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Sample records for fast reactors razrabotka

  1. 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.)

  2. 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

  3. 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...

  4. 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.)

  5. 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)

  6. 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)

  7. 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

  8. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

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

  9. Fast reactors worldwide

    International Nuclear Information System (INIS)

    Hall, R.S.; Vignon, D.

    1985-01-01

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

  10. 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

  11. 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.

  12. 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.)

  13. 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)

  14. 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

  15. 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.)

  16. 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

  17. 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.

  18. 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.)

  19. 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

  20. 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,

  1. 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.)

  2. 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)

  3. 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

  4. 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)

  5. 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.)

  6. 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

  7. Safeguards challenges of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Ko, H. S.

    2010-01-01

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

  8. 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

  9. 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)

  10. 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)

  11. 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

  12. 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.

  13. 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)

  14. 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

  15. 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.)

  16. 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.)

  17. 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

  18. 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

  19. Integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

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

  20. 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

  1. 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

  2. Economic Issues of Fast Reactor in China

    International Nuclear Information System (INIS)

    Yang Hongyi

    2013-01-01

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

  3. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

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

  4. 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)

  5. 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)

  6. 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.)

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

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

  18. Decommissioning of fast reactors after sodium draining

    International Nuclear Information System (INIS)

    2009-11-01

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

  19. 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)

  20. 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)

  1. 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)

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

    International Nuclear Information System (INIS)

    Dombey, N.

    1980-01-01

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

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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)

  8. 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

  9. 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.)

  10. 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

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. 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)

  18. 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

  19. 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)

  20. 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.

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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)

  6. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

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

  7. 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

  8. 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.

  9. 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

  10. 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.)

  11. 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.

  12. 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.

  13. Advances by the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  14. 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)

  15. 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.)

  16. 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)

  17. 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.

  18. 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

  19. 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.

  20. 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

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

    International Nuclear Information System (INIS)

    Stephens, M.

    1981-01-01

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

  2. 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

  3. Status of fast reactor activities in Russia

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  4. 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)

  5. The Integral Fast Reactor

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  6. 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

  7. 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

  8. 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.)

  9. 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

  10. 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.

  11. 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)

  12. 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.

  13. 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)

  14. 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.)

  15. 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.)

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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.)

  1. 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.

  2. 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)

  3. 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.)

  4. 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

  5. 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.

  6. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1985-01-01

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

  7. 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)

  8. 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)

  9. 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)

  10. A review of the Italian fast reactor programme

    International Nuclear Information System (INIS)

    Pierantoni, F.; Tavoni, R.

    1984-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  12. 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.

  13. 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

  14. 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

  15. The integral fast reactor concept

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  16. 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

  17. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1989-01-01

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

  18. Fission energy: The integral fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1989-01-01

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

  19. Aspects of the fast reactors fuel cycle

    International Nuclear Information System (INIS)

    Zouain, D.M.

    1982-06-01

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

  20. 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

  1. 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.

  2. 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)

  3. 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.

  4. 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)

  5. 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)

  6. 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)

  7. 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)

  8. 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

  9. 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.

  10. 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

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

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

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

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

    International Nuclear Information System (INIS)

    Tanaka, Nobuo

    2017-01-01

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

  16. 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.

  17. 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.

  18. 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.)

  19. 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

  20. 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)

  1. 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.

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

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

    International Nuclear Information System (INIS)

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

    1980-08-01

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

  4. 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)

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. Intrinsically secure fast reactors with dense cores

    International Nuclear Information System (INIS)

    Slessarev, Igor

    2007-01-01

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

  12. Status of fast reactor activities in Brazil

    International Nuclear Information System (INIS)

    Menezes, Artur

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  14. Programme and current status of fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    Suita, T.; Oyama, A.

