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

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)

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)

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)

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)

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)

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

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

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

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

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

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

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

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

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

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

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.

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.

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

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.

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)

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

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

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)

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

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

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

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

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

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.

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

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)

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

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

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

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.

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

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

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

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

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

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

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.

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.

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

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.

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)

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

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

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)

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

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

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

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

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

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

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

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.)

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

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

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.

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.

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.

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

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

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)

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)

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...

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

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

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.)

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)

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)

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

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

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.

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

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

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)

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

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

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.)

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.)

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

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

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)

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

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

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

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

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)

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.)

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.

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

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)

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

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

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

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)

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

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.

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

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

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

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.)

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)

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

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

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)

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

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

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

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

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.

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

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.)

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)

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

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

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

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,

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

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)

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)

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)

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

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

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.

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.

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

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

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.)

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)

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

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

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)

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

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.

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

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)

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.)

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

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”.

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

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

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

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

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

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

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)

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)

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.

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)

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

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

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

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.)

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)

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

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.)

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)

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

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

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

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

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)

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

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

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.

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)

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

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

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.)

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)

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)

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.)

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.

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.

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.

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.

Modeling the behavior of metallic fast reactor fuels during extended transients

International Nuclear Information System (INIS)

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

1993-01-01

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

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

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

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

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

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

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

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

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

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

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)

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)

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.)

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

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

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.

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

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

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

International Nuclear Information System (INIS)

1998-01-01

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

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)

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

Status of the R and D activities on fast reactors and ADS in Brazil

International Nuclear Information System (INIS)

Maiorino, Jose Rubens

2001-01-01

Research and Development in Nuclear Science and Technology is conducted by Research Institutes of the Brazilian Nuclear Energy Commission. In Fast Reactor, R and D activities started in the sixties, and in 1972 a small Na loop (100 kW) was constructed. At the same time, during the seventies at IPEN, research in cooperation with GA for Gas Cooled Fast Breeder Reactor was conducted. The motivation of such research was Thorium Fuel Cycle. As a result of this research a Helium Loop was constructed and a Split Table Critical Assembly (ZPR) was designed. During the eighties, an agreement with ANSALDO-NIRA resulted in an acquisition of a Sodium Loop for Thermohydraulics studies, however it never had been assembled. At the same time, a concept of a Binary Breeder Reactor using two cycles, Th and U, was developed. During the nineties, a National Program to conduct R and D (pyroprocess; U-Zr Metallic Fuel; HT-9; Electromagnetic Pump; and a conceptual design of a Experimental Reactor (60/20 MWth/MWe)) was proposed, however it was closed at the end of the decade. Now, only academic research is being conducted, and it is summarized in this report. Basically, they are: an integral lead fast reactor concept for developing countries, and an alternative concept for a fast energy amplifier accelerator driven system. The first is an combination of best characteristics of the American Integral Fast Reactor and the Russian Lead Cooled Reactor. The second is a conceptual design of ADS helium cooled imbedded in a solid lead subcritical array of fuel, using more than one point of spallation trying to reduce the requirement for energy and current of the accelerator

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)

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

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)

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.)

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

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

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)

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.)

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.)

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

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

Modeling the behavior of metallic fast reactor fuels during extended transients

International Nuclear Information System (INIS)

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

1992-01-01

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

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)

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)

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)

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

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)

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

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

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

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.

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)

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

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.

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)

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.)

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

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.

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.

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

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

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)

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)

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.)

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

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)

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.

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

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.

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)

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.

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

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)

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

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

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)

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.

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

Current status and perspective of advanced loop type fast reactor in fast reactor cycle technology development project

International Nuclear Information System (INIS)

Niwa, Hajime; Aoto, Kazumi; Morishita, Masaki

2007-01-01

After selecting the combination of the sodium-cooled fast reactor (SFR) with oxide fuel, the advanced aqueous reprocessing and the simplified pelletizing fuel fabrication as the most promising concept of FR cycle system, 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was finalized in 2006. Instead, a new project, Fast Reactor Cycle Technology Development Project (FaCT Project) was launched in Japan focusing on development of the selected concepts. This paper describes the current status and perspective of the advanced loop type SFR system in the FaCT Project, especially on the design requirements, current design as well as the related innovative technologies together with the development road-map. Some considerations on advantages of the advanced loop type design are also described. (authors)

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

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...

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.)

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)

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.

