiation hazard; technology of closed fuel cycle preventing its use for Pu extraction and permitting physical protection from fuel thefts;economic competitiveness of nuclear power for most of countries and regions, i.e. primarily the cost of NPPs with fat reactors is to be below the cost of modern LWR plants, etc

  18. Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Batte, G.L.; Pahl, R.G.; Hofman, G.L.

    1993-01-01

    This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  20. Fast reactor fuel reprocessing in the UK

    International Nuclear Information System (INIS)

    Allardice, R.H.; Williams, J.; Buck, C.

    1977-01-01

    Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the U.K. since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium based fast reactor system and the importance of establishing at an early stage fast reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high burn-up thermal reactor oxide fuel. In consequence, the U.K. has decided to reprocess irradiated fuel from the 250 MW(E) Prototype Fast Reactor as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small scale fully active demonstration plant have been carried out over the past 5 years and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant a parallel development programme has been initiated to provide the basis for the design of a large scale fast reactor fuel reprocessing plant to come into operation in the late 1980s to support the projected U.K. fast reactor installation programme. The paper identifies the important differences between fast reactor and thermal reactor fuel reprocessing technologies and describes some of the development work carried out in these areas for the small scale P.F.R. fuel reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast reactor fuel reprocessing plant is outlined and the current design philosophy is discussed

  1. Fast reactor programme

    International Nuclear Information System (INIS)

    Hoekstra, E.K.

    1976-11-01

    Estimated reactivity effects of fission products in the SNR-300 fast breeder are given. Neutron cross sections of 127 I and 129 I are also given. Results of the in-pile canning failure experiments on fuel pins R54-F35 and F39 are discussed. Sinter experiments using mixed UC-UN powders are reported. Results of tensile tests on high-dose and low-dose irradiated specimens of 18Cr1 1Ni stainless steel (DIN 1.4948) used in the SNR-300 reactor vessel are given. It is shown that the aerosol behaviour in condensing sodium vapour can be described by the same MADCA model developed for the decay of aerosols in condensing water vapour. Results of heat transfer measurements in the electrically heated 28-rod bundle under liquid-phase and subsequently under two-phase conditions are commented on

  2. Uranium and the fast reactor

    International Nuclear Information System (INIS)

    Price, T.

    1982-01-01

    The influence of uranium availability upon the future of the fast reactor is reviewed. The important issues considered are uranium reserves and resources, uranium market prices, fast reactor economics and the political availability of uranium to customers in other countries. (U.K.)

  3. The safety of fast reactors

    International Nuclear Information System (INIS)

    Justin, F.

    1976-01-01

    A response is made to the main questions that a man in the street may arise concerning fast breeder reactors, in particular: the advantages of this line, dangerous materials contained in fast breeder reactors, containment shells protecting the environment from radiations, main studies now in progress [fr

  4. Fast reactor physics - an overview

    International Nuclear Information System (INIS)

    Lee, S.M.

    2004-01-01

    An introduction to the basic features of fast neutron reactors is made, highlighting the differences from the more conventional thermal neutron reactors. A discussion of important feedback reactivity mechanisms is given. Then an overview is presented of the methods of fast reactor physics, which play an important role in the successful design and operation of fast reactors. The methods are based on three main elements, namely (i) nuclear data bases, (ii) numerical methods and computer codes, and (iii) critical experiments. These elements are reviewed and the present status and future trends are summarized. (author)

  5. Fast mixed spectrum reactor concept

    International Nuclear Information System (INIS)

    Kouts, H.J.C.; Fischer, G.J.; Cerbone, R.J.

    1979-04-01

    The Fast Mixed Spectrum Reactor is a highly promising concept for a fast reactor with improved features of proliferation resistance, and excellent utilization of uranium resources. In technology, it can be considered to be a branch of fast breeder development, though its operation and implications are different from those of FBR'S in important respects. Successful development programs are required in several areas to bring FMSR to reality, but the payoff from a successful program can be high

  6. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1981-06-01

    The accuracy requirements and the status of the evaluated fission-product cross sections for fast reactors are reviewed; the work on calculating the sensitivity of the sodium void effect to fission-product cross sections is described; some results of the intercomparison of adjusted data sets for capture cross sections of fission-products (RCN-2A and CARNAVAL-IV) are discussed; the applicability of the maximum-likelihood method for the analysis of resolved resonance parameters for a large class of fission-product nuclides is demonstrated; the neutron cross sections for corrosion product 64 Ni are evaluated. Some results of post-irradiation examination of a loss-of-cooling experiment are given; the progress in testing the equipment and instrumentation for transient-overpower experiments is reported. The proceedings in the thermochemical investigations on uranium compounds with some fission-products are described. The creep behaviour of a heat of DIN 1.4948 parent metal is investigated with respect to the changes in strain with different test temperatures. Sodium smoke aerosols have been produced and analysed with respect to their aerodynamic behaviour and morphology. The two-phase local boiling experiments have been analysed to find criteria for the occurrence of different boiling regimes with the objection to deduce general dryout correlations

  7. Fast reactor recharging device

    International Nuclear Information System (INIS)

    Artemiev, L.N.; Kurilkin, V.V.

    1979-01-01

    Disclosure is made of a device for recharging a fast-neutron reactor, intended for the transfer of fuel assemblies and rods of the control and safety system, having profiled heads to be gripped on the outside. The device comprises storage drums whose compartments for rods of the control and safety system are identical to compartments for fuel assemblies. In order to store and transport rods of the control and safety system from the storage drums to the recharging mechanism provision is made for sleeve-type holders. When placed in such a holder, the dimensions of a rod of the control and safety system are equal to those of a fuel assembly. To join a holder to a rod of the control and safety system, on the open end of each holder there is mounted a collet, whereas on the surface of each rod of the control and safety system, close to its head, there is provided an encircling groove to interact with the collet. The grip of the recharging mechanism is provided with a stop interacting with the collet in order to open the latter and withdraw the safety and control system rod from its holder

  8. Methodological study of the adjustment of multigroup cross sections for fast reactors, by means of integral data

    International Nuclear Information System (INIS)

    Bastos, H.F.B.N.

    1979-01-01

    In this work a study of the methodology of the adjustment of multigroup cross sections by means of integral data is presented. A synthesis of the principal methods existent and the mathematical development of the adaptation of one of them are made. A calculational system is built from this reference method, with the basic conditions for the operation of the process of adjustment. In order to test the system developed and analyze several problems related to the adjustment, a series of trial adjustments was made with the value of the U 235 fission cross section from the infinite dilution library used in the calculational system for fast reactors of the Instituto de Engenharia Nuclear. (author)

  9. Designing for a safe response to operational and severe accident initiators in the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1994-01-01

    A method is described for optimizing the plant control strategy for a liquid metal reactor with respect to safety margins sustained in unprotected upset events. The optimization is performed subject to the normal requirements for reactor startup, load change and compensation for reactivity changes over the cycle. The method provides a formal approach to the process of exploiting the innate self-regulating property of a metal fueled reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error

  10. Fast reactor database. 2006 update

    International Nuclear Information System (INIS)

    2006-12-01

    Liquid metal cooled fast reactors (LMFRs) have been under development for about 50 years. Ten experimental fast reactors and six prototype and commercial size fast reactor plants have been constructed and operated. In many cases, the overall experience with LMFRs has been rather good, with the reactors themselves and also the various components showing remarkable performances, well in accordance with the design expectations. The fast reactor system has also been shown to have very attractive safety characteristics, resulting to a large extent from the fact that the fast reactor is a low pressure system with large thermal inertia and negative power and temperature coefficients. In addition to the LMFRs that have been constructed and operated, more than ten advanced LMFR projects have been developed, and the latest designs are now close to achieving economic competitivity with other reactor types. In the current world economic climate, the introduction of a new nuclear energy system based on the LMFR may not be considered by utilities as a near future option when compared to other potential power plants. However, there is a strong agreement between experts in the nuclear energy field that, for sustainability reasons, long term development of nuclear power as a part of the world's future energy mix will require the fast reactor technology, and that, given the decline in fast reactor development projects, data retrieval and knowledge preservation efforts in this area are of particular importance. This publication contains detailed design data and main operational data on experimental, prototype, demonstration, and commercial size LMFRs. Each LMFR plant is characterized by about 500 parameters: physics, thermohydraulics, thermomechanics, by design and technical data, and by relevant sketches. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors with complete technical information of a total of 37 LMFR

  11. Fast reactor collaboration in Europe

    International Nuclear Information System (INIS)

    Smith, G.E.I.

    1987-01-01

    Fast reactors have been developed in several European countries, the United Kingdom, France, Germany and Italy. A suggestion to collaborate on fast reactor research and development resulted in an Intergovernmental Memorandum of Understanding signed in 1984 by the UK, France, Germany, Italy and Belgium. Holland was expected to join later. This provided for co-operation between electric utilities, reactor design, research and development companies and fuel cycle companies. Three steering committees have so far been set up, the European fast reactor utilities Group, the European research and development and the European fuel cycle steering committees. Progress on these is detailed. The main areas of technology exchange are listed in the Appendix. The possibility exists for a series of three large demonstration plants to be built in Europe and a fuel reprocessing plant to confirm the reactor system. (U.K.)

  12. Review of fast reactor activities in India

    Energy Technology Data Exchange (ETDEWEB)

    Paranjpe, S R [Reactor Research Centre, Kalpakkam, Tamil Nadu (India)

    1981-05-01

    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.

  13. Review of fast reactor activities in India

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1981-01-01

    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

  14. Review of fast reactor activities

    International Nuclear Information System (INIS)

    1982-01-01

    A description of some highlights of the activities performed by the Commission of the European Communities in the field of fast reactors is given. They fall into two categories: coordinating and harmonizing activities and research activities. The former are essentially performed in the frame of the Fast Reactor Coordinating Committee (FRCC), the latter in the Commission's Joint Research Center and to some extent under contract in research centers of the Member States

  15. The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.R.

    1986-01-01

    This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

  16. Review of fast reactor activities

    International Nuclear Information System (INIS)

    Haeussermann, W.; Royen, J.

    1978-01-01

    Since 1971, when the Co-ordinating Group on Gas-Cooled Fast reactors Development was set up, the participating countries have maintained an interest in keeping this option as a back-up solution to the sodium cooled fast reactors. Two different concepts were investigated, one based on coated particle type fuel elements and the other on pin type fuel elements. The coated particles studies have been brought to an end, and resources were concentrated on the further development of the pin type concept. The work done in previous years covered design and safety investigations, heat transfer studies and irradiation experiments in thermal reactors

  17. Review of fast reactor activities in India

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1982-01-01

    A review of fast reactor activities in India is introduced. One stage of construction of the Fast Breeder Test Reactor (FBTR) and design studies for 500MWe Prototype Fast Breeder Reactor (PFBR) are briefly summarized. The emphasis is on fast reactor physics, materials studies, radiochemistry, and the safety and fuel reprocessing programme

  18. Fast breeder reactor

    International Nuclear Information System (INIS)

    Ito, Shin-ichi; Maki, Koichi.

    1975-01-01

    Object: To conserve loaded fuel, aquire controllable surplus reaction degree, increase the breeding index, flatten output and improve sealing of neutrons by inserting a decelerating substance in a blanket section. Structure: A decelerating substance such as beryllium or beryllium oxide is inserted in a blanket section between an outer reactor core and reflector. With this arrangement, neutrons are decelerated to increase the low energy components, which are partly subjected to reflection by the outer reactor core to thereby reduce leakage of neutrons from the reactor core. (Kamimura, M.)

  19. Astrid (fast breeder nuclear reactor)

    International Nuclear Information System (INIS)

    2014-01-01

    This document presents ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a French project of sodium-cooled fast breeder reactor, fourth generation reactor which should be fuelled by uranium 238 rather than uranium 235, and should therefore need less extracted natural uranium to produce electricity. The operation principle of fast breeder reactors is described. They notably directly consume plutonium, allow an easier radioactive waste management as they transform long life radioactive elements into shorter life elements by transmutation. The regeneration process is briefly described, and the various operation modes are evoked (iso-generator, sub-generator, and breeder). Some peculiarities of sodium-cooled reactors are outlined. The Astrid operation principle is described, its main design innovations outlined. Various challenges are discussed regarding safety of supply and waste processing, and the safety of future reactors. Major actors are indicated: CEA, Areva, EDF, SEIV Alcen, Toshiba, Rolls Royce, and Comex. Some key data are indicated: expected lifetime, expected availability rate, cost. The projected site is Marcoule and fast breeder reactors operated or under construction in the world are indicated. The document also proposes an overview of the background and evolution of reactors of 4. generation

  20. Fast reactors: the industrial perspective

    International Nuclear Information System (INIS)

    Vaughan, R.D.

    1986-01-01

    Industrial participation in the development of the fast reactor is reviewed, from the construction of PFR at Dounreay to the initial steps towards collaboration in Europe. The optimum design of the fast reactor has changed considerably from the days when it was needed urgently to forestall a shortage of uranium to today when uranium is abundant and cheap. The evolution of the reactor design over this period is described. Collaboration in Europe is shown to be the only answer to high development costs and the search for a reactor which will compete with thermal reactors in today's environment. The partner countries in this collaboration are all motivated differently, and this is leading to some delays in concluding the necessary agreements. The objective on the industrial front is now to participate in the two or three demonstration fast reactors that will be built in Europe during the remainder of the century leading, it is hoped, to a competitive reactor design by the year 2000. (author)

  1. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1979-10-01

    Various experiments being performed at the SNR reactor are described including: capture cross sections of various nuclei; fuel can failure; creep testing of welded joints; gas leakage through concrete/steel interfaces; testing of the test section of the four rod bundle for Laser Doppler Anemometry

  2. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1996-01-01

    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

  3. Integrity Evaluation of Control Rod Assembly for Sodium-Cooled Fast Reactor due to Drop Impact

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun Seung; Yoon, Kyung Ho; Kim, Hyung Kyu; Cheon, Jin Sik; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-03-15

    The CA (Control Assembly) of an SFR has a CRA(Control Rod Assembly) with an inner duct and control rod. During an emergency situation, the CRA falls into the duct of the CA for a rapid shut-down. The drop time and impact velocity of the CRA are important parameters with respect to the reactivity insertion time and the structural integrity of the CRA. The objective of this study was to investigate the dynamic behavior and integrity of the CRA owing to a drop impact. The impact analysis of the CRA under normal/abnormal drop conditions was carried out using the commercial FEM code LS-DYNA. Results of the drop impact analysis demonstrated that the CRA maintained structural integrity, and could be safely inserted into the flow hole of the damper under abnormal conditions.

  4. Integrity Evaluation of Control Rod Assembly for Sodium-Cooled Fast Reactor due to Drop Impact

    International Nuclear Information System (INIS)

    Lee, Hyun Seung; Yoon, Kyung Ho; Kim, Hyung Kyu; Cheon, Jin Sik; Lee, Chan Bock

    2017-01-01

    The CA (Control Assembly) of an SFR has a CRA(Control Rod Assembly) with an inner duct and control rod. During an emergency situation, the CRA falls into the duct of the CA for a rapid shut-down. The drop time and impact velocity of the CRA are important parameters with respect to the reactivity insertion time and the structural integrity of the CRA. The objective of this study was to investigate the dynamic behavior and integrity of the CRA owing to a drop impact. The impact analysis of the CRA under normal/abnormal drop conditions was carried out using the commercial FEM code LS-DYNA. Results of the drop impact analysis demonstrated that the CRA maintained structural integrity, and could be safely inserted into the flow hole of the damper under abnormal conditions.

  5. The instrumentation of fast reactor

    International Nuclear Information System (INIS)

    Endo, Akira

    2003-03-01

    The author has been engaged in the development of fast reactors over the last 30 years with both an involvement with the early technology development on the experimental breeder reactor Joyo, and latterly continuing this work on the prototype breeder reactor, Monju. In order to pass on this experience to younger engineers this paper is produced to outline this experience in the sincere hope that the information given will be utilised in future educational training material. The paper discusses the wide diversity on the associated instrument technology which the fast breeder reactor requires. The first chapter outlines the fast reactor system, followed by discussions on reactor instrumentation, measurement principles, temperature dependencies, and verification response characteristics from various viewpoints, are discussed in chapters two and three. The important issues of failed fuel location detection, and sodium leak detection from steam generators are discussed in chapters 4 and 5 respectively. Appended to this report is an explanation on the methods of measuring response characteristics on instrumentation systems using error analysis, random signal theory and measuring method of response characteristic by AR (autoregressive) model on which it appears is becoming an indispensable problem for persons involved with this technology in the future. (author)

  6. Fast Reactors and Nuclear Nonproliferation

    International Nuclear Information System (INIS)

    Avrorina, E.N.; Chebeskovb, A.N.

    2013-01-01

    Conclusion remarks: 1. Fast reactor start-up with U-Pu fuel: – dependent on thermal reactors, – no needs in U enrichment, – needs in SNF reprocessing, – Pu is a little suitable for NED, – practically impossible gun-type NED, – difficulties for implosion-type NED: necessary tests, advanced technologies, etc. – Pu in blankets is similar to WPu by isotopic composition, – Use of blanket for production isotopes (e.g. 233 U), – Combined reprocessing of SNF: altogether blanket and core, – Blanket elimination: decrease in Pu production – No pure Pu separation. 2. Fast reactor start-up with U fuel: - Needs in both U enrichment and SNF reprocessing, - Independent of thermal reactors, - Good Pu bred in the core let alone blankets, - NED of simple gun-type design, - Increase of needs in SWU, - Increased demands in U supply. 3. Fast reactors for export: - Uranium shortage, - To replace thermal reactors in future, - No blankets (depends on the country, though), - Fuel supply and SNF take back, - International centers for rendering services of NFC. Time has come to remove from FRs and their NFC the label unfairly identifying them as the most dangerous installations of nuclear power from the standpoint of being a proliferation problem

  7. Investigation of molten salt fast reactor

    International Nuclear Information System (INIS)

    Kubota, Kenichi; Konomura, Mamoru

    2002-01-01

    On survey research for practicability strategy of fast reactor (FR) (phase 1), to extract future practicability image candidates of FR from wide options, in addition to their survey and investigation objects of not only solid fuel reactors of conventional research object but also molten salt reactor as a flowing fuel reactor, investigation on concept of molten salt FR plant was carried out. As a part of the first step of the survey research for practicability strategy, a basic concept on plant centered at nuclear reactor facility using chloride molten salt reactor capable of carrying out U-Pu cycle was examined, to perform a base construction to evaluate economical potential for a practical FBR. As a result, a result could be obtained that because of inferior fuel inventory and heat transmission to those in Na cooling reactor in present knowledge, mass of reactor vessel and intermediate heat exchanger were to widely increased to expect reduction of power generation unit price even on considering cheapness of its fuel cycle cost. Therefore, at present step further investigation on concept design of the chloride molten salt reactor plant system is too early in time, and it is at a condition where basic and elementary researches aiming at upgrading of economical efficiency such as wide reduction of fuel inventory, a measure expectable for remarkable rationalization effect of reprocessing system integrating a reactor to a processing facility, and so on. (G.K.)

  8. Fast reactor research in Switzerland

    International Nuclear Information System (INIS)

    Brogli, R.; Hudina, M.; Pelloni, S.; Sigg, B.; Stanculescu, A.

    1998-01-01

    The small Swiss research program on fast reactors serves to further understanding of the role of LMFR for energy production and to convert radioactive waste to more environmentally benign forms. These activities are on the one hand the contribution to the comparison of advanced nuclear systems and bring on the other to our physical and engineers understanding. (author)

  9. Development and Validation of a Momentum Integral Numerical Analysis Code for Liquid Metal Fast Reactor

    International Nuclear Information System (INIS)

    Chen, Xiangyi; Suh, Kune Y.

    2016-01-01

    In this work, this benchmark problem is conducted to assess the precision of the upgraded in-house code MINA. Comparison of the results from different best estimate codes employed by various grid spacer pressure drop correlations is carried out to suggest the best one. By modifying In's method, it presents good agreement with the experiment data which is shown in Figure 7. The reason for the failure of the prediction in previous work is caused by the utilization of Rehme's method which is categorized into four groups according to different fitting strategy. Through comparison of drag coefficients calculated by four groups of Rheme's method, equivalent drag coefficient calculated by In's method and experiment data shown in Figure 8, we can conclude that Rehme's method considerably underestimate the drag coefficients in grid spacers used in HELIOS and In's method give a reasonable prediction. Starting from the core inlet, the accumulated pressure losses are presented in figure 9 along the accumulated length of the forced convection flow path; the good agreement of the prediction from MINA with the experiment result shows MINA has very good capability in integrated momentum analysis makes it robust in the future design scoping method development of LFR.