    1977-01-01

    In 1967 the Japan Atomic Energy Commission revised her long term programme after a two year study for giving principles to her nuclear energy development programme, which indicated the dominant role of nuclear energy mid 1980's in the electric power generation and stressed the necessity of developing fast breeder reactors. It also recommended to organize a nucleus to undertake this nation-wide project, bringing together the total capability available throughout the country. Accordingly, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established in 1967 to develop two sodium-cooled fast reactors, an experimental fast reactor of about 100 MW thermal and a prototype fast breeder reactor of about 300 MW electrical, both using mixed oxide fuels. Construction of the experimental fast reactor started in 1970 and was essentially completed at the end of in 1974. The precommissioning test was followed in parallel with re-evaluating quality assurance of all systems. Physics test will be initiated around the end of 1976. The conceptual design of the prototype fast breeder reactor is now toward its final stage. Surveys on its proposed site have just started. Construction will start in 1978. Beside R and D works conducted by many organizations in Japan as well as under the international cooperation, several key test facilities were installed by PNC itself to conduct in-sodium test of full-size prototype components including 50 MW steam generators and post-irradiation-examination of fuels and materials. Recently an interim report was issued to an ad-hoc committee organized by JAEC to evaluate future prospect of the fuel cycle and power reactors. This recommended start of construction of the prototype reactor as scheduled and the large demonstration reactor to be followed to the prototype. Thus the fast breeder reactor is indicated as the most indispensable in 1990's

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

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1997-01-01

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

  17. Fast reactor development strategy targets study in China

    International Nuclear Information System (INIS)

    Xu Mi

    2008-01-01

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

  18. 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

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

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

    International Nuclear Information System (INIS)

    Wu Xingman

    2011-01-01

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

  2. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1996-01-01

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

  3. The problems of thermohydraulics of prospective fast reactor concepts

    International Nuclear Information System (INIS)

    Sedov, A.A.

    2000-01-01

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

  4. The Integral Fast Reactor (IFR) concept

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  5. Use of fast reactors for actinide transmutation

    International Nuclear Information System (INIS)

    1993-03-01

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

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

    International Nuclear Information System (INIS)

    Vrban, B.

    2015-01-01

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

  7. Total decay heat estimates in a proto-type fast reactor

    International Nuclear Information System (INIS)

    Sridharan, M.S.

    2003-01-01

    Full text: In this paper, total decay heat values generated in a proto-type fast reactor are estimated. These values are compared with those of certain fast reactors. Simple analytical fits are also obtained for these values which can serve as a handy and convenient tool in engineering design studies. These decay heat values taken as their ratio to the nominal operating power are, in general, applicable to any typical plutonium based fast reactor and are useful inputs to the design of decay-heat removal systems

  8. 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

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

    International Nuclear Information System (INIS)

    2000-01-01

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

  10. 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)

  11. 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)

  12. Discharges from a fast reactor reprocessing plant

    International Nuclear Information System (INIS)

    Barnes, D.S.

    1987-01-01

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

  13. Effect of Fast Neutron to MA/PU Burning/Transmutation Characteristic Using a Fast Reactor

    International Nuclear Information System (INIS)

    Marsodi; Lasman, As Natio; Kimamoto, A.; Marsongkohadi; Zaki, S.

    2003-01-01

    MA/Pu burning/transmutation has been studied and evaluated using fast neutrons. Generally, neutron density at this fast burner reactor and transmutation has spectrum energy level around 0.2 MeV with wide enough variation, i.e. from low neutron spectrum to its peak is 0.2 MeV. This neutron spectrum energy level depends on the kind of cooler material or fuel used. Neutron spectrum higher than fast power reactor neutron spectrum is found by means of changing oxide fuel by metallic fuel and changing natrium cooler material by metallic or gas cooler material. This evaluation is conducted by various variations in accordance with the kind of fuel or cooler, MA/Pu fractions and fuel comparison fraction with respect to its cooler in order to get better neutron usage and MA/Pu burning speed. Reactor calculation evaluation in this paper was conducted with 26-group nuclear data cross section energy spectrum. The main purpose of the discussion is to know the effect of fast neutrons to burning/transmutation MA/Pu using fast neutrons

  14. 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

  15. Materials requirements for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  16. Fast reactor system factors affecting reprocessing plant design

    International Nuclear Information System (INIS)

    Allardice, R.H.; Pugh, O.