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)

Development of design procedures for fast reactors in the United Kingdom

International Nuclear Information System (INIS)

Rose, R.T.; Tomkins, B.; Townley, C.H.A.

1989-01-01

A considerable amount of research has been carried out in the United Kingdom during the past two decades to quantify the factors which control the integrity of structural components. The work which has been aimed at understanding the performance of structures at high temperature, is particularly relevant to the Fast Reactor. At the same time, because of the need to demonstrate the tolerance to defects in the low temperature as well as the high temperature components, defect assessment criteria are also of great importance. Emphasis is now being given to the development of design procedures specifically for Fast Reactors, making use of the research so far completed. The United Kingdom proposals are being integrated with those from France, Federal Republic of Germany and Italy as part of the European collaborative venture. The paper outlines the major developments which are currently in hand, and brings up to date the review of United Kingdom activities presented at Tokyo in 1986. (author)

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.)

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.)

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

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

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

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)

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.)

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.

International standardization of safety requirements for fast reactors

International Nuclear Information System (INIS)

2011-06-01

Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

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.)

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

International Nuclear Information System (INIS)

1998-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-04-01

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

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.

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)

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)

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)

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)

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

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.)

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.

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

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)

Survey of creep data on structural materials of fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, S.

1977-11-01

The reactor vessels and other components of fast breeder reactor is affected by high neutron irradiation at elevated temperatures. However, in this regard, related test data on creep property of component materials and welds at elevated temperatures are a few in Japan, and especially, there are no data available on the irradiation effect. It will take 3 to 7 years before the results of currently planned research and development on prototype fast breeder become available. On the other hand, establishment of design base for prototype fast breeder and other needs call for early solution to such problems. The Committee should, therefore, collect from documents the latest data on experiments on structural materials overseas and in our country, and survey and analyze the problems in order to proceed with the future research and development in the most effective way. It was for this purpose that the Fourth Subcommittee at Technical Research Association for Integrity of Structures at Elevated Service Temperatures was commissioned by Power Reactor and Nuclear Fuel Development Corporation to conduct the examination and study of related data by establishing Group 41G. This collection of data is the compilation of the above results. (author)

International Working Group on Fast Reactors Sixth Annual Meeting. Summary Report

International Nuclear Information System (INIS)

1973-01-01

The Agenda of the Meeting was as follows: 1. Review of IWGFR Activities - 1a. Approval of the minutes of the Fifth IWGFR Meeting. 1b. Report by Scientific Secretary regarding the activities of the Group. 2. Comments on National Programmes on Fast Breeder Reactors. 3. International Coordination of the Schedule for Major Fast Reactor Meetings and other major international meetings having a predominant fast reactor interest. 4. Consideration of Conferences on Fast Reactors. 4a. IAEA Symposium on Fuel and Fuel Elements for Fast Reactors, Brussels, Belgium 2-6 July 1973. 4b. International Symposium on Physics of Fast Reactors, Tokyo, Japan, 16 to 23 October 1973. 4c. International Conference on Fast Reactor Power Stations, London, UK, 11 to 14 March 1974 . 4d. Suggestions of the IWGFR members on other conferences. 5. Consideration of a Schedule for Specialists' Meetings in 1973-74. 6. Other Business - 6a. First-aid in Sodium Burns. 6b. Principles of Good Practice for Safe Operation of Sodium Circuits. 6c. Bibliography on Fast Reactors. 7. The Date and Place of the Seventh Annual Meeting of the IWGFR

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

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

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

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.

UKAEA fast reactor project research and development programme on fuel element cladding and sub-assembly wrapper materials

International Nuclear Information System (INIS)

Harries, D.R.

1977-01-01

Research and development work on fuel element component (cladding, subassembly wrappers, etc.) materials for the U.K. sodium cooled fast reactor programme has been conducted at the United Kingdom Atomic Energy Authority (UKAEA) establishments at Dounreay, Harwell, Risley, and Springfields during the past fifteen years or so. This work has formed an integral part of, and has been co-ordinated by, the UKAEA Fast Reactor Project and has involved close liaison with the Nuclear Power Company (NPC) and the Central Electricity Generating Board (CEGB). The research and development were initially related to the Prototype Fast Reactor (PFR) but the scope has now been extended to cover the first Civil Fast Reactor (CFR1), which has recently been re-designated the Civil Demonstration Fast Reactor (CDFR). The paper outlines the present status of the development of sodium cooled fast reactors in the U.K. and proceeds to summarize the principal PFR and CDFR core and fuel element parameters which have determined the planning and direction of the fuel element materials programme. The current position on the fuel element cladding and wrapper research and development programme is reviewed, and the facilities and future irradiation programme to be carried out in PFR are described

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

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.