  10. Development and Validation of a Momentum Integral Numerical Analysis Code for Liquid Metal Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xiangyi; Suh, Kune Y. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this work, this benchmark problem is conducted to assess the precision of the upgraded in-house code MINA. Comparison of the results from different best estimate codes employed by various grid spacer pressure drop correlations is carried out to suggest the best one. By modifying In's method, it presents good agreement with the experiment data which is shown in Figure 7. The reason for the failure of the prediction in previous work is caused by the utilization of Rehme's method which is categorized into four groups according to different fitting strategy. Through comparison of drag coefficients calculated by four groups of Rheme's method, equivalent drag coefficient calculated by In's method and experiment data shown in Figure 8, we can conclude that Rehme's method considerably underestimate the drag coefficients in grid spacers used in HELIOS and In's method give a reasonable prediction. Starting from the core inlet, the accumulated pressure losses are presented in figure 9 along the accumulated length of the forced convection flow path; the good agreement of the prediction from MINA with the experiment result shows MINA has very good capability in integrated momentum analysis makes it robust in the future design scoping method development of LFR.

  11. Review of fast reactor activities

    Energy Technology Data Exchange (ETDEWEB)

    Balz, W [Commission of the European Communities, Brussels (Belgium)

    1978-07-01

    The Commission of the European Communities continued its activities on the following lines: activities aimed at preparing for commercialization of fast breeder reactors which are essentially performed in the frame of Fast Reactor Coordinating Committee (FRCC); the execution of its own research program in the Joint Research Center. The report covers activities of the FRCC, of the Safety Working Group (SWG), the Whole Core Accident Code (WAC) subgroup, Containment (CONT) subgroup, Codes and Standards Working Group (CSWG). Research and development activities are concerned with LMFBR safety, subassembly thermal hydraulics, fuel-coolant interactions, post-accident heat removal, dynamic load response, safety related material properties, utilization limits of fast breeder fuels, plutonium and actinide aspects of nuclear fuel cycle.

  12. Review of fast reactor activities

    International Nuclear Information System (INIS)

    Balz, W.

    1978-01-01

    The Commission of the European Communities continued its activities on the following lines: activities aimed at preparing for commercialization of fast breeder reactors which are essentially performed in the frame of Fast Reactor Coordinating Committee (FRCC); the execution of its own research program in the Joint Research Center. The report covers activities of the FRCC, of the Safety Working Group (SWG), the Whole Core Accident Code (WAC) subgroup, Containment (CONT) subgroup, Codes and Standards Working Group (CSWG). Research and development activities are concerned with LMFBR safety, subassembly thermal hydraulics, fuel-coolant interactions, post-accident heat removal, dynamic load response, safety related material properties, utilization limits of fast breeder fuels, plutonium and actinide aspects of nuclear fuel cycle

  13. Status of national programmes on fast reactors

    International Nuclear Information System (INIS)

    1994-04-01

    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

  14. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  15. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985

  16. Fast reactor core monitoring device

    International Nuclear Information System (INIS)

    Sanda, Toshio; Inoue, Kotaro; Azekura, Kazuo.

    1982-01-01

    Purpose: To enable the rapid and accurate on-line identification of the state of a fast reactor core by effectively utilizing the measured data on the temperature and flow rate of the coolant. Constitution: The spacial power distribution and average assembly power are quickly calculated using an approximate calculating method, the measured values and the calculated values of the inlet and outlet temperature difference, flow rate and coolant physical values of an assembly are combined and are individually obtained, the most definite respective values and their errors are obtained by a least square method utilizing a formula of the relation between these values, and the power distribution and the temperature distribution of a reactor core are estimated in this manner. Accordingly, even when the measuring accuracy and the calculating accuracy are equal as in a fast reactor, the power distribution and the temperature distribution can be accurately estimated on-line at a high speed in a nuclear reactor, information required for the operator is provided, and the reactor can thus be safely and efficiently operated. (Yoshihara, H.)

  17. Introduction of the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Matsuba, Ken-ichi; Kawahara, Hirotaka; Aoyama, Takafumi

    2006-01-01

    The experimental fast reactor JOYO at O-arai Engineering Center of Japan Nuclear Cycle Development Institute is the first liquid metal cooled fast reactor in Japan. This paper describes the plant outline, experiences on the fast reactor technology and test results accumulated through twenty eight years successful operation of JOYO. (author)

  18. Fast reactor programme

    International Nuclear Information System (INIS)

    Hoekstra, E.K.

    1978-09-01

    Recently evaluated neutron capture cross section data of 142 143 144 145 146 150 Nd, natural Nd and Mo as well as adjustment of the capture cross sections of 152 154 Sm to fit integral data measured in STEK and CFRMF are discussed. The progress made with preparations for the HFR-TOP transient overpower experiments on fuel pins under irradiation in the pool-side facility of the HFR is reported on. Results are given of tensile tests on irradiated as well as on heat-treated stainless steel DIN 1.4948 specimens subjected to varying numbers of fatigue cycles. In the field of aerosol research, measured gas flow rates and pressure drops in stainless steel capillaries of various dimensions are compared with theory; the gas flow and aerosol penetration in cracks, artificially introduced in concrete test specimens, have been determined. Criteria in selecting the right light-scattering particles for use in Laser Doppler Anemometer measurements are given; the results of single and two-phase experiments with the second 28-rod bundle and the hydrodynamics during single-bubble boiling in the first bundle are discussed. (Auth.)

  19. EDF research on fast neutron reactors

    International Nuclear Information System (INIS)

    In order to make possible the calculation of the temperatures of the sodium, of the sheath and of the fuel in fast reactor assemblies, taking into account the mixing phenomena induced by the helicoidal wires, two design codes have been developed. Those codes have then been adapted for their integration in the Superalcyon system. This system shall constitute the reference tool for the development of those codes that shall manage Phenix, and other reactors of the family. Cooling accidents, thermohydraulic studies, and steam generator studies are also in progress

  20. The 'SURA' fast reactor program

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The Commissariat a l'Energie Atomique's SURA program on fast reactor safety consists of two specific testing programs on fastbreeder reactor safety: the Cabri and Scarabee programs. Both Cabri and Scarabee are examples of multinational research collaboration. The CEA and the Karlsruhe Nuclear Research Center are each covering half of the construction costs. Britain, the US and Japan are also due to participate in these experiments. The aim of the programs is to examine the behaviour of fuel in sodium cooled fast reactors. The Cabri program consists of setting off a reactivity accident in a power reactor core which is cooled with liquid sodium, such an accident occurring after a sharp increase in reactivity or as a result of the pump suddenly breaking down without there at the same time being any fall in the control rods. In 1967 the Commissariat a l'Energie Atomique started its Scarabee research program which is trying to analyse the sort of things that can go wrong with fuel cooling systems and what the consequences can be [fr

  1. 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,

  2. Breeding description for fast reactors and symbiotic reactor systems

    International Nuclear Information System (INIS)

    Hanan, N.A.

    1979-01-01

    A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

  3. Fast neutron nuclear reactor with lightened internal structure

    International Nuclear Information System (INIS)

    Artaud, R.; Aubert, M.; Renaux, C.

    1984-01-01

    The invention concerns an integrated type fast reactor. The inner vessel comprises two truncated shells, of which the large bases are connected either directly, or by a cylindrical shell of large diameter. The small base of the upper truncated shell is prolongated by a shell of small diameter and the small base of the lower truncated shell supports the reactor core. The invention allows the construction of simpler and less expansive fast reactors [fr

  4. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Wheeler, R.C.; Bramman, J.I.

    1988-04-01

    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)

  5. A fast spectrum dual path flow cermet reactor

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  6. Electrochemistry in fast reactor technology

    International Nuclear Information System (INIS)

    Mathews, C.K.

    1987-01-01

    Electrochemistry plays a significant role in the production, characterisation or behaviour of the fuel, the coolant and structural materials used in fast reactor systems. The role of electrochemistry in sodium production, in the fuel cycle, in the development of electrochemical meters used for the on-line monitoring of the various impurities at sub ppm levels and in the recovery of plutonium and uranium are discussed. The advantage of voltammmetric techniques in the analysis of impurities and the application of electrochemical meters have been investigated. (author). 5 figs., 15 refs

  7. Metrological certification of systems to monitor the seal integrity of fuel-element cladding based on exposed fuel in sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Eliseev, A.V.; Filonov, V.S.; Ushakov, V.M.; Belov, S.P.; Pedyash, B.V.; Zemtsev, B.V.; Skorikov, N.V.

    1992-01-01

    In sodium-cooled fast reactors, the clad monitoring system for seal integrity of the fuel element cladding is practically the only source of operator information on the serviceability of fuel elements in the core. The monitoring system can be used as the basis for critical decisions whether the reactor must be shut down of whether operation can continue, but only if the meterologically provided measurements are reliable. This article describes a method developed for certifying working rods on the basis of the domestic standard. The method includes a combined irradiation of the sample and the rod to be certified in an arbitrary field of a plutonium-beryllium neutron source with an output rate greater than 10 8 sec -1 , which is mounted in a paraffin moderator. The positive results of the metrological certification of the system to monitor cladding seal integrity leads the authors to recommend this method for other current and planned sodium-cooled fast reactors. 6 refs., 2 tabs

  8. Fast reactor fuel reprocessing. An Indian perspective

    International Nuclear Information System (INIS)

    Natarajan, R.; Raj, Baldev

    2005-01-01

    The Department of Atomic Energy (DAE) envisioned the introduction of Plutonium fuelled fast reactors as the intermediate stage, between Pressurized Heavy Water Reactors and Thorium-Uranium-233 based reactors for the Indian Nuclear Power Programme. This necessitated the closing of the fast reactor fuel cycle with Plutonium rich fuel. Aiming to develop a Fast Reactor Fuel Reprocessing (FRFR) technology with low out of pile inventory, the DAE, with over four decades of operating experience in Thermal Reactor Fuel Reprocessing (TRFR), had set up at the India Gandhi Center for Atomic Research (IGCAR), Kalpakkam, R and D facilities for fast reactor fuel reprocessing. After two decades of R and D in all the facets, a Pilot Plant for demonstrating FRFR had been set up for reprocessing the FBTR (Fast Breeder Test Reactor) spent mixed carbide fuel. Recently in this plant, mixed carbide fuel with 100 GWd/t burnup fuel with short cooling period had been successfully reprocessed for the first time in the world. All the challenging problems encountered had been successfully overcome. This experience helped in fine tuning the designs of various equipments and processes for the future plants which are under construction and design, namely, the DFRP (Demonstration Fast reactor fuel Reprocessing Plant) and the FRP (Fast reactor fuel Reprocessing Plant). In this paper, a comprehensive review of the experiences in reprocessing the fast reactor fuel of different burnup is presented. Also a brief account of the various developmental activities and strategies for the DFRP and FRP are given. (author)

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1977-01-01

    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

  10. Reactivity worth measurements on fast burst reactor Caliban - description and interpretation of integral experiments for the validation of nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Richard, B. [Commissariat a l' Energie Atomique et Aux Energies Alternatives CEA, DAM, VALDUC, F-21120 Is-sur-Tille (France)

    2012-07-01

    Reactivity perturbation experiments using various materials are being performed on the HEU fast core CALIBAN, an experimental device operated by the CEA VALDUC Criticality and Neutron Transport Research Laboratory. These experiments provide valuable information to contribute to the validation of nuclear data for the materials used in such measurements. This paper presents the results obtained in a first series of measurements performed with Au-197 samples. Experiments which have been conducted in order to improve the characterization of the core are also described and discussed. The experimental results have been compared to numerical calculation using both deterministic and Monte Carlo neutron transport codes with a simplified model of the reactor. This early work led to a methodology which will be applied to the future experiments which will concern other materials of interest. (authors)

  11. International Experience with Fast Reactor Operation & Testing

    International Nuclear Information System (INIS)

    Sackett, John I.; Grandy, C.

    2013-01-01

    Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

  12. Advanced Safeguards Approaches for New Fast Reactors

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  13. 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.

  14. The Argentine-Brazilian fast reactor programme

    International Nuclear Information System (INIS)

    Gho, C.J.; Mauricio, A.

    1989-01-01

    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

  15. Fast-reactor fuel reprocessing in the United Kingdom

    International Nuclear Information System (INIS)

    Allardice, R.H.; Buck, C.; Williams, J.

    1977-01-01

    Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the United Kingdom since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium-based fast-reactor system, and the importance of establishing at an early stage fast-reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high-burnup thermal-reactor oxide fuel. The United Kingdom therefore decided to reprocess irradiated fuel from the 250MW(e) Prototype Fast Reactor (PFR) as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small-scale fully active demonstration plant has been carried out since 1972, and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste-management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant, a parallel development programme has been initiated to provide the basis for the design of a large-scale fast-reactor fuel-reprocessing plant to come into operation in the late 1980s to support the projected UK fast-reactor installation programme. The paper identifies the important differences between fast-reactor and thermal-reactor fuel-reprocessing technologies and describes some of the development work carried out in these areas for the small-scale PFR fuel-reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast-reactor fuel-reprocessing plant is outlined and the current design philosophy discussed. (author)

  16. Holography for fast reactor inspection

    International Nuclear Information System (INIS)

    Tozer, B.A.

    1980-01-01

    Holography, an optical process whereby an image of the original subject can be reconstructed in three dimensions, is being developed for use as an optical inspection tool. With a potential information storage density of 10 16 bits/m 2 , the ability to reconstruct in 3 dimensions, a depth of field of up to 8 metres, extremely wide angle of view, and potentially diffraction limited resolution, holography should be invaluable for the optical recording of fast reactors during construction, and the inspection of optically accessible regions during operation, or maintenance down-times. The photographic emulsions used for high resolution holography are fine-grained and fog only very slowly when subjected to γ-radiation, so that inspection of highly radio-active regions and components can be effected satisfactorily. Some of the practical limitations affecting holography are described and ways of overcoming them discussed. Some preliminary results are presented. (author)

  17. Interfacial effects in fast reactors

    International Nuclear Information System (INIS)

    Saidi, M.S.; Driscoll, M.J.

    1979-05-01

    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

  18. The UK commercial demonstration fast reactor design

    International Nuclear Information System (INIS)

    Holmes, J.A.G.

    1987-01-01

    The paper on the UK Commercial Demonstration Fast Reactor design was presented to the seminar on 'European Commercial Fast Reactor Programme, London 1987. The design is discussed under the topic headings:- primary circuit, intermediate heat exchangers and pumps, fuel and core, refuelling, steam generators, and nuclear island layout. (U.K.)

  19. Aspects of the fast reactors fuel cycle

    International Nuclear Information System (INIS)

    Zouain, D.M.

    1982-06-01

    The fuel cycle for fast reactors, is analysed, regarding the technical aspects of the developing of the reprocessing stages and the fuel fabrication. The environmental impact of LMFBRs and the waste management of this cycle are studied. The economic aspects of the fuel cycle, are studied too. Some coments about the Brazilian fast reactors programs are done. (E.G.) [pt

  20. Euratom contributions in Fast Reactor research programmes

    International Nuclear Information System (INIS)

    Fanghänel, Th.; Somers, J.

    2013-01-01

    The Sustainable Nuclear Initiative: • demonstrate long-term sustainability of nuclear energy; • demonstration reactors of Gen IV: •more efficient use of resources; • closed fuel cycle; • reduced proliferation risks; • enhanced safety features. • Systems pursued in Europe: • Sodium-cooled fast reactor SFR; • Lead-cooled fast reactor LFR; • Gas-cooled fast reactor GFR. Sustainable Nuclear Energy Technology Platform SNE-TP promotes research, development and demonstration of the nuclear fission technologies necessary to achieve the SET-Plan goals

  1. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki

    1982-07-01

    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)

  2. Economic Issues of Fast Reactor in China

    International Nuclear Information System (INIS)

    Yang Hongyi

    2013-01-01

    Conclusions: 1. More and more fast reactors could be appearing in the world currently and near future. 2. China gets little experience and practice about the economics issues of sodium cooled fast reactors. 3. The economic issues become more and more important for the deplot of fast reactors. Suggestions: 1. An authoritative economic evaluation solution for fast reactor and related fuel cycles facilities is necessary. The solution may be developed by the interested country in order to share the few data, experience and methodology. 2. A new initiative to help to share the economic information for fast reactor and related fuel cycle facilities is necessary. A meeting like TM-44899 organized by the IAEA is very beneficial for this topic and hopefully will continue

  3. Status of fast reactor activities in Russia

    International Nuclear Information System (INIS)

    Poplavski, V.M.; Ashurko, Yu.M.; Zverev, K.V.

    1998-01-01

    This paper outlines state-of-the-art of the Russian nuclear power as of 1997 and its prospects for the nearest future. Results of the BR-10, BOR-60 and BN-600 reactors operation are described, as well as activity of the Russian institutions on scientific and technological support of the BN-350 reactor. Analysis of current status of the BN-800 reactor South-Urals NPP and Beloyarskaya NPP designs is given in brief, as well as prospects of their construction and possible ways of fast reactor technology improvement. Studies on fast reactors now under way in Russia are described. (author)

  4. General remarks on fast neutron reactor physics

    International Nuclear Information System (INIS)

    Barre, J.Y.

    1980-01-01

    The main aspects of fast reactor physics, presented in these lecture notes, are restricted to LMFBR's. The emphasis is placed on the core neutronic balance and the burn-up problems. After a brief description of the power reactor main components and of the fast reactor chronology, the fundamental parameters of the one-group neutronic balance are briefly reviewed. Then the neutronic burn-up problems related to the Pu production and to the doubling time are considered

  5. Pool type liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Guthrie, B.M.

    1978-08-01

    Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

  6. Fast reactor fuel design and development

    International Nuclear Information System (INIS)

    Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

    1977-01-01

    Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

  7. Safeguards challenges of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Ko, H. S.

    2010-01-01

    Although the safeguards system of Sodium Fast Reactor (SFR) seems similar to that of Light Water Reactor (LWR), it was raised safeguards challenges of SFR that resulted from the visual opacity of liquid sodium, chemical reactivity of sodium and other characteristics of fast reactor. As it is the basic concept stage of the safeguards of SFR in Korea, this study tried to analyze the latest similar study of safeguards issues of the Fast Breeder Reactor (FBR) at Joyo and Monju in Japan. For this reason, this study is to introduce some potential safeguards challenges of Fast Breeder Reactor. With this analysis, future study could be to address the safeguards challenges of SFR in Korea

  8. Power from plutonium: fast reactor fuel

    International Nuclear Information System (INIS)

    Bishop, J.F.W.

    1981-01-01

    Points of similarity and of difference between fast reactor fuel and fuels for AGR and PWR plants are established. The flow of uranium and plutonium in fast and thermal systems is also mentioned, establishing the role of the fast reactor as a plutonium burner. A historical perspective of fast reactors is given in which the substantial experience accumulated in test and prototype is indicated and it is noted that fast reactors have now entered the commercial phase. The relevance of the data obtained in the test and prototype reactors to the behaviour of commercial fast reactor fuel is considered. The design concepts employed in fuel are reviewed, including sections on core support styles, pin support and pin detail. This is followed by a discussion of current issues under the headings of manufacture, performance and reprocessing. This section includes a consideration of gel fuel, achievable burn-up, irradiation induced distortions and material choices, fuel form, and fuel failure mechanisms. Future development possibilities are also discussed and the Paper concludes with a view on the logic of a UK fast reactor strategy. (U.K.)

  9. Development of physical conceptions of fast reactors

    International Nuclear Information System (INIS)

    Khomyakov, Yu.S.; Matveev, V.I.; Moiseev, A.V.

    2013-01-01

    • Russian experience in developing fast reactors has proved clearly scientific justification of conceptual physical principles and their technical feasibility. • However, the potential of fast reactors caused by their physical features has not been fully realized. • In order to assure the real possibility of transition to the nuclear power with fast reactors by about 2030 it is necessary to consistently update fast reactor designs for solving the following key problems: - increasing of self-protection level of reactor core; - improvement of technical and economical characteristics; - solution of the problems related to the fuel supply of nuclear power and assimilation of closed nuclear fuel cycle; - disposal of long lived radioactive waste and transmutation of minor actinides. • Russian program (2010-2020) on the development of basic concepts of the new generation reactors implies successive solution of the above problems. • New technical decisions will be demonstrated by development and assimilation of the new reactors: - BN-800 – development of the fuel cycle infrastructure and mastering of the new types of fuel; - BN-1200 reactor – demonstration economical efficiency of fast reactor and new level of safety; - BREST development and demonstration new heavy liquid metal coolant technology and alternative design concept

  10. Some questions and answers concerning fast reactors

    International Nuclear Information System (INIS)

    Marshall, W.

    1980-01-01

    The theme of the lecture is the place of the fast reactor in an evolving nuclear programme. The whole question of plutonium is first considered, ie its method of production and the ways in which it can be used in the fast reactor fuel cycle. Whether fast reactors are necessary is then discussed. Their safety is examined with particular attention to those design features which are most criticised ie high volumetric power density of the core, and the use of liquid sodium as coolant. Attention is then paid to environmental and safeguard aspects. (U.K.)

  11. Fast breeder reactors--lecture 4

    International Nuclear Information System (INIS)

    Marshall, W.; Davies, L.M.