    1982-01-01

    The introduction of a commercial fast reactor electricity generating system is very dependent on the availability of an efficient nuclear fuel cycle. Selection of fuel element constructional materials, the fuel element design approach and the reactor operation have a significant influence on the technical feasibility and efficiency of the reprocessing and waste management plants. Therefore the fast reactor processing plant requires liaison between many design teams -reactor, fuel design, reprocessing and waste management -often with different disciplines and conflicting objectives if taken in isolation and an optimised approach to determining several key parameters. A number of these parameters are identified and the design approach discussed in the context of the reprocessing plant. Radiological safety and its impact on design is also briefly discussed. (author)

  17. Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

    International Nuclear Information System (INIS)

    Friess, Friederike Renate

    2017-01-01

    According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

  18. Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Friess, Friederike Renate

    2017-07-12

    According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  20. Fast-reactor-data testing of ENDF/B-V at ORNL

    International Nuclear Information System (INIS)

    Wright, R.Q.; Ford, W.E. III; Lucius, J.L.; Webster, C.C.; Marable, J.H.

    1982-01-01

    The Cross Section Evaluation Working Group (CSEWG) is coordinating a program to assess the adequacy of ENDF/B-V cross sections for both fast- and thermal-reactor design applications. A secondary goal is to evaluate cross-section processing codes, cross-section libraries, and radiation-transport codes. Fast reactor data testing (FRDT) goals are accomplished, in part, by comparison of calculated results with documented performance parameters of CSEWG fast reactor benchmarks and with results obtained by other data testers. The purpose of this paper is to describe the results of FRDT at Oak Ridge National Laboratory

  1. Fast reactors: R and D targets and outlook for their introduction

    International Nuclear Information System (INIS)

    Poplavsky, V.; Barre, B.; Aizawa, K.

    1997-01-01

    In this paper the current status of fast reactors development is briefly outlined, including experimental, demonstration, and commercial installations. Data on the experience gained in development and operation of NPPs with reactors of this type are presented. The issues are discussed in connection with possibilities of fast reactor development in the nuclear power structure for the near (up to 2010-2020) and distant future. In the final part of the paper, an analysis is given of possible ways for R and D development in the field of NPPs with fast neutron reactors. (author)

  2. Fast reactors will eat nuclear waste from LWR

    Energy Technology Data Exchange (ETDEWEB)

    Tokiwai, Moriyasu [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Res., Lab.

    1999-12-01

    Although nuclear power is one of the indispensable energy sources to support modern life styles in developed countries, it becomes harder and harder to increase its capacity. Newspaper reported that there are numbers of evidences showing the suppression effect on cancer by the low level of radiation. It is expected for public people that the fear for radiation induced harm on health will mitigate through the explanation based on scientific evidences. Safe management of radioactive waste is one of the most serious issues to be solved. The neutron at fast reactors can eat more effectively the long lived several nuclear waste materials from light water reactor system, The key issue is to develop the fast reactor fuel cycle system technologies that are more economical, more proliferation resistant and higher breeding ratio. The Metallic Fuel Cycle is one of the options for the future fast breeder reactor and its related fuel cycle that enable to give the answer for the radioactive waste issues. The attractiveness of the metallic fuel cycle concept is briefly described. (author)

  3. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    1998-01-01

    This report describes the development and activities on fast reactor in Japan for the period of April 1996 - March 1997. During this period, the 30th duty cycle operation has been started in the Experimental Fast Reactor ''''Joyo''''. The cause investigation on the sodium leak incident has completed and the safety examination are being performed in the Prototype Fast Breeder Reactor ''''Monju''''. The three years design study since FY1994 on the plant optimization of the Demonstration FBR has been completed by the Japan Atomic Power Company (JAPC). Related research and development works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which is composed of Power Reactor and Nuclear Fuel Development Corporation (PNC), JAPC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). In November 1996, the Japan Atomic Energy Commission (JAEC) established a Social Gathering Meeting to discuss generally the significance of FBR development in Japan for the future. (author)

  4. Fast reactors will eat nuclear waste from LWR

    International Nuclear Information System (INIS)

    Tokiwai, Moriyasu

    1999-01-01

    Although nuclear power is one of the indispensable energy sources to support modern life styles in developed countries, it becomes harder and harder to increase its capacity. Newspaper reported that there are numbers of evidences showing the suppression effect on cancer by the low level of radiation. It is expected for public people that the fear for radiation induced harm on health will mitigate through the explanation based on scientific evidences. Safe management of radioactive waste is one of the most serious issues to be solved. The neutron at fast reactors can eat more effectively the long lived several nuclear waste materials from light water reactor system, The key issue is to develop the fast reactor fuel cycle system technologies that are more economical, more proliferation resistant and higher breeding ratio. The Metallic Fuel Cycle is one of the options for the future fast breeder reactor and its related fuel cycle that enable to give the answer for the radioactive waste issues. The attractiveness of the metallic fuel cycle concept is briefly described. (author)