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)

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.

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Nascimento, Jamil Alves do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, A. dos; Nascimento, J.A. do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

Argentinean integrated small reactor design and scale economy analysis of integrated reactor

International Nuclear Information System (INIS)

Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

2000-01-01

This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

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)

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)

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

International Nuclear Information System (INIS)

Alekseev, P.; Fomichenko, P.; Mikityuk, K.; Nevinitsa, V.; Shchepetina, T.; Subbotin, S.; Vasiliev, A.

2001-01-01

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

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)

Foundational development of an advanced nuclear reactor integrated safety code

International Nuclear Information System (INIS)

Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

2010-01-01

This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

Foundational development of an advanced nuclear reactor integrated safety code.

Energy Technology Data Exchange (ETDEWEB)

Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

2010-02-01

This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

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

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)

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

Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

Energy Technology Data Exchange (ETDEWEB)

Scaller, K; Vrillon, B

1980-02-01

Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component.

Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

International Nuclear Information System (INIS)

Scaller, K.; Vrillon, B.

1980-01-01

Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component

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.

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

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.

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.

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.)

Status of fast reactors and ADS programmes in France in 2001

International Nuclear Information System (INIS)

Astegiano, J.C.

2002-01-01

Status of French fast reactor and ADS program in France covers the following topics: data on power generation from NPPs; status of fast reactors, namely Rapsodie, Phenix and Super Phenix; research and development programs concerned with fast gas cooled and sodium cooled reactors

ORACLE: an adjusted cross-section and covariance library for fast-reactor analysis

International Nuclear Information System (INIS)

Yeivin, Y.; Marable, J.H.; Weisbin, C.R.; Wagschal, J.J.

1980-01-01

Benchmark integral-experiment values from six fast critical-reactor assemblies and two standard neutron fields are combined with corresponding calculations using group cross sections based on ENDF/B-V in a least-squares data adjustment using evaluated covariances from ENDF/B-V and supporting covariance evaluations. Purpose is to produce an adjusted cross-section and covariance library which is based on well-documented data and methods and which is suitable for fast-reactor design. By use of such a library, data- and methods-related biases of calculated performance parameters should be reduced and uncertainties of the calculated values minimized. Consistency of the extensive data base is analyzed using the chi-square test. This adjusted library ORACLE will be available shortly

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

US/DOE Man-Machine Integration program for liquid metal reactors

International Nuclear Information System (INIS)

D'Zmura, A.P.; Seeman, S.E.

1985-03-01

The United States Department of Energy (DOE) Man-Machine Integration program was started in 1980 as an addition to the existing Liquid Metal Fast Breeder Reactor safety base technology program. The overall goal of the DOE program is to enhance the operational safety of liquid metal reactors by optimum integration of humans and machines in the overall reactor plant system and by application of the principles of human-factors engineering to the design of equipment, subsystems, facilities, operational aids, procedures and environments. In the four years since its inception the program has concentrated on understanding the control process for Liquid Metal Reactors (LMRs) and on applying advanced computer concepts to this process. This paper describes the products that have been developed in this program, present computer-related programs, and plans for the future

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)

The irradiation test program for transmutation in the French Phenix fast reactor

International Nuclear Information System (INIS)

Guidez, J.; Chaucheprat, P.; Fontaine, B.; Brunon, E.

2004-01-01

Put on commercial operation in July 1974, the French fast reactor Phenix reached a 100 000 hours operation time in september 2003. When the French law relative to long lived radioactive waste management was promulgated on December 1991, priority was given to Phenix to be run as a research reactor and to carry on a wide irradiation program dedicated to study transmutation of minor actinides and long-lived fission products. After a major renovation program required to extend the reactor lifetime, Phenix power buildup took place in 2003. Experimental irradiations have been loaded in the core, involving components for heterogeneous and homogeneous transmutation modes, americium targets, technetium 99 metal pins and isolated isotopes for integral cross-sections measurements. Associated post- irradiated examination programs are already underway or planned. With new experiments to be loaded in the core in 2006 the Phenix reactor remains to be a powerful tool providing an important experimental data on fast reactors and on transmutation of minor actinides and long-lived fission products, as well as it will contribute to gain further experience in the framework of the GENERATION IV International Forum. (authors)

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

Status of fast reactor control rod development in the United Kingdom

International Nuclear Information System (INIS)

Kelly, B.T.