    1986-01-01

    This paper discusses the economics of fast breeder reactors. An algebraic background is presented which represents the various views expressed by different nations regarding the cost of fast breeder reactors and their associated fuel cycle services, the timescale by which they might be available, and the simultaneous variations in the price of uranium. Actual presentations made by individual countries in recent discussions serve to verify the general nature of this present discussion. It is assumed that if nuclear power is to make a long term contribution to the needs of the world, the introduction of fast breeder reactors is both essential and necessary

  12. 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

  13. U.S. Status of Fast Reactor Research and Technology

    International Nuclear Information System (INIS)

    Hill, Robert

    2012-01-01

    Summary: • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction) and safety: 1. System Integration and Concept Development; 2. Safety Technology; 3. Advanced Materials; 4. Ultrasonic Viewing; 5. Advanced Energy Conversion (Supercritical CO 2 Brayton cycle); 6. Reactor Simulation; 7. Nuclear Data; 8. Advanced Fuels. • Fast reactors have flexible capability for actinide management: – A wide variety of fuel cycle options are being considered; • International R&D collaboration pursued in Generation-IV and multilateral arrangements

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

    International Nuclear Information System (INIS)

    Bramman, J.I.; John, C.T.; Wheeler, R.C.

    1986-01-01

    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

  15. Concept and designs of new-generation fast reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.

    1993-01-01

    This article discusses the general safety requirements and characteristics for future nuclear power plants. It examines various designs - loop, block, and integrated layouts for reactors. Specifically, the article focuses an integrated design for sodium-cooled fast reactors noting that the BN-600 reactor has operated accident-free over the past 12 years. An obvious advantage of this scheme is that the coolant of the primary loop is localized in one volume (in a vessel), there are no short connections and large-diameter pipes, which of course sharply reduces the probability in coolant leaks. With an integrated scheme the problem of embrittlement of the reactor vessel by neutron irradiation is obviated. The neutron fluence for the vessels of the AST-500 and VPBER-600 reactors, built with an integrated scheme, is less than 10 17 cm -2 . Such a fluence does not cause any appreciable change in the mechanical properties of the vessel steel. The integrated layout of the reactor makes it possible to build a containment vessel. In this case it is possible to eliminate the danger of the reactor core drying out and thus cooling of the reactor in emergency situations can be simplified substantially. In an integrated layout, however, access is more difficult to the equipment inside the reactor, thus limiting or complicating maintenance work. The integrated layout, therefore, requires the use of highly reliable equipment built according to designs that have been proven in operation and have been passed representative service-life tests under laboratory conditions. The integrated layout considerably increases the mass and size characteristics of the reactor. New solutions thus are needed for the organization of work on reactor fabrication and assembly. In the case of the BN-600 and Superphenix reactors the welding of the reactor vessels and the assembly work were done on the building site

  16. Report of investigation into allegations of retaliation for raising safety and quality of work issues regarding Argonne National Laboratory's Integral Fast Reactor project

    International Nuclear Information System (INIS)

    1991-12-01

    In August 1990 James A. Smith resigned his position as an experimenter at Argonne National Laboratory-West (ANL-W), located near Idaho Falls, Idaho. Smith who holds a Ph.D. in metallurgy, had worked at the Laboratory since 1988, primarily on its Integral Fast Reactor (IFR) project. He alleged that the quality of the Laboratory's work on that project had been undermined by fundamental errors in metallurgy and related sciences, at least some of which had nuclear safety implications; that the Laboratory had published false and misleading accounts of its work; that prevailing attitudes at the Laboratory were antithetical to quality scientific work; and that because he had expressed concerns about these matters his job was threatened by his managers. Evidence gathered during an investigation by the Department of Energy's Office of Nuclear Safety (NS) is presented and conclusions and recommendations are provided

  17. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  18. Decommissioning of fast reactors after sodium draining

    International Nuclear Information System (INIS)

    2009-11-01

    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

  19. The energy gap and the fast reactor

    International Nuclear Information System (INIS)

    Hill, J.

    1977-01-01

    The background to the development of fast reactors is summarized. In Britain, the results of the many experiments performed, the operation of the Dounreay Fast Reactor for the past 18 years and the first year's operation of the larger Prototype Fast Reactor have all been very encouraging, in that they demonstrated that the performance corresponded well with predictions, breeding is possible, and the system is exceptionally stable in operation. The next step in fast reactor engineering is to build a full-scale fast reactor power station. There would seem to be little reason to expect more trouble than could reasonably be expected in constructing any large project of this general nature. However, from an engineering point of view continuity of experience is required. If a decision to build a commercial fast reactor were taken today there would be a 14-year gap between strating this and the start of the Prototype Fast Reactor. This is already much too long. From an environmental standpoint we have to demonstrate that we can manufacture and reprocess fast reacctor fuel for a substantial programme in a way that does not lead to pollution of the environment, and that plutonium-containing fuel can be transported in the quantities required in safety and in a way that does not attract terrorists or require a private army to ensure its security. Finally, we have to find a way to allow many countries to obtain the energy they need from fast reactors, without leading to the proliferation of nuclear weapons or weapons capability. (author)

  20. Optimal reactor strategy for commercializing fast breeder reactors

    International Nuclear Information System (INIS)

    Yamaji, Kenji; Nagano, Koji

    1988-01-01

    In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

  1. The fast reactor and energy supply

    International Nuclear Information System (INIS)

    1979-01-01

    The progress made with fast reactor development in many countries is summarised showing that the aim is to provide to the nation concerned an ability to instal fast reactor power stations at the end of this century or early in the next one. Accepting the importance of fast reactors as a potential independent source of energy, problems concerning economics, industrial capability, technical factors, public acceptibility and in particular plutonium management, are discussed. It is concluded that although fast reactors have reached a comparatively advanced stage of development, a number of factors make it likely that their introduction for electricity generation will be a gradual process. Nevertheless it is necessary to complete demonstration and development phases in good time. (U.K.)

  2. Economic evaluation of fast reactor fuel cycling

    International Nuclear Information System (INIS)

    Hu Ping; Zhao Fuyu; Yan Zhou; Li Chong

    2012-01-01

    Economic calculation and analysis of two kinds of nuclear fuel cycle are conducted by check off method, based on the nuclear fuel cycling process and model for fast reactor power plant, and comparison is carried out for the economy of fast reactor fuel cycle and PWR once-through fuel cycle. Calculated based on the current price level, the economy of PWR one-through fuel cycle is better than that of the fast reactor fuel cycle. However, in the long term considering the rising of the natural uranium's price and the development of the post treatment technology for nuclear fuels, the cost of the fast reactor fuel cycle is expected to match or lower than that of the PWR once-through fuel cycle. (authors)

  3. Fast reactors - Dounreay and the future

    International Nuclear Information System (INIS)

    Jordan, G.

    1988-01-01

    In 1960 at Dounreay, the Dounreay Fast Reactor (DFR) supplied the world's first fast reactor grid electricity, and went on to a highly successful career as a test facility, as fuel designs advanced. In the 1960s, the Prototype Fast Reactor (PFR) was designed and built, beginning operation in 1974. The PFR was built to provide a sound technical and experienced base to support the UK's future Fast Reactor development and design. The in-vessel fuel handling facilities have demonstrated the flexibility of the pool design and a considerable body of in-core fuel handling experience is available. A key issue for further Fast Reactor application is the performance of fuel and, because PFR was designed to take full-scale fuel assemblies, the fuel performance experience is directly relevant to commercial designs. The original PFR design irradiation target of 60000 MWd/t U (equivalent to 7.5 % burn-up) has already been exceeded by a factor of more than two and a 15.9 % burn-up sub-assembly has been discharged and reprocessed without difficulty. Soon a 20 % sub-assembly will follow. Also the PFR reprocessing plant has demonstrated the safety and efficiency of this essential adjunct to Fast Reactor operation. The safety and the environmental protection features of both the PFR and its fuel reprocessing plant have been demonstrated over the last 14 years. 2 refs., 3 figs

  4. Advanced liquid metal fast breeder reactor designs

    International Nuclear Information System (INIS)

    Sayles, C.W.

    1978-01-01

    Fast Breeder reactor power plants in the 1000-1200 MW(e) range are being built overseas and are being designed in this country. While these reactors have many characteristics in common, a variety of different approaches have been adopted for some of the major features. Some of those alternatives are discussed

  5. A glossary of terms for fast reactors

    International Nuclear Information System (INIS)

    Wheeler, R.C.

    1979-04-01

    The glossary aims to provide definitions of technical terms likely to be used in a fast reactor enquiry and to encourage the use of the same set of consistent terms in any documents intended for such an inquiry. In some cases definitions are formulated in the limited context of LMFBRS rather than applying to all types of reactors. A brief guide is presented to the different reactor types. (author)

  6. The fast breeder reactor Rapsodie (1962)

    International Nuclear Information System (INIS)

    Vautrey, L.; Zaleski, C.P.

    1962-01-01

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors) [fr

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

    International Nuclear Information System (INIS)

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

    2012-01-01

    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.

  8. 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.

  9. Review of Fast Reactor Activities, March 1980

    International Nuclear Information System (INIS)

    Balz, W.

    1980-01-01

    As in previous years, a short outline of the major achievements made since the last IWGFR meeting is given in the following. On 18 February 1980 the Council of Ministers has approved a resolution in which they recognise the strategic importance of fast breeder reactors and the need to continue the efforts towards maintaining an effective fast breeder option in the Member States

  10. The dismantling of fast reactors: sodium processing

    International Nuclear Information System (INIS)

    Rodriguez, G.; Berte, M.; Serpante, J.P.

    1999-01-01

    Fast reactors require a coolant that does not slow down neutrons so water can not be used. Metallic sodium has been chosen because of its outstanding neutronic and thermal properties but sodium reacts easily with air and water and this implies that sodium-smeary components can not be considered as usual nuclear wastes. A stage of sodium neutralizing is necessary in the processing of wastes from fast reactors. Metallic sodium is turned into a chemically stable compound: soda, carbonates or sodium salts. This article presents several methods used by Framatome in an industrial way when dismantling sodium-cooled reactors. (A.C.)

  11. Fast reactor research activities in Brazil

    International Nuclear Information System (INIS)

    Menezes, A.

    1998-01-01

    Fast reactor activities in Brazil have the objective of establishing a consistent knowledge basis which can serve as a support for a future transitions to the activities more directly related to design, construction and operation of an experimental fast reactor, although its materialization is still far from being decided. Due to the present economic difficulties and uncertainties, the program is modest and all efforts have been directed towards its consolidation, based on the understanding that this class of reactors will play an important role in the future and Brazil needs to be minimally prepared. The text describes the present status of those activities, emphasizing the main progress made in 1996. (author)

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

    International Nuclear Information System (INIS)

    Koch, M.; Kazimi, M.S.

    1991-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

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

  14. Stationary Liquid Fuel Fast Reactor

    International Nuclear Information System (INIS)

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

    2015-01-01

    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

  15. 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

  16. Calculation of the neutron parameters of fast thermal reactor

    International Nuclear Information System (INIS)

    Kukuleanu, V.; Mocioiu, D.; Drutse, E.; Konstantinesku, E.

    1975-01-01

    The system of neutron calculation for fast reactors is given. This system was used for estimation of physical parameters of fast thermal reactors investigated. The results obtained and different specific problems of the reactors of this type are described. (author)

  17. Strategies for minority actinides transmutation in fast reactors

    International Nuclear Information System (INIS)

    Perez-Martin, S.; Martin-Fuertes, F.; Alvarez-Velarde, F.

    2010-01-01

    Presentation of the strategies that can be followed in fast reactors designed for the fourth generation to reduce the inventory of minority actinides generated in current light water reactors, as the actinides generation in fast reactor.

  18. Actinides burnup in a sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Pineda A, R.; Martinez C, E.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    The burnup of actinides in a nuclear reactor is been proposed as part of an advanced nuclear fuel cycle, this process would close the fuel cycle recycling some of the radioactive material produced in the open nuclear fuel cycle. These actinides are found in the spent nuclear fuel from nuclear power reactors at the end of their burnup in the reactor. Previous studies of actinides recycling in thermal reactors show that would be possible reduce the amounts of actinides at least in 50% of the recycled amounts. in this work, the amounts of actinides that can be burned in a fast reactor is calculated, very interesting results surge from the calculations, first, the amounts of actinides generated by the fuel is higher than for thermal fuel and the composition of the actinides vector is different as in fuel for thermal reactor the main isotope is the {sup 237}Np in the fuel for fast reactor the main isotope is the {sup 241}Am, finally it is concluded that the fast reactor, also generates important amounts of waste. (Author)

  19. A worldwide survey of fast breeder reactors

    International Nuclear Information System (INIS)

    Hennies, H.H.

    1986-01-01

    While the completion of the SNR 300 was accompanied by manifold discussions on questions relevant to safety and energy policies in the Federal Republic of Germany and as a result considerable scheduling delays and exceeding of budgets were recorded, breeder reactor technology has been progressing worldwide. The transition from the development phase with small trial reactors to the construction and operation of large performance reactors was completed systematically, in particular in France and the Soviet Union. Even though the uranium supply situation does not make a short-term and comprehensive employment of fast breeder reactors essential, technology has meanwhile been advanced to such a level and extensive operating experience is on hand to enable the construction and safe operation of fast breeder reactors. A positive answer has long been found to the question of the realization of a breeding rate to guarantee the breeding effect. There remain now the endeavors to achieve a reduction in investment and fuel cycle costs. (orig.) [de

  20. Nuclear Burning Wave Modular Fast Reactor Concept

    International Nuclear Information System (INIS)

    Kodochigov, N.G.; Sukharev, Yu.P.

    2014-01-01

    The necessity to provide nuclear power industry, comparable in a scope with power industry based on a traditional fuel, inspired studies of an open-cycle fast reactor aimed at: - solution of the problem of fuel provision by implementing the highest breeding characteristics of new fissile materials of raw isotopes in a fast reactor and applying accumulated fissile isotopes in the same reactor, independently on a spent fuel reprocessing rate in the external fuel cycle; - application of natural or depleted uranium for makeup fuel, which, with no spent fuel reprocessing, forms the most favorable non-proliferation conditions; - application of inherent properties of the core and reactor for safety provision. The present report, based on previously published papers, gives the theoretical backgrounds of the concept of the reactor with a nuclear burning wave, in which an enriched-fuel core (driver) is replaced by a blanket, and basic conditions for nuclear burning wave initiating and keeping are shown. (author)

  1. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    , mainly pressurized heavy water reactors (PHWRs) to .... plug housing 12 absorber rod drive mechanisms is supported on ... state-of-art erection equipments and construction methodologies and .... This decision is taken after.

  2. Fast reactors and problems in their development. Chapter 6

    International Nuclear Information System (INIS)

    Dombey, N.

    1980-01-01

    The main differences between fast reactors, in particular the liquid-metal fast breeder reactor (LMFBR), and thermal reactors are discussed. The view is taken, based on the intrinsic physics of the systems, that fast reactors should be considered as a different genus from thermal reactors. Some conclusions are drawn for fast reactor development generally and for the British programme in particular. Physics, economics and safety aspects are covered. (U.K.)

  3. European ERANOS formulaire for fast reactor core analysis

    International Nuclear Information System (INIS)

    Rimpault, Gerald

    2003-01-01

    ERANOS code scheme was developed within the European collaboration on fast reactors. It contains all the functions required to calculate a complete set of core, shielding and fuel cycle parameters for LMFR cores. Nuclear data are taken from recent evaluations (JEF2.2) and adjusted on integral experiments (ERALIB1). Calculational scheme uses the ECCO cell code to generate cross section data. Whole core calculations are carried out using the spatial modules BISTRO (Sn) and TGVNARIANT (nodal method). Validation is based on integral and power reactor experiments. Integral experiments are also used for adjustment of nuclear data

  4. Utilization of large electromagnetic pumps in the fast breeder reactors

    International Nuclear Information System (INIS)

    Deverge, C.; Lefrere, J.P.; Peturaud, P.; Sauvage, M.

    1984-04-01

    After an overview concerning the induction annular electromagnetic pumps and the dimensioning methods usually utilized, development of these components for a fast breeder integrated reactor is considered: - utilization of cooled EMP in the intermediate circuit, - utilization of immersed pumps, coupled with the intermediate exchanger, for the primary pumping; dimensioning, energetic aspects, and effects on the power plant geometrical configurations [fr

  5. The safety of the fast reactor

    International Nuclear Information System (INIS)

    Matthews, R.R.

    1977-01-01

    Verbatim of an address by R.R. Matthews, Chief Nuclear Health and Safety Officer, UK Central Electricity Generating Board given on January 15th 1977. The object of this address was to give some opinions on the safety issues of fast reactors as seen from an operational point of view. An outline of the basic responsibilities for nuclear safety is first given, and it is emphasized that the Central Electricity Generating Board has a statutory responsibility for the safe operation of its nuclear plant. The Nuclear Installations Act places absolute responsibility on the operator for ensuring that injury to persons and damage to property do not occur, and the new Health and Safety at Work Act does likewise. In addition the Board has a Nuclear Health and Safety Department that has to ensure that adequate provision for safety is made in the design, construction, and operation of nuclear plant, and safety at operational stations is monitored continuously by inspectors. In addition the requirements of the Nuclear Installations Inspectorate, laid down in the site licence conditions, must be satisfied. All these requirements are here discussed in the light of application to commercial fast reactors. It is considered that the hazards to fast reactor operating personnel are small and little different from those of other types of reactor, and in some respects the fast reactor has advantages, particularly in regard to the use of a Na coolant. The possibility of various types of accident is considered. Radioactive effluent discharge is also considered. The fast reactor as an international problem is discussed, including security matters. The extensive experience gained in operation of the experimental and prototype fast reactors at Dounreay is emphasized. (U.K.)

  6. Fast Reactor Programme. Second Quarter 1969. Progress Report. RCN Report

    International Nuclear Information System (INIS)

    Hoekstra, E.K.

    1969-12-01

    This progress report covers fast reactor research carried out by RCN during the second quarter 1969 forming part of the integrated fast breeder research and development programme also in progress at the national nuclear research centres Karlsruhe and Mol. The combined effort is based on a memorandum of co-operation in this field signed by the respective governments in 1968 and on a memorandum of understanding signed by the research centres. The RCN contribution is mainly concerned with the core of the fast breeder reactor and related safety aspects and, as such, must be looked upon as being complementary to the industrial research pro- field of fast reactors. The contribution comprises the following six items: - A Æéatîtôr , physics programme to determine the influence of fission products on several main characteristics of the reactor core such as void coefficient, Doppler coefficient and breeding ratio; - A fuel performance programme in which both stationary and transient irradiations are being carried out to establish the temperature and power limits of fuel rods; also the consequences of loss- of-cooling will be investigated; - Investigation into the change in mechanical properties of fuel canning materials due to high fast neutron doses; - A study of the corrosion behaviour of canning materials and their compatibility with the fuel under conditions of high temperature and high pressure; - Investigation into the behaviour of aerosols of fission products which could be formed after a fast reactor accident; a thorough understanding is of utmost importance for the reactor safety assessment ; - Studies on heat transfer in the reactor core. As fast breeders operate at high power densities, an accurate knowledge on the heat transfer phenomena is required

  7. A new neutron noise technique for fast reactors

    International Nuclear Information System (INIS)

    Zhuo Fengguan; Jin Manyi; Yao Shigui; Su Zhuting

    1987-12-01

    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

  8. Overview of the fast reactors fuels program

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

  9. Fast breeder reactor safety : a perspective

    International Nuclear Information System (INIS)

    Kale, R.D.

    1992-01-01

    Taking into consideration India's limited reserves of natural and vast reserves of thorium, the fast reactor route holds a great promise for India's energy supply in future. The fast reactor fueled with 239 Pu/ 238 U (unused or depleted) produces (breeds) more fissionable fuel material 239 Pu than it consumes. Calculations show that a fast breeder reactor (FBR) increases energy potential of natural uranium by about 60 times. As the fast reactor can also convert 232 Th into 233 U which is a fissionable material, it can make India's thorium reserves a source of almost inexhaustible energy supply for a long time to come. Significant advantage of FBR plants cooled by sodium and their world-wide operating experience are reviewed. There are two main safety issues of FBR, one nuclear and the other non-nuclear. The nuclear issue concerns core disruptive accident and the non-nuclear one concerns the high chemical energy potential of sodium. These two issues are analysed and it is pointed that they are manageable by current design, construction and operational practices. Main findings of safety research during the last six to eight years in West European Countries and United States of America (US) are summarised. Three stage engineered safety provision incorporated into the design of the sodium cooled Fast Breeder Test Reactor (FBTR) commissioned at Kalpakkam are explained. The important design safety features of FBTR such as primary system containment, emergency core cooling, plant protection system, inherent safety features achieved through reactivity coefficients, and natural convection cooling are discussed. Theoretical analysis and experimental research in fast reactor safety carried out at the Indira Gandhi Centre for Atomic Research during the past some years are reviewed. (M.G.B.)

  10. Fast-acting nuclear reactor control device

    International Nuclear Information System (INIS)

    Kotlyar, O.M.; West, P.B.

    1993-01-01

    A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position

  11. Fast neutron reactors: the safety point of view

    International Nuclear Information System (INIS)

    Laverie, M.; Avenas, M.

    1984-01-01

    All versions of nuclear reactors present favourable and unfavourable characteristics from the point of view of safety. The safety of the installations is obtained by making efforts to utilize in the best possible way those which are favourable and by taking proper steps in the face of those which are unfavourable. The present article shows how this general principle has been applied as regards the fast neutron reactors of integrated design which have been developped in France, taking into account the specific features of this version. A qualitative method to compare the safety of this version with that of pressurized water reactors which has been widely put to the test commercially all over the world is presented. These analyses make, generally speaking, several positive characteristics stand out for these fast neutron reactors from the safety aspects [fr

  12. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

    Parcy, J.P.