  5. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    1989-07-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. The status of fast reactor technology development in China

    International Nuclear Information System (INIS)

    Xu Mi

    2000-01-01

    Considering the future clean energy supply in China, a rather consistent opinion is to develop nuclear power step by step with the contribution from a supplementary one up to an important one. The large scale utilization of nuclear energy obviously determines the interest in fast breeders; China right now already has about 300 GWe total electricity capacity using conventional energy resources. As the first step for fast reactor technology development in the country, the China Experimental Fast Reactor (CEFR) project is still under detail design stage, which is a sodium cooled pool type fast reactor with 65 MW thermal power matched with a turbine-generator of 25 MW. The ordering of the components is continuing. The site is ready and the steel works for the 3 m x 69 m x 82.5 m foundation base of reactor building are being arranged layer by layer. The review to the PSAR by the China National Nuclear Safety Administration (CNNSA) is going to the final stage, if everything goes smoothly. The first pouring of the concrete for the reactor building will be in the middle of the year 2000. The brief introduction of the CEFR design, safety characteristics, the main results of the safety analysis and design test demonstration are given in the paper. (author)

  8. Sodium fires at fast reactors: RF status report

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  9. A glossary of terms for fast reactors

    International Nuclear Information System (INIS)

    Wheeler, R.C.

    1979-04-01

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

  10. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  12. 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

  13. Fast breeder reactor-block antiseismic design and verification

    International Nuclear Information System (INIS)

    Martelli, A.; Forni, M.

    1988-02-01

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

  14. Fast reactor development programme in France during 1995

    International Nuclear Information System (INIS)

    Le Rigoleur, C.

    1996-01-01

    In 1995, the total amount of electricity produced in France was 471 TWh, out of which 358.2 TWh (76 %) were produced by nuclear power plants, 36.9 TWh (7.8 %) by conventional thermal plants, and 75.5 TWh (16 %) by hydraulic plants. The net electrical power consumption was 368.7 TWh. At the end of 1995, 'Electricite de France' had 54 PWR units in operation. The availability factor for these units was maintained at 81%. 1995 was marked by a decrease of unexpected shutdowns (1.8% in 1995 instead of 2.2% in 1994), a new reduction in programmed shutdown periods, and a good safety level was maintained. In the field of Fast Reactors, the main events of 1995 were the following. At the end of December 1994, the PHENIX reactor was authorized to perform its 49th cycle at 350 MW th (143 MWe). This 49th cycle was completed without any significant problems on April 7, 1995. During the remainder of the year, the reactor had been shut down in order to carry out several tasks within the scope of the ten-year extension of the PHENIX reactor's lifetime. Concerning the CREYS-MALVILLE plant (SUPER-PHENIX) the first part of the year was devoted to repairing argon leak of one of the IHX. Authorization to restart the reactor was given on August 22. The end of the year was beset by a number of minor incidents. The reactor was restarted at the end of 1995 and reactor power was increased by successive steps (30% Pn (Nominal Power) up to February 6 1996; followed by 50 %...). The 'Decret d'Autorisation de Creation' stipulates that because of its prototype character, SUPER PHENIX will have to be operated under conditions explicitly giving priority to safety and knowledge acquisition, with an objective of research and demonstration. In this context, the so-called 'knowledge acquisition' programme designed to prove the capacity of a large FBR to produce electricity on an industrial scale, to test the consumption of plutonium and minor actinides in a large fast reactor, as well as to provide

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

    International Nuclear Information System (INIS)

    Ishiguro, Y.