1984-01-01

The two large fast reactors constructed in the United Kingdom, that is the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR) differed substantially in their control systems. DFR was controlled by variation of the neutron leakage from the core while PFR uses conventional control rods containing neutron absorbing materials. This paper describes the development of the PFR control systems, the progressive design of the control systems for the prototype Civil Fast Reactor (CFR) and the supporting research and development programmes. (author)

Commission of the European Communities review of fast reactor activities, March 1981

Energy Technology Data Exchange (ETDEWEB)

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

1981-05-01

The Commission of the European Communities continued its activities in the field of fast reactors development essentially in the frame of the Fast Reactor Coordinating Committee (FRCC) and by execution of a Reactor Programme at its Joint Research Center (JRC). The study was concerned with introducing fast reactors into European Community, elaboration of preliminary safety criteria and guidelines for typical fast reactor accidents; codes and standards; LMFBR safety, fuel, fuel cycle safety.

Commission of the European Communities review of fast reactor activities, March 1981

International Nuclear Information System (INIS)

Balz, W.

1981-01-01

The Commission of the European Communities continued its activities in the field of fast reactors development essentially in the frame of the Fast Reactor Coordinating Committee (FRCC) and by execution of a Reactor Programme at its Joint Research Center (JRC). The study was concerned with introducing fast reactors into European Community, elaboration of preliminary safety criteria and guidelines for typical fast reactor accidents; codes and standards; LMFBR safety, fuel, fuel cycle safety

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

NEA activities in preserving, evaluating and applying data from fast reactor experiments

International Nuclear Information System (INIS)

Gulliford, Jim; Cornet, S.M.; Hill, I.; Yamaji, A.

2013-01-01

Conclusions: Progress to date: • Extensive programme of work to preserve and evaluate data from integral experiments has been established since the mid 1990s. • NEA Data Bank maintains and distributes several databases of these integral experiments, notably through the ICSBEP and IRPhE projects. • More recently programmes of work have been established to help preserve data from the UK Fast Reactor Programme and from various experiments related to minor actinide management. • Data obtained from these programmes are made available to the nuclear science community to provide high quality benchmarks against which modelling methods can be validated. • Involvement of younger scientists and engineers to work alongside well-established experts in the process of evaluating the information is a highly efficient means of transmitting tacit knowledge to the new generation of nuclear specialists. Conclusions: Looking ahead - • Further development of Databases and Database tools, e.g. – improved coverage of fast reactor experiments, MAs; – improved treatment of correlations in uncertainties between experiments; – production of sensitivities to facilitate identification of similar experiments. • Continuation securing UK archives and creating framework for information: – Start identifying suitable integral experiments for inclusion in NEA databases

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

International Nuclear Information System (INIS)

Tsoulfanidis, N.; Jankhah, M.H.

1979-01-01

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

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

International Nuclear Information System (INIS)

Smith, R.D.

1983-01-01

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

Research and development issues for fast reactor structural design standard (FDS)

International Nuclear Information System (INIS)

Kasahara, Naoto; Ando, Masanori; Morishita, Masaki

2003-01-01

For realization of safe 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 and validate their structural integrity, the research and development of 'Fast Reactor Structural Design Standard (FDS)' is recognized as an essential theme. FDS considers general characteristics of FRs and design needs for their rationalization. Three main subjects were settled in research and development issues of FDS. One is rationalization of failure criteria' taking characteristic design conditions into account. Next is development of 'a guideline on inelastic analysis for design' in order to predict elastic plastic and creep behaviours of high temperature components. Furthermore, efforts are being made toward preparing a guideline on thermal loads modeling' for FR component design where thermal loads are dominant. (author)

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)

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)

Development, Fabrication and Characterization of Fuels for Indian Fast Reactor Programme

International Nuclear Information System (INIS)

Kumar, Arun

2013-01-01

Development of Fast Reactor fuels in India started in early Seventies. The successful development of Mixed Carbide fuels for FBTR and MOX fuel for PFBR have given confidence in manufacture of fuels for Fast Reactors. Effort is being put to develop high Breeding Ratio Metallic fuel (binary/ternary). Few fuel pins have been fabricated and is under test irradiation. However, this is only a beginning and complete fuel cycle activities are under development. Metal fuelled Fast Reactors will provide high growth rate in Indian Fast Reactor programme

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

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.