    1982-09-01

    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 [fr

  13. Fast reactor development programme in France

    Energy Technology Data Exchange (ETDEWEB)

    Le Rigoleur, C [Direction des Reacteurs Nucleaires, CEA Centre d` Etudes de Cadarache, Saint-Paul-lez-Durance (France)

    1998-04-01

    First the general situation regarding production of electricity in France is briefly described. Then in the field of Fast Reactors, the main events of 1996 are presented. At the end of February 1996, the PHENIX reactor was ready for operation. After review meetings, the Safety Authority has requested safety improvements and technical demonstrations, before it examines the possibility of authorizing a new start-up of PHENIX. The year 1996 was devoted to this work. In 1996, SUPERPHENIX was characterized by excellent operation throughout the year. The reactor was restarted at the end of 1995 after a number of minor incidents. The reactor power was increased by successive steps: 30% Pn up to February 6, followed by 50% Pn up to May then 60% up to October and 90% Pn during the last months. A programmed shutdown period occurred during May, June and mid-July 1996. The reactor has been shutdown at the end of 1996 for the decenial control of the steam generators. The status of the CAPRA project, aimed at demonstrating the feasibility of a fast reactor to burn plutonium at as high a rate as possible and the status of the European Fast Reactor are presented as well as their evolution. Finally the R and D in support of the operation of PHENIX and SUPERPHENIX, in support of the ````knowledge-acquisition```` programme, and CAPRA and EFR programmes is presented, as well as the present status of the stage 2 dismantling of the RAPSODIE experimental fast reactor. (author). 4 refs, figs, 2 tabs.

  14. Current status of fast reactor physics

    International Nuclear Information System (INIS)

    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

  15. Fast Reactor Knowledge Management at IGCAR, India

    International Nuclear Information System (INIS)

    Kuriakose, K.K.

    2013-01-01

    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

  16. Slow clean-up for fast reactor

    Science.gov (United States)

    Banks, Michael

    2008-05-01

    The year 2300 is so distant that one may be forgiven for thinking of it only in terms of science fiction. But this is the year that workers at the Dounreay power station in Northern Scotland - the UK's only centre for research into "fast" nuclear reactors - term as the "end point" by which time the site will be completely clear of radioactive material. More than 180 facilities - including the iconic dome that housed the Dounreay Fast Reactor (DFR) - were built at at the site since it opened in 1959, with almost 50 having been used to handle radioactive material.

  17. Discharges from a fast reactor reprocessing plant

    International Nuclear Information System (INIS)

    Barnes, D.S.

    1987-01-01

    The purpose of this paper is to assess the environmental impact of the calculated routine discharges from a fast reactor fuel reprocessing plant. These assessments have been carried out during the early stages of an evolving in-depth study which culminated in the design for a European demonstration reprocessing plant (EDRP). This plant would be capable of reprocessing irradiated fuel from a series of European fast reactors. Cost-benefit analysis has then been used to assess whether further reductions in the currently predicted routine discharges would be economically justified

  18. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Smith, R.D.

    1982-01-01

    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

  19. Universal Fast Breeder Reactor Subassembly Counter manual

    International Nuclear Information System (INIS)

    Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

    1984-08-01

    A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications

  20. The behaviour of materials in fast reactors

    International Nuclear Information System (INIS)

    Matthews, J.R.

    1977-01-01

    Fast neutron damage in fast reactors can limit the life of structural components through the growth voids. The main features of the current theory of point defect production and condensation are surveyed. The role of metallurgical structures and radiation produced extended defects is outlined and used to demonstrate the development of volume swelling and radiation hardening. Mechanisms of radiation creep are described in the context of the preceding treatment of point defect behaviour. Finally, future trends in the field are briefly explored. (author)

  1. Universal Fast Breeder Reactor Subassembly Counter manual

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

    1984-08-01

    A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications.

  2. A study of reactor vessel integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

    1999-02-15

    The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

  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)

    Neil Todreas; Pavel Hejzlar

    2008-01-01

    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. Small size modular fast reactors in large scale nuclear power

    International Nuclear Information System (INIS)

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

    2005-01-01

    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)

  6. Liquid metal fast reactor transient design

    International Nuclear Information System (INIS)

    Horak, C.; Purvis, E. III

    2000-01-01

    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)

  7. Thermophysical properties of fast reactor fuel

    International Nuclear Information System (INIS)

    Fink, J.K.

    1984-01-01

    This paper identifies the fuel properties for which more data are needed for fast-reactor safety analysis. In addition, a brief review is given of current research on the vapor pressure over liquid UO 2 and (U,PU)O/sub 2-x/, the solid-solid phase transition in actinide oxides, and the thermal conductivity of molten urania

  8. Charging machine for a fast production reactor

    International Nuclear Information System (INIS)

    Artem'ev, L.N.; Kurilkin, V.V.

    1971-01-01

    Charging machine for a fast production reactor is described. The machine contains charging mechanism, mechanism for positioning fresh fuel and spent fuel assemtlies, storage drums with sockets for control rod assemtlies and collet tongs for control rods. Recharging is conducted by means of ramp channel

  9. Fast breeder reactor at Kalkar. Pt. 2

    International Nuclear Information System (INIS)

    Degen, G.

    1979-02-01

    After a brief description of the previous development of the case the legal decisions are documented and commented on. The concept of the then FDP-Minister of Economy of North Rhine Westphalia (Riemer, Pu-combustion plant) is presented and the prospects and risk for the fast breeder reactor after the 3. partial construction license are discussed. (orig./HP) [de

  10. Use of fast reactors for actinide transmutation

    International Nuclear Information System (INIS)

    1993-03-01

    The management of radioactive waste is one of the key issues in today's discussions on nuclear energy, especially the long term disposal of high level radioactive wastes. The recycling of plutonium in liquid metal fast breeder reactors (LMFBRs) 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. The International Working Group on Fast Reactors (IWGFR) decided to include the topic of actinide transmutation in liquid metal fast breeder reactors in its programme. The IAEA organized the Specialists Meeting on Use of Fast Breeder Reactors for Actinide Transmutation in Obninsk, Russian Federation, from 22 to 24 September 1992. The specialists agree that future progress in solving transmutation problems could be achieved by improvements in: Radiochemical partitioning and extraction of the actinides from the spent fuel (at least 98% for Np and Cm and 99.9% for Pu and Am isotopes); technological research and development on the design, fabrication and irradiation of the minor actinides (MAs) containing fuels; nuclear constants measurement and evaluation (selective cross-sections, fission fragments yields, delayed neutron parameters) especially for MA burners; demonstration of the feasibility of the safe and economic MA burner cores; knowledge of the impact of maximum tolerable amount of rare earths in americium containing fuels. Refs, figs and tabs

  11. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  12. The development of fast reactors in France

    International Nuclear Information System (INIS)

    Vautrey, L.

    1982-01-01

    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)

  13. Status of fast reactor activities in Brazil

    International Nuclear Information System (INIS)

    Menezes, Artur

    1996-01-01

    This text describes the present status of fast reactor activities in Brazil, emphasizing the strategies being used to preserve this reactor concept as a viable alternative for future electricity generation in the country. The program is mostly research-oriented and has the objective of establishing a consistent knowledge basis which can serve as a support for the transition to the activities more directly related to design, construction and operation of an experimental fast reactor. Due to the present economic difficulties, the program is still modest but it is gradually growing. A report which has been finalized in December, 1995 and submitted to the authorities indicates the existence of the grounds for enlarging and consolidating the program. (author)

  14. Nuclear data for advanced fast reactors

    International Nuclear Information System (INIS)

    Rabotnov, N.S.

    2001-01-01

    Interest revives to fast reactors as the only proven technology obviously able of satisfying human energy needs for the next millennium by using full energy content of both natural uranium resources and of vast stocks of depleted uranium. This interest stimulates revision and improvement of fast reactor ND. Progress in reactor calculations accuracy due to better codes and much faster computers also increases relative importance of the input data uncertainties, especially in case of small reactivity margin and fuels of equilibrium compositions. The main objects of corresponding R and D efforts should be minor actinides and heavy liquid metal coolant. Data error bands and covariance information also gain importance as necessary components of neutron physics calculations. (author)

  15. Fast breeder reactor electromagnetic pump

    International Nuclear Information System (INIS)

    Araseki, Hideo; Murakami, Takahiro

    2008-01-01

    Main pumps circulating sodium in the FBR type reactor have been mechanical types, not electromagnetic pumps. Electromagnetic pump of 1-2 m 3 /min has been used as an auxiliary pump. Large sized electromagnetic pumps such as several hundred m 3 /min have not been commercialized due to technical difficulties with electromagnetic instability and pressure pulsations. This article explained electromagnetic and fluid equations and magnetic Reynolds number related with electromagnetic pumps and numerical analysis of instability characteristics and pressure pulsations and then described applications of the results to FBR system. Magnetic Reynolds number must be chosen less than one with appropriate operating frequency and optimum slip of 0.2-0.4. (T. Tanaka)

  16. Integrity assessment of the ferritic / austenitic dissimilar weld joint between intermediate heat exchanger and steam generator in fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, T.; Laha, K.; Chandravathi, K. S.; Parameswaran, P.; Goyal, S.; Kumar, J. G.; Mathew, M. D. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam- 603 102 (India)

    2012-07-01

    Integrity of the modified 9Cr-1Mo / alloy 800 dissimilar joint welded with Inconel 182 electrodes has been assessed under creep condition based on the detailed analysis of microstructure and stress distribution across the joint by finite element analysis. A hardness peak at the ferritic / austenitic weld interface and a hardness trough at the inter-critical heat affected zone (HAZ) in ferritic base metal developed. Un-tempered martensite was found at the ferritic / austenitic weld interface to impart high hardness in it; whereas annealing of martensitic structure of modified 9Cr-1Mo steel by inter-critical heating during welding thermal cycle resulted in hardness tough in the inter-critical HAZ. Creep tests were carried out on the joint and ferritic steel base metal at 823 K over a stress range of 160-320 MPa. The joint possessed lower creep rupture strength than its ferritic steel base metal. Failure of the joint at relatively lower stresses occurred at the ferritic / austenitic weld interface; whereas it occurred at inter-critical region of HAZ at moderate stresses. Cavity nucleation associated with the weld interface particles led to premature failure of the joint. Finite element analysis of stress distribution across the weld joint considering the micro-mechanical strength inhomogeneity across it revealed higher von-Mises and principal stresses at the weld interface. These stresses induced preferential creep cavitation at the weld interface. Role of precipitate in enhancing creep cavitation at the weld interface has been elucidated based on the FE analysis of stress distribution across it. (authors)

  17. Scenario for commercialization of fast breeder reactors

    International Nuclear Information System (INIS)

    Kumaoka, Yoshio; Sato, Morihiko

    1989-01-01

    To realize the commercialization of fast breeder reactors (FBRs), it is essential to reduce construction costs to the same level as those for the current light water reactors. For this target to be attained, a highly important factor is to reduce to the lowest-levels possible the quantities of materials and volume of the buildings required for the primary and secondary sodium loops of the FBR. In this direction, an innovative compact FBR plant concept which holds promise for commercialization has been developed by introducing the pooltype reactor concept with the shortest possible secondary sodium loops, realized by coupling electromagnetic pumps with the steam generators. In comparison with the French Super Phenix reactor, for example, the construction of this 1,300-MWe FBR plant could be achieved with half the material quantities and plant volume required by the former type. (author)

  18. Conceptual design of reactor assembly of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

    1996-01-01

    The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

  19. 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.

  20. Intrinsically secure fast reactors with dense cores

    International Nuclear Information System (INIS)

    Slessarev, Igor

    2007-01-01

    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: ·Ultra-plate 'dense cores' of the ordinary (monolithic) type with negative total coolant void effects. ·Modular type cores. Multiple dense modules can be embedded in the common reflector for achieving the desired NPP total power. The modules can be used also independently (as

  1. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  2. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

    Bourgeois, M.; Le Bouhellec, J.; Eymery, R.; Viala, M.

    1984-08-01

    Simultaneous with the effort on fast breeder reactors launched several years ago in France, equivalent investigations have been conducted on the fuel cycle, and in particular on reprocessing, which is an indispensable operation for this reactor. The Rapsodie experimental reactor was associated with the La Hague reprocessing plant AT1 (1 kg/day), which has reprocessed about one ton of fuel. The fuel from the Phenix demonstration reactor is reprocessed partly at the La Hague UP2 plant and partly at the Marcoule pilot facility, undergoing transformation to reprocess all the fuel (TOR project, 5 t/y). The fuel from the Creys Malville prototype power plant will be reprocessed in a specific plant, which is in the design stage. The preliminary project, named MAR 600 (50 t/y), will mobilize a growing share of the CEA's R and D resources, as the engineering needs of the UP3 ''light water'' plant begins to decline. Nearly 20 tonnes of heavy metals irradiated in fast breeder reactors have been processed in France, 17 of which came from Phenix. The plutonium recovered during this reprocessing allowed the power plant cycle to be closed. This power plant now contains approximately 140 fuel asemblies made up with recycled plutonium, that is, more than 75% of the fuel assemblies in the Phenix core

  3. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Johnson, R.N.

    1984-04-01

    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

  4. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Kapoor, R.P.; Srinivasan, G.; Ellappan, T.R.; Ramalingam, P.V.; Vasudevan, A.T.; Iyer, M.A.K.; Lee, S.M.; Bhoje, S.B.

    2000-01-01

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  5. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    Scott, D.

    1979-01-01

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  6. Fuel management codes for fast reactors

    International Nuclear Information System (INIS)

    Sicard, B.; Coulon, P.; Mougniot, J.C.; Gouriou, A.; Pontier, M.; Skok, J.; Carnoy, M.; Martin, J.

    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)

  7. Liquid metal cooled experimental fast reactor simulator

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Braz Filho, Francisco; Borges, Eduardo M.; Rosa, Mauricio A.P.; Rocamora, Francisco; Hirdes, Viviane R.

    1997-01-01

    This paper is a continuation of the work that has been done in the area of fast reactor component dynamic analysis, as part of the REARA project at the IEAv/CTA-Brazil. A couple of preceding papers, presented in other meetings, introduced major concept design components of the REARA reactor. The components are set together in order to represent a full model of the power plant. Full model transient results will be presented, together with several parameters to help us to better establish the REARA experimental plant concept. (author). 8 refs., 6 figs., 3 tabs

  8. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Picker, C.; Ainsworth, K.F.

    1996-01-01

    The general position with regard to nuclear power and fast reactors in UK during 1995 is described. The status of fast reactor studies made in UK is outlined and a description and statement regarding the conclusions of the programme of studies associated with the closure of the Prototype Fast Reactor is included. (author)

  9. A review of the UK fast reactor programme

    Energy Technology Data Exchange (ETDEWEB)

    Picker, C [AEA Technolgy plc, Risley, Warrington, Cheshire (United Kingdom); Ainsworth, K F [British Nuclear Fuels plc, Sellafield, Cumbria (United Kingdom)

    1996-07-01

    The general position with regard to nuclear power and fast reactors in UK during 1995 is described. The status of fast reactor studies made in UK is outlined and a description and statement regarding the conclusions of the programme of studies associated with the closure of the Prototype Fast Reactor is included. (author)

  10. Material choices for the commercial fast reactor steam generators

    International Nuclear Information System (INIS)

    Willby, C.; Walters, J.

    1978-01-01

    Experience with fast reactor steam generators has shown them to be critical components in achieving a high availability. This paper presents the designers views on the use of ferritic materials for steam generators and describes the proposed design of the steam generators for the Commercial Fast Reactor (CFR), prototype of which are to be inserted in the Prototype Fast Reactor at Dounreay. (author)

  11. The fast reactor and electricity supply, a utility view

    International Nuclear Information System (INIS)

    Wright, J.K.; Hall, R.S.; Kemmish, W.B.; Thorne, R.T.

    1982-01-01

    The significance of the fast reactor is discussed from the viewpoint of the Central Electricity Generating Board. The need for the fast reactor and a possible timescale for its introduction are examined. It is emphasised that demonstration of the commercial and environmental acceptability of the fuel cycle will be needed before any commitment can be made to fast reactors. (U.K.)

  12. What is the future for fast reactor technology?

    International Nuclear Information System (INIS)

    Kraev, Kamen

    2017-01-01

    NucNet spoke to Vladimir Kriventsev, team leader for fast reactor technology development at the International Atomic Energy Agency (IAEA), about the possibilities and challenges of technology development in the fast reactor sector. Today, the field of fast reactors is vibrant and full of fascinating developments, some which will have an impact in the nearer term and others in the longer term.

  13. What is the future for fast reactor technology?

    Energy Technology Data Exchange (ETDEWEB)

    Kraev, Kamen [NucNet, Brussels (Belgium). The Independent Global Nuclear News Agency

    2017-08-15

    NucNet spoke to Vladimir Kriventsev, team leader for fast reactor technology development at the International Atomic Energy Agency (IAEA), about the possibilities and challenges of technology development in the fast reactor sector. Today, the field of fast reactors is vibrant and full of fascinating developments, some which will have an impact in the nearer term and others in the longer term.

  14. Review of fast reactor activities in India (1982-83)

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1983-01-01

    A review of fast reactor activities in India in 1982-1983 is given. One stage of construction of Fast Breeder Test Reactor (FBTR) is briefly described. The emphasis is on design studies for the 500 MWe Prototype Fast Breeder Reactor (PFBR). The main features of this design are introduced

  15. Fast reactor versions: elements of choice

    International Nuclear Information System (INIS)

    Tassart, J.; Zerbib, J.C.

    1984-01-01

    This paper has the objective of explaining in detail the economical, political, social and technical elements on which the CFDT (French Trade Union) bases its opposition to the commercial development of the version of fast reactors. An examination of the different choices which were investigated does not point to any legitimate grounds for this choice. What has to be done is to present the facts which enable the greatest possible number of workers or civilians to take up a position on the choices concerning them. A technical comparison of the fast neutron reactor with those operating at present is put forward (France and United Kingdom). It covers the different radioactive waste products and the results of the individual and collective monitoring of the workmen [fr

  16. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

    Shoai Tehrani, Bianka; Da Costa, Pascal

    2013-01-01

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Gatley, J.A.

    1979-01-01

    Breeder fuel sub-assemblies with electromagnetic brakes and fluidic valves for liquid metal cooled fast breeder reactors are described. The electromagnetic brakes are of relatively small proportions and the valves are of the controlled vortex type. The outlet coolant temperature of at least some of the breeder sub-assemblies are maintained by these means substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (UK)

  19. Non-electric Applications of Fast Reactors

    International Nuclear Information System (INIS)

    Safa, Henri; Borgard, Jean-Marc

    2013-01-01

    Conclusions: → Most of industrial applications (80%) require low temperature heat below 540°C; → Fast Reactors are technically suitable to provide industrial steam at temperatures not accessible by standard LWRs; → As an illustrative example, the application at an oil refinery site has been studied showing the economic benefits; → Nuclear Cogeneration enhances the overall energy efficiency of the power plant; • Nuclear Cogeneration allows massive cut in CO 2 emissions

  20. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Thatcher, G.; Mitchell, A.J.

    1981-01-01

    Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

  1. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

    A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

  2. Nodal method for fast reactor analysis

    International Nuclear Information System (INIS)

    Shober, R.A.

    1979-01-01

    In this paper, a nodal method applicable to fast reactor diffusion theory analysis has been developed. This method has been shown to be accurate and efficient in comparison to highly optimized finite difference techniques. The use of an analytic solution to the diffusion equation as a means of determining accurate coupling relationships between nodes has been shown to be highly accurate and efficient in specific two-group applications, as well as in the current multigroup method

  3. Some aspects of fast reactor economics

    International Nuclear Information System (INIS)

    Kazachkovskij, O.D.

    1996-01-01

    Expedient approach to evaluation of economic efficiency of fast reactors is discussed. It is concluded that determination of electric power generation cost should be based on the fact, that plutonium cost is dictated only by expenses for its extraction from the spent fuel. The cost of the first critical load is not included into capital investments, and investment charges should be sufficiently lower, than standard ones. 5 refs

  4. Technical feasibility study of 60 MWe fast reactor concept: RAPID

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Ueda, Nobuyuki; Uotani, Masaki

    1993-01-01

    A study has been performed on the passive safety features and technical feasibility of an inherently safe 60 MWe fast reactor concept RAPID to meet various power requirements in Japan. The system dynamic analyses on the UTOP and ULOF transients revealed that the enhanced reactivity feedback derived from an annular core configuration and the integrated fuel assembly provides a high margin of self-protection. Structural integrity of the integrated fuel assembly has also been confirmed. The following innovative key technologies have been demonstrated; Lithium Injection Modules (LIM) for ultimate shutdown, Lithium Expansion Modulus (LEM) for inherent reactivity feedback and Void Leading Channel (VLC) for the sodium void worth reduction. (author)

  5. 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”.

  6. Thermal baffle for fast-breeder reactor

    International Nuclear Information System (INIS)

    Rylatt, J.A.