    1983-01-01

    In order to assure the continued utilization of fission energy, development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed as the best type for future Brazilian nuclear systems. The inherent safety characteristics are superior to current FBRs and an efficient utilization of the abundant thorium is possible. A first step and a basic tool for the development of FBR technologies is the construction and operation of an experimental fast reactor (EFR). A series of core designs for a 90 MW EFR is studied and possible options and the magnitudes of principal parameters are identified. Flexible modifications of the core and sufficiently high fast fluxes for fuel and materials irradiations appear possible. (Author) [pt

  16. Some basic concepts of fast breeder reactor safeguards

    International Nuclear Information System (INIS)

    Tkharev, E.; Walford, F.J.

    1987-04-01

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

  17. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-01-01

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

  18. Integral Fast Reactor concept

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  19. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

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

    1993-06-01

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

  20. Fast reactor versions: elements of choice

    International Nuclear Information System (INIS)

    Tassart, J.; Zerbib, J.C.

    1984-01-01

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

  1. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  4. Advanced liquid metal fast breeder reactor designs

    International Nuclear Information System (INIS)

    Sayles, C.W.

    1978-01-01

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

  5. Materials development for fast reactor applications

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-15

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

  6. Safety Analysis Of Actinide Recycled Fast Power Reactor

    International Nuclear Information System (INIS)

    Taufik, Mohammad

    2001-01-01

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

  7. 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”.

  8. 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

  9. Environmental sensitivity studies for Gen-IV roadmap fast reactor scenario

    International Nuclear Information System (INIS)

    Jeong, Chang Joon

    2004-03-01

    The environmental effect of the self-sufficient fast reactor scenario, which is considered as one of the full fissile plutonium and transuranic recycle scenario in Gen-IV roadmap, has been analyzed by using the dynamic analysis method. Through the parametric calculations for the fast reactor deployment time and capacity, the environmental effects of the fuel cycle for important parameters such as the amount of spent fuel and the combined amounts of plutonium and minor actinides were estimated and compared to those of the once-through LWR fuel cycle. The results of the sensitivity calculations showed that an early deployment of the fast reactor with a high capacity can reduce the accumulation of spent fuel by up to 37%. Furthermore, the recycling of plutonium and minor actinides can reduce the key repository parameter (long term decay heat). Therefore the favorable environmental effects can be expected with the implementation of the symbiotic fast reactor scenario

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

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

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

  11. 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

  12. 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

  13. Design characteristics of zero power fast reactor Lasta

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  14. FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

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

    International Nuclear Information System (INIS)

    Imbert, Ch.

    1997-01-01

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

  16. 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

  17. 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.

  18. 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.

  19. Activities of the OECD-NEA in the field of fast reactors

    International Nuclear Information System (INIS)

    Royen, J.

    1977-01-01

    The OECD-NEA is performing the following activities in the field of fast reactors: Held ad hoc meetings of senior experts on safety, development and economics of LMFBR type reactors; publishing a Nuclear Safety Research Index (the index is now expanded to cover fast reactors) and distribution; collect test computer programmes, as well as neutron data

  20. The IAEA Activities in the Field of Fast Reactors Technology Development

    International Nuclear Information System (INIS)

    Monti, Stefano

    2011-01-01

    Main activities of the IAEA Programme on Fast Reactor: Carry out Collaborative Research Projects (CRPs) of common interest to the TWG-FR Member States in the field of FRs and ADS; Secure Training and Education in the field of fast neutron system physics, technology and applications; Support Fast Reactor data retrieval and knowledge preservation activities in MSs; Provide support to IAEA Nuclear Safety and Security Department for preparation of fast reactor Safety standards / requirements / guides. IAEA TWG-FR Functions: Provide advice and guidance, and marshal support in their countries for implementation of IAEA’s programmatic activities in the area of advanced technologies and R&D for fast reactors and sub-critical hybrid systems for energy production and for utilization/transmutation of long-lived nuclides; Provide a forum for information and knowledge sharing on national and international development programs; Act as a link between IAEA’s activities in the specific area of the TWG-FR and national scientific communities, delivering information from and to national communities

  1. Fast Reactor Knowledge Management at IGCAR, India

    International Nuclear Information System (INIS)

    Kuriakose, K.K.

    2013-01-01

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

  2. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

    Shoai Tehrani, Bianka; Da Costa, Pascal

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

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

  4. A review of the Indian fast reactor programme

    International Nuclear Information System (INIS)

    Chetal, S.C.