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

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

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

Utilization of MVP for research on fast reactor

International Nuclear Information System (INIS)

Yokoyama, Kenji

2001-01-01

Utilization of the continuous energy Monte-Carlo code, MVP, for research on fast reactor in Power Reactor and Nuclear Fuel Development Corporation(PNC) is described. In this report, three types of utilization are reviewed; (1) a comparison of the eigenvalues calculated by MVP with the results by the deterministic methods, (2) an improvement of U-238 reaction rate evaluation in JUPITER experimental Analysis and (3) an evaluation of heterogeneity effects for Am reaction rates of the moderated subassemblies. Since the results of MVP can be used as references, MVP is very useful code in research on fast reactor. It is one of indispensable tools in order to verify the models in the deterministic methods. Furthermore, it can be used so as to investigate the new concept reactors, such as a reactor aiming to transmute minor actinides(MA). On the other hand, a problem of the variance reduction remains. Especially, a small reactivity cannot be estimated by MVP because of large variances. The development of a Monte-Carlo method for a small reactivity calculation will promote the utilization of MVP for research on fast reactor. (author)

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

International Nuclear Information System (INIS)

Pesic, M.

1994-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

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

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

International Nuclear Information System (INIS)

1992-01-01

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

UK fast reactor components - sodium removal decontamination and requalification

Energy Technology Data Exchange (ETDEWEB)

Donaldson, D M [FRDD, UKAEA, Risley (United Kingdom); Bray, J A; Newson, I H [UKAEA, Dounreay Nuclear Power Establishment, Thurso (United Kingdom)

1978-08-01

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

UK fast reactor components - sodium removal decontamination and requalification

International Nuclear Information System (INIS)

Donaldson, D.M.; Bray, J.A.; Newson, I.H.

1978-01-01

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

Proceedings of 'workshop on Pb-alloy cooled fast reactor'

International Nuclear Information System (INIS)

Kim, Sang Ji; Kim, Yong Hee; Hong, Ser Gi

2003-06-01

The objective of 'Workshop on Pb-Alloy Cooled Fast Reactor', held in Taejeon, Korea on May 6, 2003, is to enhance the basic knowledge in this area by facilitating the exchange of information and discussions about problematic area of design aspects. There were five presentations from three different countries and about 25 participants gathered during the workshop. The topics covered in the workshop include benefits and drawbacks of Pb-alloy and Sodium coolant, two Pb-alloy cooled 900 MWt reactor designs using both B4C rods and NSTs, BREST-300 breakeven reactor and transmutation effectiveness of LLFPs in the typical thermal/fast neutron systems. The generic conclusion for the Pb-alloy cooled fast reactor from this workshop is as follows: 1) It has a potential to satisfy the goals established for the Generation-IV reactor concepts, so it has a bright future. 2) As a fast neutron system with a moderate breeding or a conversion, it is flexible in its roles and has superior safety characteristics over sodium coolant because of Pb-alloy's chemical inertness with water/air and high boiling temperature

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

International Nuclear Information System (INIS)

1969-01-01

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

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

The development of accurate data for the desing of fast reactors

International Nuclear Information System (INIS)

Rossouw, P.A.

1976-04-01

The proposed use of nuclear power in the generation of electricity in South Africa and the use of fast reactors in the country's nuclear porgram, requires a method for fast reactor evluation. The availability of accurate neutron data and neutronics computation techniques for fast reactors are required for such an evaluation. The reacotr physics and reactor parameters of importance in the evaluation of fast reacotrs are discussed, and computer programs for the computation of reactor spectra and reacotr parameters from differential nuclear data are presented in this treatise. In endeavouring to increase the accuracy in fast reactor design, two methods for the improvement of differential nuclear data were developed and are discussed in detail. The computer programs which were developed for this purpose are also given. The neutron data of the most important fissionable and breeding nuclei (U 235 x U 238 x Pu 239 and Pu 240 ) are adjusted using both methods and the improved neutron data are tested by computation with an advanced neutronics computer program. The improved and orginal neutron data are compared and the use of the improved data in fast reactor design is discussed

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

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

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)

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

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

International Nuclear Information System (INIS)

1996-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-04-01

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

Fuel Development For Gas-Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

M. K. Meyer

2006-06-01

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

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.

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

Intrinsically secure fast reactors with dense cores