    1977-01-01

    A liquid-metal-cooled fast-breeder reactor includes a bridge structure for separating hot outlet coolant from relatively cool inlet coolant consisting of an annular stainless steel baffle plate extending between the core barrel surrounding the core and the thermal liner associated with the reactor vessel and resting on ledges thereon, there being inner and outer circumferential webs on the lower surface of the baffle plate and radial webs extending between the circumferential webs, a stainless steel insulating plate completely covering the upper surface of the baffle plate and flex seals between the baffle plate and the ledges on which the baffle plate rests to prevent coolant from washing through the gaps therebetween. The baffle plate is keyed to the core barrel for movement therewith and floating with respect to the thermal liner and reactor vessel. 3 claims, 2 figures

  7. Fast reactor operation in the United States

    International Nuclear Information System (INIS)

    Smith, R.R.; Cissel, D.W.

    1978-01-01

    Of the many American facilities dedicated to fast reactor technology, six qualify as liquid-metal-cooled fast reactors. All of these satisfy the following criteria: an unmoderated neutron spectrum, highly enriched fuel material, substantial heat production, and the use of a liquid metal coolant. These include the following: EBR-I Clementine, LAMPRE, EBR-II, EFFBR, and SEFOR. Collectively, these facilities encompassed all of the more important features of liquid-metal-cooled fast reactor technology. Coolant types ranged from mercury in Clementine, to NaK in EBR-I, and sodium in the others. Fuels included enriched-uranium metallic alloys in EBR-I, EBR-II, and EFFBR; metallic plutonium in Clementine; molten plutonium alloy in LAMPRE; and a mixed UO 2 -PuO 2 ceramic in SEFOR. Heat removal techniques ranged from air-blast cooling in LAMPRE and SEFOR; steam-electrical generation in EBR-I, EBR-II, and EFFBR; to a mercury-to-water heat dump in Clementine. Operational experience with such diverse systems has contributed heavily to the U.S. Each of the six systems is described from the viewpoints of purpose, history, design, and operation. Attempts are made to limit descriptive material to the most important features and to refer the reader to a few select references if additional information is needed

  8. Fast-neutron nuclear reactor vessel

    International Nuclear Information System (INIS)

    Presciuttini, L.

    1984-01-01

    The reactor vessel comprises a cylindrical shell, of which axis is vertical, coupled at its lower part with a dished bottom. The reactor core rests on a support plate bearing on the bottom of the vessel. The above dished bottom comprises a spherical sector having the same radius and the same axis as the cylindrical shell and joining the lower part of the shell, and a spherical head of which radius is a little more important than the spherical sector one. A cylindrical support for the reactor core is attached to the vessel at the joint between the two dished sections. The invention applies more particularly to integrated type reactors cooled by liquid sodium [fr

  9. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

    Wild, E.; Mack, K.J.; Gegenheimer, M.

    1984-11-01

    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.) [de

  10. 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.

  11. Today's attitudes and future prospects of fast reactors in Italy

    International Nuclear Information System (INIS)

    Barabaschi, S.; Cicognani, G.; Pierantoni, F.

    1982-01-01

    The Italian fast reactor programme is reviewed. The 15 year collaboration with France has resulted in the construction of the PEC reactor, development of the Superphenix-1 and a common R and D programme for future large fast reactors. The CNEN 4th five year (1980-84) plan is outlined. The budget breakdown for different areas shows the importance attached to the fast reactor. (U.K.)

  12. Status of Fast Reactor Research and Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-04-01

    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.

  13. Status of Fast Reactor Research and Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-01

    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.

  14. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    2012-01-01

    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.

  15. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    2013-01-01

    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.

  16. Materials development for fast reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, T.; Mathew, M.D.; Laha, K.; Sandhya, R., E-mail: san@igcar.gov.in

    2013-12-15

    Highlights: • A modified version of alloy D9 designated as IFAC-1 has been developed. • Oxide dispersion strengthened Grade 91 steel with good creep strength developed. • 0.14 wt% nitrogen in 316LN stainless steel leads to improved mechanical properties. • Type IV cracking resistant Grade 91 steel with boron addition developed. • Mechanical properties of SFR materials evaluated in sodium environment. -- Abstract: Materials play a crucial role in the economic competitiveness of electricity produced from fast reactors. It is necessary to increase the fuel burn-up and design life in order to realize this objective. The burnup is largely limited by the void swelling and creep resistance of the fuel cladding and wrapping materials. India's 500 MWe Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction. The major structural materials chosen for PFBR with MOX fuel are D9 austenitic stainless steel as fuel clad and wrapper material, 316LN austenitic stainless steel for reactor components and piping and modified 9Cr-1Mo steel for steam generator. In order to improve the burnup, titanium, phosphorous and silicon contents in alloy D9 have been optimized for decreased void swelling and increased creep strength and this has led to the development of a modified version of alloy D9 as IFAC-1. Ferritic steels are inherently resistant to void swelling. The disadvantage is their poor creep strength. Creep resistance of 9Cr-ferritic steel has been improved with the dispersion of nano-size yttria to develop oxide dispersion strengthened (ODS) steel clad tube with long-term creep strength, comparable to alloy D9 so as to achieve higher fuel burnup. Improved versions of 316LN stainless steel with nitrogen content of about 0.14 wt% having higher creep strength to increase the life of fast reactors and modified 9Cr-1Mo steel with reduced nitrogen content and controlled addition of boron to improve type IV cracking resistance for steam generator

  17. Primary system thermal hydraulics of future Indian fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

    2015-12-01

    Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

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

    International Nuclear Information System (INIS)

    Bramman, J.I.; Wheeler, R.C.

    1987-03-01

    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.)

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1979-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R D

    1979-07-01

    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.

  1. Status of the DEBENE fast breeder reactor development, March 1979

    International Nuclear Information System (INIS)

    Daeunert, U.; Kessler, G.

    1979-01-01

    Status report of the Fast-breeder reactor development in Germany covers the following: description of the political situation in Federal republic of germany during 1978; international cooperation in the field of fast reactor technology development; operation description of the KNK-II fast core experimental power plant; status of construction of the SNR-300; results of the research and development programs concerned with fuel element, cladding, absorber rods and core structural materials development; sodium effects; neutron irradiation effects on SS properties; reactor physics related to experiments in fast critical assemblies; fast reactor safety issues; core disruption accidents; sodium boiling experiments, measuring methods developed; component tests

  2. Status of the DEBENE fast breeder reactor development, March 1979

    Energy Technology Data Exchange (ETDEWEB)

    Daeunert, U; Kessler, G

    1979-07-01

    Status report of the Fast-breeder reactor development in Germany covers the following: description of the political situation in Federal republic of germany during 1978; international cooperation in the field of fast reactor technology development; operation description of the KNK-II fast core experimental power plant; status of construction of the SNR-300; results of the research and development programs concerned with fuel element, cladding, absorber rods and core structural materials development; sodium effects; neutron irradiation effects on SS properties; reactor physics related to experiments in fast critical assemblies; fast reactor safety issues; core disruption accidents; sodium boiling experiments, measuring methods developed; component tests.

  3. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

    1985-01-01

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author) [pt

  4. Fast Reactor Programme. Third Quarter 1969. Progress Report

    International Nuclear Information System (INIS)

    Hoekstra, E.K.

    1970-02-01

    The RCN research programme on fast spectrum nuclear reactors comprises reactor physics, fuel performance, radiation damage in canning materials, corrosion behaviour in canning materials, aerosol research and heat transfer and hydraulics. An overview is given of the fast reactor experiments at the STEK critical facility in Petten, the Netherlands, in the third quarter of 1969

  5. Superalloy applications in the fast breeder reactor

    International Nuclear Information System (INIS)

    Powell, R.W.

    1976-01-01

    The economics of the LMFBR are dependent on the breeding of new fuel in the reactor core and this can be improved by the use of advanced alloys as core structural components. The environment of the core makes superalloys a natural choice for these components, but phenomena related directly to neutron irradiation necessitate extensive testing. Consequently, commercially-available superalloys, together with a number of developmental alloys are being tested in existing LMFBR's and by simulation techniques to determine the best alloy for use in the LMFBR core. It presently appears that such materials will indeed be capable of the performance required, and will greatly facilitate the commercial realization of the fast breeder reactor

  6. The economics of fast breeder reactors

    International Nuclear Information System (INIS)

    Rapin, M.

    1990-01-01

    The overall status of the fast breeder reactor (FBR) system is periodically reviewed in France. In 1983, a report was prepared on the status and prospects of the FBR system at the request of the then Minister of Industry. Five years later, Electricite de France (EdF) and the French Atomic Energy Commission (CEA) jointly updated this report. The FBR reactor system economic considerations mentioned here are taken from the work performed in 1987-88 for this updating. The position in 1983 is reviewed to highlight concrete developments. Developments that have occurred since then are presented, along with the prospects that today enable us to define better the technical and economic potential of the FBR system. In conclusion, the effects of these findings on desirable directions are discussed, in particular with regard to European FBR cooperation. (author)

  7. Design and selection of materials for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Chetal, S.C.

    2011-01-01

    Sodium cooled fast reactors are currently in operation, under construction or under design by a number of countries. The design of sodium cooled fast reactor is covered by French RCC - MR code and ASME code NH. The codes cover rules as regards to materials, design and construction. These codes do not cover the effect of irradiation and environment. Elevated temperature design criteria in nuclear codes are much stringent in comparison to non nuclear codes. Sodium corrosion is not an issue in selection of materials provided oxygen impurity in sodium is controlled for which excellent reactor operating experience is available. Austenitic stainless steels have remained the choice for the permanent structures of primary sodium system. Stabilized austenitic stainless steel are rejected because of poor operating experience and non inclusion in the design codes. Route for improved creep behaviour lies in compositional modifications in 316 class steel. However, the weldability needs to be ensured. For cold leg component is non creep regime, SS 304 class steel is favoured from overall economics. Enhanced fuel burn up can be realized by the use of 9-12%Cr 1%Mo class steel for the wrapper of MOX fuel design, and cladding and wrapper for metal fuel reactors. Minor compositional modifications of 20% cold worked 15Cr-15Ni class austenitic stainless steel will be a strong candidate for the cladding of MOX fuel design in the short term. Long term objective for the cladding will be to develop oxide dispersion strengthened steel. 9%Cr 1%Mo class steel (Gr 91) is an ideal choice for integrated once through sodium heated steam generators. One needs to incorporate operating experience from reactors and thermal power stations, industrial capability and R and D feedback in preparing the technical specifications for procurement of wrought products and welding consumables to ensure reliable operation of the components and systems over the design life. The paper highlights the design approach

  8. Gas cooled fast reactor research in Europe

    International Nuclear Information System (INIS)

    Stainsby, Richard; Peers, Karen; Mitchell, Colin; Poette, Christian; Mikityuk, Konstantin; Somers, Joe

    2011-01-01

    Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce. GFR research within Europe is performed directly by those states who have signed the 'System Arrangement' document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with

  9. Safe Management Of Fast Reactors: Towards Sustainability

    International Nuclear Information System (INIS)

    Dreimanis, Andrejs

    2015-01-01

    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)

  10. GENIUS & the Swedish Fast Reactor programme

    International Nuclear Information System (INIS)

    Wallenius, Janne

    2012-01-01

    Concluding remarks: Sweden’s growing fast reactor programme focuses on LFR technology, but we also participate in ASTRID. • An innovative facility for UN fabrication, an LBE thermal hydraulics loop and a lead corrosion facility are operational. • A plutonium fuel fabrication lab is is under installation (this week!) • The government is assessing the construction of ELECTRA-FCC, a centre for Gen IV-system R&D, at a tentative cost of ~ 140±20 M€. • Location: Oskarshamn (adjacent to intermediate repository) • Date of criticality: 2023 (best case) • Swedish participation in IAEA TWG-FR should intensify

  11. Heterogeneous cores for fast breeder reactor

    International Nuclear Information System (INIS)

    Schroeder, R.; Spenke, H.

    1980-01-01

    Firstly, the motivation for heterogeneous cores is discussed. This is followed by an outline of two reactor designs, both of which are variants of the combined ring and island core. These designs are presented by means of figures and detailed tables. Subsequently, a description of two international projects at fast critical zero energy facilities is given. Both of them support the nuclear design of heterogeneous cores. In addition to a survey of these projects, a typical experiment is discussed: the measurement of rate distributions. (orig.) [de

  12. Core of a fast neutron nuclear reactor

    International Nuclear Information System (INIS)

    Giacometti, Christian; Mougniot, J.-C.; Ravier, Jean.

    1974-01-01

    The fast neutron nuclear reactor described includes an internal area in fissile material completely enclosed in an area of fertile material forming the outside blanket. The internal fissile area is provided with housings exclusively filled with fertile material forming one or more inside blankets. In this core the internal blankets are shaped like rings vertically separating superimposed rings of fissile material. The blanket of material nearest to the periphery is circumscribed externally by a contour having an indented shape on its straight section so as to increase the contact area between this blanket and the external blanket [fr

  13. Nuclear fuel assembly for fast neutron reactors

    International Nuclear Information System (INIS)

    Ilyunin, V.G.; Murogov, V.M.; Troyanov, M.F.; Rinejskij, A.A.; Ustinov, G.G.; Shmelev, A.N.

    1982-01-01

    The fuel assembly of a fast reactor consists of fuel elements comprising sections with fissionable and breeding material and tubes with hollows designed for entrapping gaseous fission products. Tubes joining up to the said sections are divided in a middle and a peripheral group such that at least one of the tube groups is placed in the space behind the coolant inlet ports. The configuration above allows reducing internal overpressure in the fuel assembly, thus reducing the volume of necessary structural elements in the core. (J.B.)

  14. Accident analysis for US fast burst reactors

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  15. Reprocessing of fast neutron reactor fuel

    International Nuclear Information System (INIS)

    Bourgeois, M.

    1981-05-01

    A PUREX process specially adapted to fast neutron reactor fuels is employed. The results obtained indicate that the aqueous process can be applied to this type of fuel: almost 10 years operation at the AT 1 plant which processes fuel from RAPSODIE; the good results obtained at the MARCOULE pilot plant on large batches of reference fuels. The CEA is continuing its work to transfer this technology onto an industrial scale. Industrial prototypes and the launching of the TOR (traitement d'oxydes rapides) project will facilitate this transfer. In 1984, it is expected that fast fuels will be able to be processed on a significant scale and that supplementary R and D facilities will be available [fr

  16. Trends and Developments for Fast Neutron Reactors and Related Fuel Cycles

    International Nuclear Information System (INIS)

    Carré, Frank

    2013-01-01

    • FR13 – A unique and dedicated framework to share updates on national programs of Fast Reactor developments, projects of new builds and plans for the future: - Near term projects of sodium and lead-alloy Fast Reactors; - Gen-IV visions of sodium-cooled and alternative types of Fast Neutron Reactors (GFR, LFR…). • FR13 – A special emphasis put on Fast Reactor Safety, Sustainability of nuclear fuel cycle and Young Generation perspective. • FR13 – A catalyst for further collaborations and alliances: - To share visions of goals and advisable options for future Fast Reactors and Nuclear Fuel Cycle; - To share cost of R&D and large demonstrations (safety, security, recycling); - To progress towards harmonized international standards; - To integrate national projects into a consistent international roadmap

  17. Status of liquid metal cooled fast reactor technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-04-01

    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 Refs, figs, tabs

  18. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    1999-04-01

    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

  19. Review of fast reactor activities in Italy, April 1978

    Energy Technology Data Exchange (ETDEWEB)

    Pierantoni, F [CNEN Fast Reactor Programme, Bologna (Italy)

    1978-07-01

    In summary, the Italian fast reactor programme was developing in the following directions: PEC reactor, SUPEPHENIX reactor and long-term research and development work. Research was related to sodium technology, steam generators development, pumps, tests on mechanics and thermal insulation, core fluid dynamics, noise analysis, studies of oxide and carbide fuels, reactor safety, CABRI and SCARABEE experiments.

  20. Review of fast reactor activities in Italy, April 1978

    International Nuclear Information System (INIS)

    Pierantoni, F.

    1978-01-01

    In summary, the Italian fast reactor programme was developing in the following directions: PEC reactor, SUPEPHENIX reactor and long-term research and development work. Research was related to sodium technology, steam generators development, pumps, tests on mechanics and thermal insulation, core fluid dynamics, noise analysis, studies of oxide and carbide fuels, reactor safety, CABRI and SCARABEE experiments

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

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Enuma, Yasuhiro; Kubota, Kenichi; Yoshida, Masashi; Uno, Osamu; Ishikawa, Hiroyasu; Kobayashi, Jun; Umetsu, Youichiro; Ichimiya, Masakazu

    1999-10-01

    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)

  2. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

    2007-02-07

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

  3. Research activities on fast reactors in Switzerland

    International Nuclear Information System (INIS)

    Brogli, R.; Dones, R.; Hudina, M.; Pelloni, S.

    1996-01-01

    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

  4. A small modular fast reactor as starting point for industrial deployment of fast reactors

    International Nuclear Information System (INIS)

    Chang, Yoon I.; Lo Pinto, Pierre; Konomura, Mamoru

    2006-01-01

    The current commercial reactors based on light water technology provide 17% of the electricity worldwide owing to their reliability, safety and competitive economics. In the near term, next generation reactors are expected to be evolutionary type, taking benefits of extensive LWR experience feedbacks and further improved economics and safety provisions. For the long term, however, sustainable energy production will be required due to continuous increase of the human activities, environmental concerns such as greenhouse effect and the need of alternatives to fossil fuels as long term energy resources. Therefore, future generation commercial reactors should meet some criteria of sustainability that the current generation cannot fully satisfy. In addition to the current objectives of economics and safety, waste management, resource extension and public acceptance become other major objectives among the sustainability criteria. From this perspective, two questions can be raised: what reactor type can meet the sustainability criteria, and how to proceed to an effective deployment in harmony with the high reliability and availability of the current nuclear reactor fleet. There seems to be an international consensus that the fast spectrum reactor, notably the sodium-cooled system is most promising to meet all of the long term sustainability criteria. As for the latter, we propose a small modular fast reactor project could become a base to prepare the industrial infrastructure. The paper has the following contents: - Introduction; - SMFR project; - Core design; - Supercritical CO 2 Brayton cycle; - Near-term reference plant; - Advanced designs; - Conclusions. To summarize, the sodium-cooled fast reactor is currently recognized as the technology of choice for the long term nuclear energy expansion, but some research and development are required to optimize and validate advanced design solutions. A small modular fast reactor can satisfy some existing near-term market niche

  5. 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.

  6. MANTA. An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

    Energy Technology Data Exchange (ETDEWEB)

    Youinou, Gilles Jean-Michel [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    neutron irradiation allows to infer energy-integrated neutron cross sections, i.e. ∫₀σ(E)φ(E)dE, where φ(E) is the neutron flux “seen” by the sample. This approach, which is usually defined and led by reactor physicists, is referred to as integral and is the object of this report. These two sources of information, i.e. differential and integral, are complementary and are used by the nuclear physicists in charge of producing the evaluated nuclear data files used by the nuclear community (ENDF, JEFF…). The generation of accurate nuclear data files requires an iterative process involving reactor physicists and nuclear data evaluators. This experimental program has been funded by the ATR National Scientific User Facility (ATR-NSUF) and by the DOE Office of Science in the framework of the Recovery Act. It has been given the name MANTRA for Measurement of Actinides Neutron TRAnsmutation.

  7. Simplified simulation of an experimental fast reactor plant

    International Nuclear Information System (INIS)

    Fujii, Masaaki; Fujita, Minoru.

    1978-01-01

    Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

  8. A resting bottom sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Costes, D.

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-06-01

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

  10. 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)

    1988-11-01

    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

  11. Seminar on Heat-transfer fluids for fast neutron reactors

    International Nuclear Information System (INIS)

    Brechet, Yves; Dautray, Robert; Friedel, Jacques; Brezin, Edouard; Martin, Georges; Pineau, Andre; Carre, Francois; Gauche, Francois; Rodriguez, Guillaume; Latge, Christian; Cabet, Celine; Garnier, Jean-Claude; Bamberger, Yves; Sauvage, Jean-Francois; Buisine, Denis; Agostini, Pietro; Ulyanov, Vladimir; Auger, Thierry; Heuer, Daniel; Ghetta, Veronique; Bubelis, Evaldas; Charlaix, Elisabeth; Barrat, Jean-Louis; Boquet, Lyderic; Glickman, Evgueny; Escaravage, Claude

    2014-03-01

    This book reports the content of a two-day meeting held by the Academy of Sciences on the use of heat-transfer fluids in fast neutron reactors. After a first part which proposes an overview of scientific and technical problems related to these heat-transfer fluids (heat transfer process, nuclear properties, chemistry, materials, risks), a contribution proposes a return on experience on the use of heat-transfer fluids in the different design options of reactors of fourth generation: from mercury to NaK in the first fast neutron reactor projects, specific assets and constraints of sodium used as heat-transfer fluid, concepts of fast neutron reactors cooled by something else than sodium, perspectives for projects and research in fast neutron reactors. The next contribution discusses the specifications of future fast-neutron reactors: expectations for fourth-generation reactors, expectations in terms of performance and of safety, specific challenges. The last contribution addresses actions to be undertaken in the field of research and development: actions regarding all reactor types or specific types as sodium-cooled reactors, lead cooled reactors, molten salt reactors, and gas-cooled fast reactors

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  13. Development of the IAEA’s Knowledge Preservation Portals for Fast Reactors and Gas-Cooled Reactors Knowledge Preservation

    International Nuclear Information System (INIS)

    Batra, C.; Menahem, D. Beraha; Kriventsev, V.; Monti, S.; Reitsma, F.; Grosbois, J. de; Khoroshev, M.; Gladyshev, M.