    1989-01-01

    Fast Breeder Test Reactor (FBTR) in India is ready for restart. Satisfactory progress has been made in the design of Prototype Fast Breeder Reactor (PFBR). Conceptual design work for the important systems and components has been completed. Cost estimation is in progress. Detailed project report for the financial sanction is under completion stage and is planned to be submitted to the Government this year. Draft Safety criteria prepared by a sub-committee on behalf of the Regulatory Board have been discussed and will be issued shortly. (author)

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  6. Fast Thorium Molten Salt Reactors Started with Plutonium

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  7. Status of national programmes on fast reactors in Korea

    International Nuclear Information System (INIS)

    Kim, Y.I.; Hahn, D.

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Pierantoni, F.

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-07-01

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

  10. 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)

  11. A review of the United Kingdom fast reactor programme

    International Nuclear Information System (INIS)

    Bramman, J.I.; Hickey, H.B.; Whitlow, W.H.; Frew, J.D.; Gregory, C.V.

    1990-01-01

    Total energy consumption in the UK in 1989 was 340 million tonnes of coal or coal equivalent, made up as follows: coal 31%, petroleum 35%, natural gas 24%, nuclear electricity 8%, hydroelectricity 1% and imported electricity 1%. About half of the nuclear electricity generated came from 14 Advanced Gas-Cooled Reactors (AGRs) and about half from the 24 older gas-cooled Magnox reactors, one Steam-Generating Heavy-Water Reactor (SGHWR) and one fast reactor (the Prototype Fast Reactor, PFR, at Dounreay). The privatization of the Electricity Supply Industry (ESI) in the UK is proceeding. On 9 November 1989, however, it was announced by the Secretary of State for Energy that the privatization plan would be changed and that the CEGB's nuclear stations were to remain in state ownership, through the formation of an additional company, Nuclear Electric. At the same time, the Secretary of State for Scotland announced the formation of a similar state-owned company, Scottish Nuclear. Nuclear Electric was asked, in the interim, to examine priorities in the whole nuclear field with particular reference to the improvement of the economics and performance of existing reactors, to the development of the Sizewell and alternative reactors and to the development of longer-term options such as the fast reactor and fusion. Nuclear Electric has been asked to formulate its new policy by June 1990. The PFR programme will continue to be funded by the UK government until March 1994. AEA Technology is endeavouring to find alternative funding to maintain the operation of the PFR until at least the year 2000. The House of Commons Select Committee on Energy stated in its report that the fast reactor ''is a matter for the British Government to foster as a long-term option for the generation of electricity in this country'', and recommended that in the interim the Government reassesses its position on this new technology in the light of increasing concern about CO 2 emissions and the long

  12. Aspects of 238Pu production in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Osaka, Masahiko; Koyama, Shin-ichi; Tanaka, Kenya; Itoh, Masahiko; Saito, Masaki

    2005-01-01

    Experimental determination of 238 Pu in 237 Np samples irradiated in the experimental fast reactor JOYO was done as part of the demonstration of 238 Pu production from 237 Np in fast reactors within the framework of the protected Pu production project, which aims at reinforcement of proliferation resistance of Pu by increasing the 238 Pu isotopic ratio. 238 Pu production amount in the irradiated 237 Np samples was determined by a radioanalytical technique. Aspects of 238 Pu production were examined on the basis of the present radioanalysis. The 238 Pu production amount depends on the neutron spectrum which can range from that of a typical fast reactor to a nearly epi-thermal spectrum. It is concluded that the fast reactor has not only high potential for use in protected Pu production, but also as an incinerator for excess Pu

  13. Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

    This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

  14. 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

  15. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

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

  16. Status report on fast reactor recycle and impact on geologic disposal

    International Nuclear Information System (INIS)

    Bauer, T. H.; Morris, E. E.; Wigeland, R. A.

    2007-01-01

    The GNEP program envisions continuing the use of light-water reactors (LWRs), with the addition of processing the discharged, or spent, LWR fuel to recover actinide and fission product elements, and then recycling the actinide elements in sodium-cooled fast reactors. Previous work has established the relationship between the processing efficiencies of spent LWR fuel, as represented by spent PWR fuel, and the potential increase in repository utilization for the resulting processing waste. The purpose of this current study is to determine a similar relationship for the waste from processing spent fast reactor fuel, and then to examine the wastes from the combination of LWRs and fast reactors as would be deployed with the GNEP approach

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

    International Nuclear Information System (INIS)

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

    1993-05-01

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

  18. Uranium alloys for using in fast breeder reactors

    International Nuclear Information System (INIS)

    Moura Neto, C.; Pires, O.S.