    2016-01-01

    Full text: The IAEA has been carrying out a dedicated initiative on fast reactor knowledge preservation since 2003. The main objectives of the Fast Reactor Knowledge Portal (FRKP) initiative are to, a) halt the on-going loss of information related to fast reactors (FR), and b) collect, retrieve, preserve and make accessible existing data and information on FR. This portal will help in knowledge sharing, development, search and discovery, collaboration and communication of fast reactor related information. On similar lines a Gas Cooled Fast Reactor Knowledge Preservation portal project also started in 2013. Knowledge portals are capable to control and manage both publicly available as well as controlled information. The portals will not only incorporate existing set of knowledge and information, but will also provide a systemic platform for further preservation of new developments. It will include fast reactor and gas cooled reactor document repositories, project workspaces for the IAEA’s Coordinated Research Projects (CRPs), Technical Meetings (TMs), forums for discussion, etc. The portal will also integrate a taxonomy based search tool, which will help using new semantic search capabilities for improved conceptual retrieve of documents. The taxonomy complies with international web standards as defined by the W3C (World Wide Web Consortium). (author

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

    Energy Technology Data Exchange (ETDEWEB)

    Pierantoni, F

    1979-07-01

    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

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

    International Nuclear Information System (INIS)

    Pierantoni, F.

    1979-01-01

    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

  16. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

    International Nuclear Information System (INIS)

    Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

    2010-01-01

    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. (authors)

  17. Fast Reactor Knowledge Preservation Activities at the OECD Nuclear Energy Agency

    International Nuclear Information System (INIS)

    Hill, Ian

    2013-01-01

    • NEA is organizing significant activities are ongoing to preserve fast reactor information: – Integral experiments supporting fast reactors; – Often “use it, or lose it”; – Difficult to preserve everything, critical information is identified; – Organisation is a large but important task; – Electronic databases needed to manage the data. • Authoritative Handbooks and state of the art reports: – These tend to be the documents that last. • OECD/NEA will continue to support member countries in fast reactor knowledge preservation: – forum for exchange of information; – collaborative activities

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

    International Nuclear Information System (INIS)

    Moiseyev, A.V.

    2008-01-01

    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 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)

  19. 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)

  20. Fast reactor knowledge preservation system: Taxonomy and basic requirements

    International Nuclear Information System (INIS)

    2008-01-01

    The IAEA has taken the initiative to coordinate efforts of Member States in the preservation of knowledge in the area of fast reactors. In the framework of this initiative, the IAEA intends to create an international database compiling information from different Member States on fast reactors through a web portal. Other activities related to this initiative are being designed to accumulate and exchange information on the fast reactor area, to facilitate access to this information by users in different countries and to assist Member States in preserving the experience gained in their countries. The purpose of this publication is to develop a taxonomy of the Fast Reactor Knowledge Preservation System (FRKPS) that will facilitate the preservation of the world's fast reactor knowledge base, to identify basic requirements of this taxonomy on the basis of the experience gained in the fast reactor area, as well as results of previous IAEA activities on fast reactor knowledge preservation. The need for such taxonomy arises from the fact that during the past 15 years there has been stagnation in the development of fast reactors in the industrialized countries that were involved, earlier, in intensive development of this area. All studies on fast reactors have been stopped in countries such as Germany, Italy, the United Kingdom and the United States of America and the only work being carried out is related to the decommissioning of fast reactors. Many specialists who were involved in the studies and development work in this area in these countries have already retired or are close to retirement. In countries such as France, Japan and the Russian Federation that are still actively pursuing the evolution of fast reactor technology, the situation is aggravated by the lack of young scientists and engineers moving into this branch of nuclear power

  1. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Duncombe, E.; Thatcher, G.

    1979-01-01

    The invention described relates to a liquid metal cooled fast breeder nuclear reactor in which the fuel assembly has an inner zone comprised mainly of fissile material and a surrounding outer zone comprised mainly of breeder material. According to the invention the sub-assemblies in the outer zone include electro-magnetic braking devices (magnets, pole pieces and armature) for regulating the flow of coolant through the sub-assemblies. The magnetic fields of the electro-magnetic breaking devices are temperature sensitive so that as the power output of the breeder sub-assemblies increases the electro-magnetic resistance to coolant flow is reduced thereby maintaining the temperature of the coolant outlets from the sub-assemblies substantially constant. (UK)

  2. Analysis of fast reactor steam generator performance

    International Nuclear Information System (INIS)

    Hulme, G.; Curzon, A.F.

    1992-01-01

    A computer model for the prediction of flow and temperature fields within a fast reactor steam generator unit is described. The model combines a commercially available computational fluid dynamics (CFD) solver (PHOENICS) with a steam-tube calculation and provides solutions for the fully coupled flow and temperature fields on both the shell side and the tube side. The model includes the inlet and outlet headers and the bottom end stagnant zone. It also accounts for the effects of support grids and edge-gaps. Two and three dimensional and transient calculations have been performed for both straight tube and J-tube units. Examples of the application of the model are presented. (7 figures) (Author)

  3. Improvement of covariance data for fast reactors

    International Nuclear Information System (INIS)

    Shibata, Keiichi; Hasegawa, Akira

    2000-02-01

    We estimated covariances of the JENDL-3.2 data on the nuclides and reactions needed to analyze fast-reactor cores for the past three years, and produced covariance files. The present work was undertaken to re-examine the covariance files and to make some improvements. The covariances improved are the ones for the inelastic scattering cross section of 16 O, the total cross section of 23 Na, the fission cross section of 235 U, the capture cross section of 238 U, and the resolved resonance parameters for 238 U. Moreover, the covariances of 233 U data were newly estimated by the present work. The covariances obtained were compiled in the ENDF-6 format. (author)

  4. Trial visualization of fast reactor design knowledge

    International Nuclear Information System (INIS)

    Yoshikawa, Shinji; Minami, Masaki; Takahashi, Tadao

    2011-01-01

    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)

  5. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1976-01-01

    Reference is made to liquid metal cooled fast breeder reactors of the 'pool' kind. In this type of reactor the irradiated fuel is lowered into a transfer rotor for removal to storage facilities, this rotor normally having provision for the temporary storage of 20 irradiated fuel assemblies, each within a stainless steel bucket. For insertion or withdrawal of a fuel assembly the rotor is rotated to bring the fuel assembly to a loading or discharging station. The irradiated fuel assembly is withdrawn from the rotor within its bucket and the total weight is approximately 1000 kg, which is lifted about 27 m. In the event of malfunction the combination falls back into the rotor with considerable force. In order to prevent damage to the rotor fracture pins are provided, and to prevent damage to the reactor vessel and other parts of the reactor structure deformable energy absorbing devices are provided. After a malfunction the fractured pins and the energy absorbing devices must be replaced by remote control means operated from outside the reactor vault - a complex operation. The object of the arrangement described is to provide improved energy absorbing means for fuel assemblies falling into a fuel transfer rotor. The fuel assemblies are supported in the rotor by elastic means during transfer to storage and a hydraulic dash pot is provided in at least one position below the rotor for absorbing the energy of a falling fuel assembly. It is preferable to provide dash pots immediately below a receiving station for irradiated fuel assemblies and immediately below a discharge station. Each bucket is carried in a container that is elastically supported in the transfer rotor on a helical coil compression spring, so that, in the event of a malfunction the container and bucket are returned to their normal operating position after the force of the falling load has been absorbed by the dash pot. The transfer rotor may also be provided with recoil springs to absorb the recoil energy

  6. Fabrication of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    Mishra, Sudhir; Kumar, Arun; Kutty, T.R.G.; Kamath, H.S.

    2011-01-01

    Mixed oxide (MOX) (U,Pu)O 2 , and metallic (U,Pu ,Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity , low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion. The higher coefficient of linear expansion is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burnup, fuel cladding interaction and lower margin between operating and melting temperature. The optimal solution may lie in cermet fuel (U, PuO 2 ), where PuO 2 is dispersed in U metal matrix and combines the favorable features of both the fuel types. The advantages of this fuel include high thermal conductivity, larger margin between melting and operating temperature, ability to retain fission product etc. The matrix being of high density metal the advantage of high breeding ratio is also maintained. In this report some results of fabrication of cermet pellet comprising of UO 2 /PuO 2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

  7. 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.

  8. A contribution to the method of fast reactor thermal output calculation

    International Nuclear Information System (INIS)

    Harant, M.

    1978-01-01

    The method of stating the heat sources is discussed as being one of the factors influencing the accuracy of the thermal output calculation of fast reactors. The distribution of heat sources in the core and in other inner parts of the fast reactor is described using the least square fit method. Relations are derived of outputs of both individual components of fuel elements and of whole inner parts of the reactor. A comparison is made of various methods used for obtaining source integrals. The optimum integration method was found. (author)

  9. Plant experience of experimental fast reactor 'Joyo'

    International Nuclear Information System (INIS)

    1982-01-01

    The experimental fast reactor ''JOYO'' installed in Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan completed its operation using the first core (called MK-I core) in December, 1981, and the works to transfer to MK-2 core have been performed since January, 1982. In this report, the experiences obtained through the construction, test and operation of ''JOYO'' over 12 years from the start of erection in 1970 to the termination of operation in 1981 are described. The contents of the report are divided into design, construction, the outline of facilities, testing, operating and maintenance experiences, and the topics on MK-I operation. As for the construction, the design changes performed before the start of manufacture or construction and the improvement and trouble restoring works implemented at the start of overall functional tests are reported. As for testing, overall functional tests, criticality test, low power test and power increasing test are described in detail. The number of test items of overall functional testing reached 266. The rated output operation of the reactor at 75 MW was performed six times in 1980 and 1981 until the termination of operation. No fuel failure was detected in MK-I operation, and the stable operation performance of the FBR was proved through MK-I operation. The topics on the MK-I operation includes natural circulation test, the measurement of total leakage rate for the containment vessel, and wear-marks which are the trace of wear due to the contact of fuel pins with the wires wound around the adjacent fuel pins, found in the post irradiation examination of fuel. (Wakatsuki, Y.)

  10. Fast Reactor Knowledge Organization System (FR-KOS)

    International Nuclear Information System (INIS)

    Costantini, Livio

    2013-01-01

    FR-KOS Web-portal general concepts: FR-KOS web-portal is an integrated and unique solution for knowledge preservation, retrieval and management in fast reactor (FR) knowledge domain; The exploitation of the FR-KOS provides an effective and flexible tool to access FR knowledge repositories. The dynamic and interactive process of searching information is guided by structures that are based on the FR Taxonomy; FR-KOS can be extended to other types of reactor and technologies, such as PWR, WWER, BWR, CANDU, etc. In addition can also be utilized in other information management process such as the preservation of nuclear accidents knowledg;. FR-KOS addresses needs in non-structured knowledge preservation for both embarking countries and countries with established nuclear power; FR-KOS repositories contain more that 50.000 samples of technical and scientific publications, including full text documents in different languages (English, German, French and eventually in Russian)

  11. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  12. History of fast reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kittel, J.H. (Argonne National Lab., IL (United States)); Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France)); Bagley, K.Q. (AEA Reactor Services, Risley (United Kingdom)); Crittenden, G.C. (AEA Reactor Services, Dounreay (United Kingdom)); Dievoet, J. van (Belgonucleaire, Brussels (Belgium))

    1993-09-01

    The first fast breeder eactors, 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. (orig.)

  13. Fast reactor physics at CEA: present studies and future prospects

    International Nuclear Information System (INIS)

    Hammer, P.

    1980-09-01

    This paper aims at giving a general survey of the fast reactor core physics and shielding studies wich are in progress at CEA (1979-1983) in order to solve the neutronic problems related to: - core design optimization, - reactor operation and fuel management, - safety, for the development of fast commercial breeders in France after the SUPER-PHENIX 1 construction is achieved

  14. The problems of thermohydraulics of prospective fast reactor concepts

    International Nuclear Information System (INIS)

    Sedov, A.A.

    2000-01-01

    In this report the main requirements to fast reactors in system of future multicomponent Nuclear Power with closed U-Pu fuel cycle are regarded. The peculiarities of different liquid-metal (sodium and lead-alloyed) coolants as well as the thermohydraulics problems of prospective fast reactors (FR) concepts are discussed. (author)

  15. A review of fast reactor progress in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Tomabechi, K [Power Reactor and Nuclear Fuel Development Corporation, Tokyo (Japan)

    1978-07-01

    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.

  16. Simulating the Behaviour of the Fast Reactor Joyo (Draft)

    International Nuclear Information System (INIS)

    Juutilainen, Pauli

    2008-01-01

    Motivated by the development of fast reactors the behaviour of the Japanese experimental fast reactor Joyo is simulated with two Monte Carlo codes: Monte Carlo NParticle (MCNP) and Probabilistic Scattering Game (PSG). The simulations are based on the benchmark study 'Japan's Experimental Fast Reactor Joyo MKI core: Sodium-Cooled Uranium-Plutonium Mixed Oxide Fueled Fast Core Surrounded by UO 2 Blanket'. The study is focused on the criticality of the reactor, control rod worth, sodium void reactivity and isothermal temperature coefficient of the reactor. These features are calculated by applying both homogeneous and heterogeneous reactor core models that are built according to the benchmark instructions. The results of the two models obtained by the two codes are compared with each other and especially with the experimental results presented in the benchmark. (author)

  17. Logistical and economic obstacles to a fast reactor programme

    International Nuclear Information System (INIS)

    Sweet, C.

    1982-01-01

    Fast reactor studies used to place great emphasis on its role as a breeder of plutonium, thereby raising the prospect of a nuclear power system free from the constraint of uranium supplies. Today, not only the timing of fast reactor introduction has slipped (by two or three decades) but the perspective which was central to energy policy has changed dramatically. This article first examines the fast reactor as a system and looks at the interaction of four key variables in its logistics. It then looks at the rise in real costs, especially capital costs. Given the parameters that determine the plutonium balance and the economics of the fast reactor system, the author questions whether there is a sound basis for its introduction, and concludes by suggesting that the most pressing requirement is a study of the opportunity costs of fast reactor R and D expenditures. (author)

  18. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Picker, C.; Ainsworth, K.F.

    1998-01-01

    The general position with regard to nuclear power and fast reactors in the UK during 1996 is described. The main UK Government-funded fast reactor research and development programme was concluded in 1993, to be replaced by a smaller programme which is mainly funded and managed by British Nuclear Fuels plc. The main focus of this programme sustains the UK participation in the European Fast Reactor (EFR) collaboration and the broader international links built-up over the previous decades. The status of fast reactor studies made in the UK in 1996 is outlined and, with respect to the Prototype Fast Reactor at Dounreay, a report of progress with the closure studies, fuel reprocessing and decommissioning activities is provided. (author)

  19. Overview of fast reactor safety research and development in the USA

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Avery, R.; Marchaterre, J.F.

    1986-01-01

    The liquid metal reactor (LMR) safety R and D program in the U.S. is presently focused on support of two modular innovative reactor concepts: PRISM - the General Electric Power Reactor Inherently Safe Module and SAFR - the Rockwell International Sodium Advanced Fast Reactor. These reactor plant concepts accommodate the use of either oxide fuel or the metal fuel which is under development in the Argonne National Laboratory (ANL) Integral Fast Reactor (IFR) program. Both concepts emphasize prevention of accidents through enhancement of inherent and passive safety characteristics. Enhancement of these characteristics is expected to be a major factor in establishing new and improved safety criteria and licensing arrangements with regulatory authorities for advanced reactors. Limited work is also continuing on the Large Scale Prototype Breeder (LSPB), a large pool plant design. Major elements of the current and restructured safety program are discussed. (author)

  20. LTFR-4, Library Generated for Fast Reactor Design Program from JAERI Fast-Set Multigroup Constant

    International Nuclear Information System (INIS)

    Suzuki, Tomoo

    1971-01-01

    Nature of physical problem solved: The program processes JAERI-Fast group constants sets of less than 30-group and prepares a binary library tape for efficient usage by a series of related fast reactor design calculation programmes

  1. Safety Analysis Of Actinide Recycled Fast Power Reactor

    International Nuclear Information System (INIS)

    Taufik, Mohammad

    2001-01-01

    Simulation for safety analysis of actinide recycled fast power reactor has been performed. The objective is to know reactor response about ULOF and ULOF and UTOP simultaneous accident. From parameter result such reactivity feedback, power, temperature, and cooled flow rate can conclusion that reactor have inherent safety system, which can back to new Equilibrium State

  2. Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

    OpenAIRE

    Verma, Vasudha

    2017-01-01

    Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1981-01-01

    steam cycle, a re-evaluation of various boiler design styles have been undertaken. The types being studied cover straight tube, U and J tube, hockey stick and helical coil designs. A final choice of steam generator for CDFR has not yet been made. Key features in design selection are overall integrity to ensure high availability, ease of maintenance and repair following minor tube leaks, resistance to escalation of minor leaks to more major sodium/water reaction incidents. inspectability, relevance of previous experience and capital cost. Operation of the PFR was interrupted by a series of small leaks in the tube-to-tube plate welds in the three evaporator units. The key areas have now been shot-peened to remove tensile stresses in the surface material. the units have been repaired and operation at high power should be restarted during March 1981. The performance of the core and primary circuits has continued to be excellent. A major step forward was taken when the first 16 sub-assemblies were reprocessed in the PFR reprocessing plant. During the year a system economic study has been completed to assist in providing a sound technical basis for a fast reactor inquiry. This study was based on a set of programme assumptions and economic ground rules developed and agreed in 1978 within the UK nuclear industry and updated in 1979. Broadly, it was concluded that the introduction of the fast reactor as soon and as rapidly as feasible also leads, even on conservative uranium price assumptions. to economic benefits, the benefits being greatest when electrical demand and raw fuel prices are at their highest. The benefits are little affected by 25% increases in costs, except for an increase in fast reactor capital cost when uranium stays cheap; in this case the benefit virtually disappears if it is assumed that all other capital costs stay constant. Use of the fast reactor also produces a dramatic fall (almost three-fold) in the need uranium imports and, if market forces are

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R D [Risley Nuclear Power Development Establishment, Risley, Warrington (United Kingdom)

    1981-05-01

    steam cycle, a re-evaluation of various boiler design styles have been undertaken. The types being studied cover straight tube, U and J tube, hockey stick and helical coil designs. A final choice of steam generator for CDFR has not yet been made. Key features in design selection are overall integrity to ensure high availability, ease of maintenance and repair following minor tube leaks, resistance to escalation of minor leaks to more major sodium/water reaction incidents. inspectability, relevance of previous experience and capital cost. Operation of the PFR was interrupted by a series of small leaks in the tube-to-tube plate welds in the three evaporator units. The key areas have now been shot-peened to remove tensile stresses in the surface material. the units have been repaired and operation at high power should be restarted during March 1981. The performance of the core and primary circuits has continued to be excellent. A major step forward was taken when the first 16 sub-assemblies were reprocessed in the PFR reprocessing plant. During the year a system economic study has been completed to assist in providing a sound technical basis for a fast reactor inquiry. This study was based on a set of programme assumptions and economic ground rules developed and agreed in 1978 within the UK nuclear industry and updated in 1979. Broadly, it was concluded that the introduction of the fast reactor as soon and as rapidly as feasible also leads, even on conservative uranium price assumptions. to economic benefits, the benefits being greatest when electrical demand and raw fuel prices are at their highest. The benefits are little affected by 25% increases in costs, except for an increase in fast reactor capital cost when uranium stays cheap; in this case the benefit virtually disappears if it is assumed that all other capital costs stay constant. Use of the fast reactor also produces a dramatic fall (almost three-fold) in the need uranium imports and, if market forces are

  5. Simplified procedures for fast reactor fuel cycle and sensitivity analysis

    International Nuclear Information System (INIS)

    Badruzzaman, A.

    1979-01-01

    The Continuous Slowing Down-Integral Transport Theory has been extended to perform criticality calculations in a Fast Reactor Core-blanket system achieving excellent prediction of the spectrum and the eigenvalue. The integral transport parameters did not need recalculation with source iteration and were found to be relatively constant with exposure. Fuel cycle parameters were accurately predicted when these were not varied, thus reducing a principal potential penalty of the Intergal Transport approach where considerable effort may be required to calculate transport parameters in more complicated geometries. The small variation of the spectrum in the central core region, and its weak dependence on exposure for both this region, the core blanket interface and blanket region led to the extension and development of inexpensive simplified procedures to complement exact methods. These procedures gave accurate predictions of the key fuel cycle parameters such as cost and their sensitivity to variation in spectrum-averaged and multigroup cross sections. They also predicted the implications of design variation on these parameters very well. The accuracy of these procedures and their use in analyzing a wide variety of sensitivities demonstrate the potential utility of survey calculations in Fast Reactor analysis and fuel management

  6. Uranium self-shielding in fast reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Kadiroglu, O.K.; Driscoll, M.J.

    1976-03-01

    The effects of heterogeneity on resonance self-shielding are examined with particular emphasis on the blanket region of the fast breeder reactor and on its dominant reaction--capture in /sup 238/U. The results, however, apply equally well to scattering resonances, to other isotopes (fertile, fissile and structural species) and to other environments, so long as the underlying assumptions of narrow resonance theory apply. The heterogeneous resonance integral is first cast into a modified homogeneous form involving the ratio of coolant-to-fuel fluxes. A generalized correlation (useful in its own right in many other applications) is developed for this ratio, using both integral transport and collision probability theory to infer the form of correlation, and then relying upon Monte Carlo calculations to establish absolute values of the correlation coefficients. It is shown that a simple linear prescription can be developed for the flux ratio as a function of only fuel optical thickness and the fraction of the slowing-down source generated by the coolant. This in turn permitted derivation of a new equivalence theorem relating the heterogeneous self-shielding factor to the homogeneous self-shielding factor at a modified value of the background scattering cross section per absorber nucleus. A simple version of this relation is developed and used to show that heterogeneity has a negligible effect on the calculated blanket breeding ratio in fast reactors.

  7. Implications of Fast Reactor Transuranic Conversion Ratio

    International Nuclear Information System (INIS)

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

    2010-01-01

    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, 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 among the higher actinides, so the neutron emission varies much stronger with CR, about three orders of magnitude.

  8. Design characteristics of zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Popovic, D.; Nikolic, D.; Antic, D.; Zavaljevski, N.

    1987-01-01

    The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)

  9. Advances in neutronics calculation of fast neutron reactors - Demonstration on Super-Phenix reactor

    International Nuclear Information System (INIS)

    Czernecki, Sebastien

    1998-01-01

    The fast reactor european neutronics calculations system, ERANOS, has integrated recent improvements both in nuclear data, with the use of the adjusted nuclear library ERALIB 1 from the JEF2.2 library, and calculation methods, with the use of the new european cell code, ECCO, and the deterministic code, TGV/VARIANT. This code performs full 3-D reactor calculation in the transport theory with variational method. The aim of this work is to create and validate a new calculational scheme for fast spectrum systems offering good compromise between accuracy and running time. The new scheme is based on these improvements plus a special procedure accounting for control rod heterogeneity, which uses a reactivity equivalence homogenization. The new scheme has been validated by means of experiment/calculation comparisons, using the extensive start-up program measurements performed in Super-Phenix reactor. The validation uses also recent measurements performed in the Phenix reactor. The results are very satisfactory and show a significant improvement for almost all core parameters, especially for critical mass, control rod worth and radial subassembly power distribution. A detailed analysis of the discrepancies between the old scheme and the new one for this parameter allows to understand the separate effects of methods and nuclear data on the radial power distribution shape. (author) [fr

  10. Core disruptive accident analysis in prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Velusamy, K.; Kannan, S.E.; Singh, Om Pal; Chetal, S.C.; Bhoje, S.B.

    2002-01-01

    Liquid metal cooled fast breeder reactors, in particular, pool type have many inherent and engineered safety features and hence a core disruptive accident (CDA) involving melt down of the whole core is a very low probable event ( -6 /ry). The important mechanical consequences such as straining of the main vessel including top shield, structural integrity of safety grade decay heat exchangers (DHX) and intermediate heat exchangers (IHX) sodium release to reactor containment building (RCB) through the penetrations in the top shield, sodium fire and consequent temperature and pressure rise in RCB are theoretically analysed using computer codes. Through the analyses with these codes, it is demonstrated that an energetic CDA capability to the maximum 100 MJ mechanical energy in PFBR can be well contained in the primary containment. The sodium release to RCB is 350 kg and pressure rise in RCB is ∼10 kPa. In order to raise the confidence on the theoretical predictions, very systematic experimental program has been carried out. Totally 67 tests were conducted. This experimental study indicated that the primary containment is integral. The main vessel can withstand the energy release of ∼1200 MJ. The structural integrity of IHX and DHX is assured up to 200 MJ. The transient force transmitted to reactor vault is negligible. The average water leak measured under simulated tests for 122 MJ work potential is about 1.8 kg and the maximum leak is 2.41 kg. Extrapolation of the measured maximum leak based on simulation principles yields ∼ 233 kg of sodium leak in the reactor. Based on the above-mentioned theoretical and experimental investigations, the design pressure of 20 kPa is used for PFBR

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

    International Nuclear Information System (INIS)

    Dievoet, J.P. van

    1987-01-01

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1978-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R D [United Kingdom Atomic Energy Authority, Risley (United Kingdom)

    1978-07-01

    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.

  14. 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.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

    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.

  16. Vibrations measurement in fast and PWR reactor study

    International Nuclear Information System (INIS)

    Tigeot, Y.; Epstein, A.; Hareux, F.

    1975-01-01

    In the past severe damages have occured in several nuclear reactors, by structural vibrations induced by the primary cooling flow. To avoid this kind of troubles, the SEMT makes studies for two different types of reactors. For the light pressurized water reactors, some tests have been made on the SAFRAN test loop which is a three loop 1/8 scale internal model of a 900 MWe reactor. This study is actually undertaken jointly with Framatome. Elsewhere, measurements have been made on the Phenix fast breeder sodium reactor, and studies are planned for the Super Phenix reactor [fr

  17. Status of fast reactor activities in the USSR

    International Nuclear Information System (INIS)

    Troyanov, M.F.; Rinejskij, A.A.

    1990-01-01

    Four fast reactors are in operation in the USSR now: BR-10, BOR-60, BN-350 and BN-600. Load factor of BN-600 reactor was in 1989 about 76%. On the basis of operational experience of running reactors design of more powerful commercial size BN-800 power reactor has been completed recently and construction work has started at two sites. The BN-1600 reactor is considered to be the prototype of future commercial reactors. In 1990, it was decided to extend its design approach with the aim to find some additional solutions to provide higher safety and better economics. (author). Figs and tabs

  18. High flux-fluence measurements in fast reactors

    International Nuclear Information System (INIS)

    Lippincott, E.P.; Ulseth, J.A.

    1977-01-01

    Characterization of irradiation environments for fuels and materials tests in fast reactors requires determination of the neutron flux integrated over times as long as several years. An accurate integration requires, therefore, passive dosimetry monitors with long half-life or stable products which can be conveniently measured. In addition, burn-up, burn-in, and burn-out effects must be considered in high flux situations and use of minimum quantities of dosimeter materials is often desirable. These conditions force the use of dosimeter and dosimeter container designs, measured products, and techniques that are different from those that are used in critical facilities and other well-characterized benchmark fields. Recent measurements in EBR-II indicate that high-accuracy results can be attained and that tie-backs to benchmark field technique calibrations can be accomplished

  19. The development of fast simulation program for marine reactor parameters

    International Nuclear Information System (INIS)

    Chen Zhiyun; Hao Jianli; Chen Wenzhen

    2012-01-01

    Highlights: ► The simplified physical and mathematical models are proposed for a marine reactor system. ► A program is developed with Simulink module and Matlab file. ► The program developed has the merit of easy input preparation, output processing and fast running. ► The program can be used for the fast simulation of marine reactor parameters on the operating field. - Abstract: The fast simulation program for marine reactor parameters is developed based on the Simulink simulating software according to the characteristics of marine reactor with requirement of maneuverability and acute and fast response. The simplified core physical and thermal model, pressurizer model, steam generator model, control rod model, reactivity model and the corresponding Simulink modules are established. The whole program is developed by coupling all the Simulink modules. Two typical transient processes of marine reactor with fast load increase at low power level and load rejection at high power level are adopted to verify the program. The results are compared with those of Relap5/Mod3.2 with good consistency, and the program runs very fast. It is shown that the program is correct and suitable for the fast and accurate simulation of marine reactor parameters on the operating field, which is significant to the marine reactor safe operation.

  20. On-line chemical sensors for applications in fast reactors

    International Nuclear Information System (INIS)

    Jayaraman, V.

    2015-01-01

    Hydrogen sensors are essential components of fast reactor sodium circuits. These sensors are needed in fast reactors for the immediate detection of any steam leak into sodium during reactor operation which can lead to failure of steam generator. Depending on the operating power of the reactor, sodium-water reaction results in either an increase in dissolved hydrogen level in sodium or an increase in hydrogen content of argon cover gas used above sodium coolant. Hence, on-line monitoring of hydrogen continuously in sodium and cover circuits helps in detection of any steam leak. In the event of accidental leak of high temperature sodium, it reacts with oxygen and moisture in air leading to sodium fires. These fires produce sodium aerosol containing oxides of sodium (Na 2 O and Na 2 O 2 ) and NaOH. For early detection of sodium fires, sensor systems based on sodium ionization detector, pH measurement and modulation of conductivity of graphite films are known in the literature. This presentation deals with the development of on-line sensors for these two applications. A diffusion based sensor using a thin walled nickel coil at 773 K and a sensitive thermal conductivity detector (TCD) has been developed for monitoring hydrogen levels in argon cover gas. This sensor has a lower detection limit of 30 ppm of hydrogen in argon. To extend the detection limit of the sensor, a surface conductivity based sensor has been developed which makes use of a thin film of semi-conducting tin oxide. Integration of this sensor with the TCD, can extend the lower detection limit to 2 ppm of hydrogen in cover gas. Electrochemical sensor based on sodium-beta-alumina has been designed, fabricated and its performance in laboratory and industrial environment was evaluated. This paper presents the logical development of these sensors highlighting their merits and limitations

  1. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.

    1983-04-01

    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)

  2. Creep buckling problems in fast reactor components

    International Nuclear Information System (INIS)

    Ramesh, R.; Damodaran, S.P.; Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

    1995-01-01

    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

  3. Implications of nuclear physics in the development of Fast Breeder Reactors

    International Nuclear Information System (INIS)

    Rapeanu, S.; Ilie, P.; Vasiliu, G.; Popescu, C.; Boeriu, S.; Constantinescu, D.; Mateescu, S.

    1980-08-01

    The purpose of this paper is to point out the involved aspects of nuclear physics in the calculation and design of the fast reactors. After a brief description of the advantages of using the fast reactors in the national economy, the national programs concerning this activity are presented. The structure and operation conditions of the liquid metal fast breeder reactor (LMFBR) are also reviewed. Then, the methods aimed to calculate the core, the burn-up, the reactor dynamics, the analysis of accidents, the shielding, as well as, the materials required in the fast reactor calculation, are shortly given. Further on, it deals with the nuclear data types connected to the fast reactor calculations, with accuracy requirements for nuclear data, as well as, with the present stage of nuclear data for fissile, fertile and structural materials. The requirements for new differential data measurements, new integral data and benchmark experiments are presented. Data adjustement methods are also summarized. Some aspects of the structural material behaviour in intense gamma radiation and neutron fields existing into a fast reactor are also presented in the last part of this paper. The concluding remarks are mentioned at the end of the paper. (author)

  4. Integral physics data for fast-reactor design; Donnees de physique integrale intervenant dans les etudes de reacteur a neutrons rapides; Integral'nye fizicheskie dannye dlya raschetov reaktorov na bystrykh nejtronakh; Datos fisicos integrales para el diseno de reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Loewenstein, W B; Meneghetti, D [Argonne National Laboratory, Argonne, IL (United States)

    1962-03-15

    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-{alpha}, 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

  5. Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-04-20

    The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treated separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s

  6. Construction schedule management of China Experimental Fast Reactor

    International Nuclear Information System (INIS)

    Wang Yue

    2012-01-01

    China Experimental Fast Reactor (CEFR) in the first Fast Reactor in China, which is one of large project of the National High Technology Research and Development Program ('863' Program). On 21 st July 2011, CEFR had succeeded to connect to power grid, the target of construction had come true. To a large item, schedule management is one of the most important management, this paper a overall discussion about CEFR item. It has proved that the management of CEFR project is scientific, normative and high-efficiency, it will be valuable for lager Fast Reactor item and designers in interrelated field. (author)

  7. Upgrading program of the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Yoshida, A.; Yogo, S.

    2001-01-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

  8. 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)

  9. A new safety approach in the design of fast reactors

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Marchaterre, J.F.; Waltar, A.E.

    1987-01-01

    A new approach to achieving fast reactor safety goals is becoming really apparent in the US Fast Reactor Program. Whereas the ''defense is best'' philosophy still prevails, there has been a tangible shift toward emphasizing passive mechanisms to protect the reactor and provide public safety---rather than relying on add-on active, engineered safety systems. This paper reviews the technical basis for this new safety approach and provides discussion on its implementation in current US liquid metal-cooled reactor designs. 4 refs., 4 figs

  10. Fast reactors fuel Cycle: State in Europe

    International Nuclear Information System (INIS)

    1991-01-01

    In this SFEN day we treat all aspects (economics-reactor cores, reprocessing, experience return) of the LMFBR fuel cycle in Europe and we discuss about the development of this type of reactor (EFR project) [fr

  11. Recent developments for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, N.; Nakamuria, K.; Morishita, M.; Shibamoto, H.; Nagashima, H.; Inoue, K.

    2005-01-01

    For realization of reliable and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute(JNC) and Japan Atomic Power Company(JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts through evaluation of their structural integrity, the research and development of 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)' is recognized as an essential theme. FDS focuses on particular failure modes of FRs such as ratchet deformation and creep fatigue damages due to cyclic thermal loads. To evaluate these modes, three main developments are in progress. One is 'Refinement of Failure Criteria' for particular modes of FRs. Next is development of 'Guidelines for Inelastic Design Analysis' in order to predict elastic plastic and creep behaviors. Furthermore, efforts are being made toward preparing 'Guidelines for Thermal Load Modeling' for FR component design where thermal loads are dominant. These studies were performed under the sponsorship of the Ministry of Economy, Trade and Industry of Japanese government. (authors)

  12. Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

    The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

  13. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    Matsuno, Y.; Bando, S.

    1981-03-01

    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.)

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

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    Paternoster, R.; Kimpland, R.; Jaegers, P.; McGhee, J.

    1994-01-01

    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

  16. Philosophy of safety evaluation on fast breeder reactor

    International Nuclear Information System (INIS)

    1981-01-01

    This is the report submitted from the special subcommittee on reactor safety standard to the Nuclear Safety Commission on October 14, 1980, and it was decided to temporarily apply this concept to the safety examination on fast breeder reactors. The examination and discussion of this report were performed by taking the prototype reactor ''Monju'' into consideration, which is to be the present target, referring to the philosophy of the safety evaluation on fast breeder reactors in foreign countries and based on the experiences in the fast experimental reactor ''Joyo''. The items applicable to the safety evaluation for liquid metal-cooled fast breeder reactors (LMFBR) as they are among the existing safety examination guidelines are applied. In addition to the existing guidelines, the report describes the matters to be considered specifically for core, fuel, sodium, sodium void, reactor shut-down system, reactor coolant boundary, cover gas boundary and others, intermediate cooling system, removal of decay heat, containment vessels, high temperature structures, and aseismatic property in the safety design of LMFBR's. For the safety evaluation for LMFBR's, the abnormal transient changes in operation and the phenomena to be evaluated as accidents are enumerated. In order to judge the propriety of the criteria of locating LMFBR facilities, the serious and hypothetical accidents are decided to be evaluated in accordance with the guideline for reactor location investigation. (Wakatsuki, Y.)

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

    International Nuclear Information System (INIS)

    Wydler, P.

    1984-01-01

    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

  18. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Durston, J.G.

    1976-01-01

    It is stated that in a liquid metal cooled fast breeder reactor wherein the core, intermediate heat exchangers and liquid metal pumps are immersed in a pool of coolant such as Na, the intermediate heat exchangers are suspended from the roof, and ducting is provided in the form of a core tank or shroud interconnected with 'pods' housing the intermediate exchangers for directing coolant from the core over the heat exchanger tubes and thence back to the main pool of liquid metal. Seals are provided between the intermediate heat exchanger shells and the walls of their 'pods' to prevent liquid metal flow by-passing the heat exchanger tube bundles. As the heat exchangers must be withdrawable for servicing, and because linear differential thermal expansion of the heat exchanger and its 'pod' must be accommodated the seals hitherto have been of the sliding kind, generally known as 'piston ring type seals'. These present several disadvantages; for example sealing is not absolute, and the metal to metal seal gives rise to wear and fretting by rubbing and vibration. This could lead to seizure or jamming by the deposition of impurities in the coolant. Another difficulty arises in the need to accommodate lateral thermal expansion of the ducting, including the core tank and 'pods'. Hitherto some expansion has been allowed for by the use of expansible bellow pairs in the interconnections, or alternatively by allowing local deformations of the core tank 'pods'. Such bellows must be very flexible and hence constitute a weak section of the ducting, and local deformations give rise to high stress levels that could lead to premature failure. The arrangement described seeks to overcome these difficulties by use of a gas pocket trapping means to effect a seal against vertical liquid flow between the heat exchanger shell and the wall of the heat exchanger housing. Full details of the arrangement are described. (U.K.)

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

    International Nuclear Information System (INIS)

    Poplavski, V.M.; Ashurko, Y.M.; Zverev, K.V.; Sarayev, O.M.; Oshkanov, N.N.; Korol'kov, A.S.

    1999-01-01

    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)

  20. Radioisotopes in the primary circuit of a fast reactor

    International Nuclear Information System (INIS)

    Berlin, M.; Cauvin, M.

    1976-01-01

    In the frame of the research performed to understand the behaviour of the radioactive isotopes of iodine in the primary coolant circuit of fast reactor, a simple theoretical model is proposed. Results concerning PHENIX and RAPSODIE are given

  1. Review of fast reactor activities at OECD (NEA)

    International Nuclear Information System (INIS)

    Stephens, M.

    1981-01-01

    The Committee on the Safety of Nuclear Installations initiated several reports in 1979. Status reports are published on: the role of fission gas release in case of fuel element failure; reactivity monitoring in a LMFBR at shutdown; increasing the reliability of fast reactor shutdown systems. A report is planned on the interactions between sodium and concrete. LMFBR safety issue that were studied are concerned with containment R and D; natural circulation cooling; and fuel failure modelling. Nuclear Development Division was concerned with Gas cooled fast reactors technology. Nuclear Science Division dealt with fast reactor physics and nuclear data for fast reactors. NEA Data Bank provides technical support and acts as a computer code library and nuclear data centre

  2. Fast wave current drive in reactor scale tokamaks

    International Nuclear Information System (INIS)

    Moreau, D.

    1992-01-01

    The IAEA Technical Committee Meeting on Fast Wave Current Drive in Reactor Scale Tokamaks, hosted by the Commissariat a l'Energie Atomique (CEA), Departement de Recherches sur la Fusion Controlee (Centres d'Etudes de Cadarache, under the Euratom-CEA Association for fusion) aimed at discussing the physics and the efficiency of non-inductive current drive by fast waves. Relevance to reactor size tokamaks and comparison between theory and experiment were emphasized. The following topics are described in the summary report: (i) theory and modelling of radiofrequency current drive (theory, full wave modelling, ray tracing and Fokker-Planck calculations, helicity injection and ponderomotive effects, and alternative radio-frequency current drive effects), (ii) present experiments, (iii) reactor applications (reactor scenarios including fast wave current drive; and fast wave current drive antennas); (iv) discussion and summary. 32 refs

  3. Slovakia: Proposal of movable reflector for fast reactor design

    International Nuclear Information System (INIS)

    Vrban, B.

    2015-01-01

    In fast reactors a larger migration area leading to a significant leak of neutrons can be observed because especially the transport cross-sections are in general smaller as compared to light water reactors. The utilization of a moveable reflector system in conjunction with dedicated safety control rods can increase the ability of accident managing due to enhanced escaping neutrons which otherwise would be reflected back into the fuel zone. The paper demonstrates the possibility of better controlling the transient reactor by additionally moving selected reflector subassemblies equipped with the neutron trap. The main purpose of the analysis of the Gas-cooled Fast Reactor (GFR) presented in the full paper is investigation of the kinetic parameters and of the control and reflector rod worth, as well as optimization of the parts used for partial reflector withdrawal. The results found in this study may serve for future design improvements of other designs such as the liquid metal cooled fast reactors

  4. Intermediate and fast neutron absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1987-10-01

    The experimental fuel channel EFC is created as one of the fast neutron fields at the RB reactor. The intermediate and fast neutron spectra in EFC are measured by activation technique. The intermediate and fast neutron absorbed doses are computed on the basis of these experimental results. At the end the obtained doses are compared. (author)

  5. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Pillai, P.; Chellapandi, P.; Chetal, S.C.; Vasudeva Rao, P.R.

    2013-01-01

    • Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

  6. Some basic concepts of fast breeder reactor safeguards

    International Nuclear Information System (INIS)

    Tkharev, E.; Walford, F.J.

    1987-04-01

    The range of discussion topics of this report is restricted to a few key areas of safeguards importance at Fast Breeder Reactors (FBR) only. The differences between thermal and fast reactors that may have safeguards significance in the case of FBRs are listed. The FBR principles of design are mentioned. The relevant safeguards objectives and criteria are given. The fundamental issues for safeguarding FBR are treated. An outline safeguards approach is presented. Model inspection activities are mentioned. 4 figs

  7. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    1989-07-01

    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

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

    International Nuclear Information System (INIS)

    Aliberti, G.; Palmiotti, G.; Salvatores, M.

    2007-01-01

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

  9. Technology development of fast reactor fuel reprocessing technology in India

    International Nuclear Information System (INIS)

    Natarajan, R.; Raj, Baldev

    2009-01-01

    India is committed to the large scale induction of fast breeder reactors beginning with the construction of 500 MWe Prototype Fast Breeder Reactor, PFBR. Closed fuel cycle is a prerequisite for the success of the fast reactors to reduce the external dependence of the fuel. In the Indian context, spent fuel reprocessing, with as low as possible out of pile fissile inventory, is another important requirement for increasing the share in power generation through nuclear route as early as possible. The development of this complex technology is being carried out in four phases, the first phase being the developmental phase, in which major R and D issues are addressed, while the second phase is the design, construction and operation of a pilot plant, called CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell. The third phase is the construction and operation of Demonstration of Fast Reactor Fuel Reprocessing Plant (DFRP) which will provide experience in fast reactor fuel reprocessing with high availability factors and plant throughput. The design, construction and operation of the commercial plant (FRP) for reprocessing of PFBR fuel is the fourth phase, which will provide the requisite confidence for the large scale induction of fast reactors

  10. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

    Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

    2013-01-01

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  11. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    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 ...

  12. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    Markoczy, G.; Hudina, M.; Richmond, R.; Wydler, P.; Stratton, R.W.; Burgsmueller, P.

    1980-03-01

    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 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  13. Design characteristics of research zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Nikolic, D.; Antic, D.; Zavaljevski, N.; Popovic, D.

    1990-01-01

    LASTA is a flexible zero power reactor with uranium and plutonium fuel designed for research in the neutron physics and in the fast reactor physics. Safety considerations and experimental flexibility led to the choice of a fixed vertical assembly with two safety blocks as the main safety elements, so that safety devices would be operated by gravity. The neutron and reactor physics, the control and safety philosophy adopted in our design, are described in this paper. Developed computer programs are presented. (author)

  14. Fast breeder reactors: can we learn from experience

    International Nuclear Information System (INIS)

    Keck, O.

    1981-01-01

    An economic analysis of FBRs, in particular the long-term benefits to be expected, with reference to the experience of the West German fast breeder reactor programme suggests ways of bringing more realism into governmental decisions on the development of new reactor types. It is suggested that if reactor manufacturers and utilities financed commercial-size demonstration plants from their own funds, then the government would get more realistic advice. (U.K.)

  15. SYNBURN: fast-reactor fuel-cycle program

    International Nuclear Information System (INIS)

    Pizzica, P.A.; Meneley, D.A.

    1976-01-01

    The SYNBURN computer program for fast reactors will calculate all the neutronics necessary to completely characterize the equilibrium cycle as well as the startup to equilibrium cycles. The program's run time is very short and this makes the program suitable for survey of parametric studies. It can search on the cycle time for a specified burnup, for the shim control necessary for criticality as well as feed enrichments and the enrichment ratio among core zones. SYNBURN synthesizes in a very simple fashion the one-dimensional fluxes in radial and axial geometry to achieve an approximate two-dimensional solution which agrees very well with the exact two-dimensional solution when measuring regional integrated quantities

  16. Nuclear data needs for fast breeder reactor shielding

    International Nuclear Information System (INIS)

    Oblow, E.M.; Perey, F.G.

    1978-11-01

    A review of neutron and gamma-ray cross section data needs for fast reactor shielding is presented in light of the recent advances made in assessing these needs through sensitivity studies. Total and partial cross sections and energy and angular distribution data for neutrons are surveyed as well as gamma-ray production cross sections. The strengths and deficiencies of currently available benchmark-quality integral experiments are also discussed with respect to their use in creating adjusted cross section libraries for design work. The availability of first round covariance data in ENDF/B-IV and plans for ENDF/B-V are also reviewed. This latter information makes it possible to quantitatively assess the quality of current cross section data libraries and also puts adjustment and data assessment procedures on a firmer basis

  17. Use of fast-spectrum reactors for actinide burning

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1991-01-01

    Finally, Integral Fast Reactor (IFR) pyroprocessing has been developed only in recent years and it appears to have potential as a relatively uncomplicated, effective actinide recovery process. In fact, actinide recycling occurs naturally in the IFR fuel cycle. Although still very much developmental, the entire IFR fuel cycle will be demonstrated on prototype-scale in conjunction with the EBR-II and its refurbished Fuel Cycle Facility starting in late 1991. A logical extension to this work, therefore, is to establish whether this IFR pyrochemical processing can be applied to extracting actinides from LWR spent fuel. This paper summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs. 4 figs., 4 tabs

  18. Remote maintenance in TOR fast reactor fuel reprocessing facility

    International Nuclear Information System (INIS)

    Eymery, R.; Constant, M.; Malterre, G.

    1986-11-01

    The TOR facility which is undergoing commissioning tests has a capacity of 5 T. HM/year which is enough for reprocessing all the Phenix fuel, with an excess capacity which is to be used for other fast reactors fuels. It is the result of enlargement and renovation of the old Marcoule pilot facility. A good load factor is expected through the use of equipment with increased reliability and easy maintenance. TOR will also be used to test new equipment developed for the large breeder fuel reprocessing plant presently in the design stage. The latter objective is specifically important for the parts of the plant involving mechanical equipment which are located in a new building: TOR 1. High reliability and flexibility will be obtained in this building thanks to the attention given to the integrated remote handling system [fr

  19. Linear and nonlinear stability analysis, associated to experimental fast reactors

    International Nuclear Information System (INIS)

    Amorim, E.S. do; Moura Neto, C. de; Rosa, M.A.P.

    1980-07-01

    Phenomena associated to the physics of fast neutrons were analysed by linear and nonlinear Kinetics with arbitrary feedback. The theoretical foundations of linear kinetics and transfer functions aiming at the analysis of fast reactors stability, are established. These stability conditions were analitically proposed and investigated by digital and analogic programs. (E.G.) [pt

  20. A review of fast reactor programme in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Masuno, Y [Experimental Fast Reactor Division, O-arai Engineering Center, PNC (Japan); Bando, S [Project Planning and Management Division, PNC, Minato-ku, Tokyo (Japan)

    1981-05-01

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report.

  1. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    Masuno, Y.; Bando, S.

    1981-01-01

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report

  2. Chemical surveillance of commercial fast breeder reactors

    International Nuclear Information System (INIS)

    Stamm, H.H.; Stade, K.Ch.

    1988-01-01

    After BN-600 (USSR) and SUPERPHENIX (France) were started succesfully, the international development of LMFBRs is standing at the doorstep of commercial use. For commercial use of LMFBRs cost reductions for construction and operation are highly desirable and necessary. Several nations developing breeder reactors have joined in a common effort in order to reach this aim by standardization and harmonization. On the base of more than 20 years of operation experience of experimental reactors (EBR-II, FFTF, RAPSODIE, DFR, BR-5/BR-10, BOR-60, JOYO, KNK-II) and demonstration plants (PHENIX, PFR, BN-350), possibilities for standardization in chemical surveillance of commercial breeder reactors without any loss of availability, reliability and reactor safety will be discussed in the following chapters. Loop-type reactors will be considered as well as pool-type reactors, although all commercial plants under consideration so far (SUPERPHENIX II, BN-800, BN-1600, CFBR, SNR-2, EFR) include pool-type reactors only. Table 1 gives a comparison of the Na inventories of test reactors, prototype plants and commercial LMFBRs

  3. Fast reactors: the future of nuclear energy

    International Nuclear Information System (INIS)

    Carvalho, H.G. de.

    1988-08-01

    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.) [pt

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

    International Nuclear Information System (INIS)

    1968-01-01

    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)

  5. Fast Thorium Molten Salt Reactors Started with Plutonium

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.; Mathieu, L.

    2006-01-01

    One of the pending questions concerning Molten Salt Reactors based on the 232 Th/ 233 U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233 U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233 U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233 U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/ 233 U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into 233 U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with 233 U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with 233 U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

  6. Sodium fires at fast reactors: RF status report

    International Nuclear Information System (INIS)

    Bagdasarov, Yu.E.; Buksha, Yu.K.; Drobyshev, A.V.; Zybin, V.A.; Ivanenko, V.N.; Kardash, D.Yu.; Kulikov, E.V.; Yagodkin, I.V.

    1996-01-01

    Scientific and engineering studies carried out in Russian Federation since 1992 up to 1996 in the sodium fire area and their main results are described. A review of activities on modification of the computer codes BOX and AERO developed at IPPE for calculating sodium fire consequences is given. Results of analysis of possible accidental situations at currently designed BN-800 reactor NPP with the use of these codes are presented. Sodium leaks occurring at our domestic fast reactors are briefly analyzed. Experimental work performed are described. Results of comparative analysis of common-cause and sodium fire hazards for fast reactor NPP are presented. (author)

  7. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

    Kalafati, D.D.; Petrov, S.I.

    1978-01-01

    Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

  8. Status of national programmes on fast reactors in Korea

    International Nuclear Information System (INIS)

    Kim, Y.I.; Hahn, D.

    2002-01-01

    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)

  9. The dissolver paradox as a coupled fast-thermal reactor

    International Nuclear Information System (INIS)

    Lutz, H.F.; Webb, P.S.

    1993-05-01

    The dissolver paradox is treated as coupled fast-thermal reactors. Each reactor is sub-critical but the coupling is sufficient to form a critical system. The practical importance of the system occurs when the fast system by itself is mass limited and the thermal system by itself is volume limited. Numerous 1D calculations have been made to calculate the neutron multiplication parameters of the separate fast and thermal systems that occur in the dissolver paradox. A model has been developed to describe the coupling between the systems. Monte Carlo calculations using the MCNP code have tested the model

  10. A review of the Italian fast reactor programme

    International Nuclear Information System (INIS)

    Pierantoni, F.; Tavoni, R.

    1984-01-01

    This review sums up the Italian situation in the field of the fast reactors on the eve of the fifth five year plan (1985-1989), in which the country undertakes to implement an important activity of research and development in the context of a greater European collaboration. Italian participation in the development of European nuclear power stations together with the completion of the PEC plant which will be used to develop a fuel element with the necessary economic and safety characteristics, remain the two principal goals of the Italian fast reactor programme. In 1983 the sum assigned by ENEA for fast reactors was about 220 billion lire of which 145 billion was for the PEC reactor

  11. 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)

  12. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1992-01-01

    In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

  13. A review of fast reactor program in Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

  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. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  16. 04 - Sodium cooled fast breeder fourth-generation reactors - The experimental reactor ALLEGRO, the other ways for fast breeder fourth-generation reactors

    International Nuclear Information System (INIS)

    2012-12-01

    The authors first present the technology of gas-cooled fast breeder reactors (basic principles, specific innovations, feasibility studies, fuel element, safety) and notably the ALLEGRO project (design options and expected performances, preliminary safety demonstration). Then, they present the lead-cooled fast-breeder reactor technology: interests and obstacles, return on experience, the issue of lead density, neutron assessment, transmutation potential, dosimetry, safety chemical properties and compatibility with the fuel, water, air and steels. The next part addresses the technology of molten-salt fast-breeder reactors: choice of the liquid fuel and geometry, reactor concept (difficulties, lack of past R and D), demonstration and demonstrators, international context

  17. Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1988-01-01

    The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs

  18. The computerized reactor period measurement system for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1996-01-01

    The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

  19. Computer measurement system of reactor period for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1997-01-01

    The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

  20. Unconventional liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.

    1989-06-01

    This report describes the rationale for, design of and analytical studies on an unconventional sodium-cooled power reactor, called the Trench Reactor. It derives its name from the long, narrow sodium pool in which the reactor is placed. Unconventional features include: pool shape; reactor shape (also long and narrow); reflector control; low power density; hot-leg primary pumping; absence of a cold sodium pool; large core boxes rather than a large number of subassemblies; large diameter metal fuel; vessel suspension from cables; and vessel cooling by natural circulation of building atmosphere (nitrogen) at all times. These features all seem feasible. They result in a system that is capable of at least a ten year reload interval and shows good safety through direct physical response to loss-of-heat-sink, loss-of-flow and limited-reactivity nuclear transients. 43 figs., 43 tabs

  1. Metallic uranium as fuel for fast reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de

    1988-01-01

    This paper presents a first overview of the use of metallic uranium and its alloys as an option for fuel for rapid reactors. Aspects are discussed concerning uranium alloys which present high solubility in the gamma phase. (author)

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1980-01-01

    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

  3. Status of fast reactor activities in the Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    Troyanov, M F; Rinejsjij, A A [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1992-07-01

    The power production program was developed before the disintegration of the USSR and CIS. This report covers therefore the current status of power production and consumption in in republics of the former USSR with a separate chapter on the status of nuclear power. It covers some general results concerned with fast reactors operational experience and BN-600 power plant operational experience. This includes radiological conditions at the BN-600 and reactor core operating experience. Separate chapters are devoted to BN-350, BOR-60, BR-10 and BN-800 reactors. Work devoted to large-size reactor design are described including research and development and fabrication.

  4. Uranium utilization of light water cooled reactors and fast breeders

    International Nuclear Information System (INIS)

    Stojadinovic, Timm

    1991-08-01

    The better uranium utilization of fast breeder reactors as compared with water cooled reactors is one argument in favour of the breeder introduction. This report tries to quantify this difference. It gives a generally valid formalism for the uranium utilization as a function of the fuel burnup, the conversion rate, fuel cycle losses and the fuel enrichment. On the basis of realistic assumptions, the ratio between the utilizations of breeder reactors to that of light water cooled reactors (LWR) amounts to 180 for the open LWR cycle and 100 in case of plutonium recycling in LWRs

  5. Status of fast reactor activities in the Russian Federation

    International Nuclear Information System (INIS)

    Troyanov, M.F.; Rinejsjij, A.A.

    1992-01-01

    The power production program was developed before the disintegration of the USSR and CIS. This report covers therefore the current status of power production and consumption in in republics of the former USSR with a separate chapter on the status of nuclear power. It covers some general results concerned with fast reactors operational experience and BN-600 power plant operational experience. This includes radiological conditions at the BN-600 and reactor core operating experience. Separate chapters are devoted to BN-350, BOR-60, BR-10 and BN-800 reactors. Work devoted to large-size reactor design are described including research and development and fabrication

  6. FAST: An advanced code system for fast reactor transient analysis

    International Nuclear Information System (INIS)

    Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh

    2005-01-01

    One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems

  7. Argentinean activities related to Fast Reactors

    International Nuclear Information System (INIS)

    Azpitarte, Osvaldo

    2012-01-01

    CNEA objectives in the area of Generation IV nuclear reactors: Implement a programme for the monitoring of the global progress of new technologies for Generation IV nuclear reactors and their fuel cycles, in order to generate and assess associated lines of R&D. – Perform studies and evaluations for defining the Generation IV line or lines on which CNEA would be interested; – Promote the participation on specific international projects; – Implementation of experimental facilities

  8. Licensing issues for inherently safe fast reactors

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Lee, S.; Okrent, D.

    1986-01-01

    There has been considerable interest recently in a new generation of liquid metal reactor (LMR) concepts in the US. Some significant changes in regulatory philosophy will be required if the anticipated cost advantages of inherently safe designs are to be achieved. The defense in depth philosophy will need to be significantly re-evaluated in the context of inherently safe reactors. It is the purpose of this paper to begin such a re-evaluation of this regulatory philosophy

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

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2014-01-01

    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)

  10. Irradiation Experiments on Plutonium Fuels for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Frost, B. R.T.; Wait, E. [Atomic Energy Research Establishment Harwell, Berks. (United Kingdom)

    1967-09-15

    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)O{sub 2} -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 Degree-Sign 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 UO{sub 2} and(U,Pu)O{sub 2} 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 {approx} 3% bum-up, and that a low density powder accommodated the swelling without excessive straining of the can. A subsequent

  11. Identification of fast power reactivity effect in nuclear power reactor

    International Nuclear Information System (INIS)

    Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

    1987-01-01

    A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

  12. Materials requirements for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Bennett, J.W.; Horton, K.E.

    1978-01-01

    Materials requirements for Liquid Metal Fast Breeder Reactors (LMFBRs) are quite varied with requisite applications ranging from ex-reactor components such as piping, pumps, steam generators and heat exchangers to in-reactor components such as heavy section reactor vessels, core structurals, fuel pin cladding and subassembly flow ducts. Requirements for ex-reactor component materials include: good high temperature tensile, creep and fatigue properties; compatibility with high temperature flowing sodium; resistance to wear, stress corrosion cracking, and crack propagation; and good weldability. Requirements for in-reactor components include most of those cited above for ex-reactor components as supplemented by the following: resistance to radiation embrittlement, swelling and radiation enhanced creep; good neutronics; compatibility with fuel and fission product materials; and resistance to mass transfer via flowing sodium. Extensive programs are currently in place in a number of national laboratories and industrial contractors to address the materials requirements for LMFBRs. These programs are focused on meeting the near term requirements of early LMFBRs such as the Fast Flux Test Facility and the Clinch River Breeder Reactor as well as the longer term requirements of larger near-commercial and fully-commercial reactors

  13. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1982-01-01

    The fast breeder reactor development project in Japan has been in progress for the past twelve months and will be continued this fiscal year, from April 1982 through March 1983, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1981. The 1982 year budget for R and D work and for construction of a prototype fast breeder reactor MONJU is approximately 20 and 27 billion yen respectively, excluding wages for the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaged in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, power increase from 50 MWt to 75 MWt was made in July 1979 and six operational cycles at 75 MWt were completed in December 1981. With respect to the prototype reactor MONJU, progress toward construction has been made and an environmental impact statement of the reactor was approved by the authorities concerned, and the licensing of the first step was completed at the end of 1981. Preliminary design studies of a large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  14. Core concept of fast power reactor with zero sodium void reactivity

    International Nuclear Information System (INIS)

    Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

    1991-01-01

    The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  17. Seismic appraisal test of control rod drive mechanism of China experiment fast reactor

    International Nuclear Information System (INIS)

    Song Qing; Yang Hongyi; Jing Yueqing; Wen Jing; Liu Guijuan; Sun Lei

    2008-01-01

    The structure of the control rod drive mechanism in pool type sodium-cooled fast reactor is the characterized by long, thin, and geometric nonlinearity, and the seismic load is multiple activation. The anti-seismic evaluation is always paid great attention by the countries developing the technology worldwide. This article introduces the seismic appraisal test of the control rod drive mechanism of China Experimental Fast Reactor (CEFR) performed on a seismic platform which is vertical shaft style and multiple activation. The result of the test shows the structural integrity and the function of the control rod drive mechanism could meet the design requirements of the earthquake intensity. (authors)

  18. History of fast reactor development in U.S.A.-I

    International Nuclear Information System (INIS)

    Ninokata, Hisashi; Sasao, Nobuyki

    2007-01-01

    History and present state of fast reactor was reviewed in series. As a history of fast reactor development in U.S.A. - I, this third lecture presented the dawn of the fast reactor development in the USA. The first fast reactor was the Clementine reactor with plutonium fuels and mercury coolant. The LAMPRE-1 reactor was the first sodium cooled and molten plutonium reactor. Experimental breeder reactor (EBR-1) was the first reactor to produce electricity and four kinds of fuels were loaded. Zero-power reactors were constructed to conduct reactor physics experiments on fast reactors. Today there are renewed interests in fast reactors due to their ability to fission actinides and reduce radioactive wastes. (T. Tanaka)

  19. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1996-01-01

    The main R and D results of Japanese activities are summarized as follows: (1) the experimental 140 MW(th) sodium cooled fast reactor 'Joyo' provided abundant experimental data and excellent operational records, attaining more than 50,000 hours of operation since its first criticality in 1977; (2) the prototype 280 MW(e) fast reactor 'Monju' reached initial criticality on 5 April 1994; presently Monju is under the cold shutdown state because of secondary sodium leak on 8 December 1995, and multiple cause investigations of the sodium leak are being performed; (3) the Japan Atomic Power Company is promoting design studies for demonstration fast reactor (DFBR) with a power output of 600 MW(e) and R and D for DFBR are being conducted under the cooperation of governmental and private sectors. (author)

  20. Economic analysis of fast reactor fuel cycle with different modes

    International Nuclear Information System (INIS)

    Ding Xiaoming

    2014-01-01

    Because of limitations on the access to technical and economic data and the lack of effective verification, the lack of in-depth study on the economy of fast reactor fuel cycle in China. This paper introduces the analysis and calculation results of the levelized cost of electricity (LCOE) under three different fuel cycle modes including fast reactor fuel cycle carried out by Massachusetts Institute of Technology (MIT). The author used the evaluation method and hypothesis parameters provided by the MIT to carry out the sensitivity analysis for the impact of the overnight cost, the discount rate and changes of uranium price on the LCOE under three fuel cycle modes. Finally, some suggestions are proposed on the study of economy in China's fast reactor fuel cycle. (authors)