    1988-08-01

    The U-Zr and U-Ti alloys are studied, given emphasis to the high solute solubility in gamma phase of uranium, which is suitable for using as metal fuel in fast breeder reactors. The alloys were prepared in electron beam furnaces and submitted to X-ray diffraction, X-ray fluorescence, microhardness, optical metallography, and chemical analysis. The obtained values are good agreements with the literature data. The study shows that the U-Zr presents better characteristics than the U-Ti for using as fuel in fast breeder reactors. (M.C.K.) [pt

  19. 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

  20. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    Poette, C.; Guedeney, P.; Stainsby, R.; Mikityuk, K.; Knol, S.

    2013-01-01

    Gas Cooled Fast Reactors: Conclusion - GFR: an attractive longer term option allowing to combine Fast spectrum & Helium coolant benefits; • Innovative SiC fuel cladding solutions were found; • A first design confirming the encouraging potential of the reactor system Design improvements are nevertheless recommended and interesting tracks have been identified (core & system design, DHR system); • The GFR requires large R&D needs to confirm its potential (fuel & core materials, specific Helium technology); • ALLEGRO prototype studies are the first step and are drawing the R&D priorities

  1. 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)

  2. 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)

  3. Current status of fast reactor physics

    International Nuclear Information System (INIS)

    Hummel, H.H.

    1979-01-01

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

  4. Application of Candle burnup to small fast reactor

    International Nuclear Information System (INIS)

    Sekimoto, H.; Satoshi, T.

    2004-01-01

    A new reactor burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. An equilibrium state was obtained for a large fast reactor (core radius is 2 m and reflector thickness is 0.5 m) successfully by using a newly developed direct analysis code. However, it is difficult to apply this burnup strategy to small reactors, since its neutron leakage becomes large and neutron economy becomes worse. Fuel enrichment should be increased in order to sustain the criticality. However, higher enrichment of fresh fuel makes the CANDLE burnup difficult. We try to find some small reactor designs, which can realize the CANDLE burnup. We have successfully find a design, which is not the CANDLE burnup in the strict meaning, but satisfies qualitatively its characteristics mentioned at the top of this abstract. In the final paper, the general description of CANDLE burnup and some results on the obtained small fast reactor design are presented.(author)

  5. Vibrations measurement in fast and PWR reactor study

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  6. Research of three-dimensional transient reactivity feedback in fast reactor

    International Nuclear Information System (INIS)

    Xu Li; Shi Gong; Ma Dayuan; Yu Hong

    2013-01-01

    To solve the three-dimensional time-spatial kinetics feedback problems in fast reactor, a mathematical model of the direct reactivity feedback was proposed. Based on the NAS code for fast reactor and the reactivity feedback mechanism, a feedback model which combined the direct reactivity feedback and feedback reflected by the cross section variation was provided for the transient calculation. Furthermore, the fast reactor group collapsing system was added to the code, thus the real time group collapsing calculation could be realized. The isothermal elevated temperature test of CEFR was simulated by using the code. By comparing the calculation result with the test result of the temperature reactivity coefficient, the validity of the model and the code is verified. (authors)

  7. Substantiation of physical concepts of fast reactors in Russia: experience and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

    2008-07-01

    The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

  8. Review of fast reactor activities in India (1983-84)

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1984-01-01

    The last year was very significant for the Indian Nuclear Energy Programme as the first indigeneously built heavy water moderated natural uranium reactor called Madras Atomic Power Plant Unit-I was made operational and connected to the grid. The power level has been gradually increased and the reactor has been operating at a power level of 200 MWe (temporarily limited by Plutonium build up during approach to equilibrium core loading). The 'plutonium peak' will be crossed shortly clearing the way for raising the reactor to the full power of 235 MWe gross. The second unit of MAPP, is well advanced and barring unforeseen difficulties, is expected to become operational during this financial year. This has been a big morale booster for the programme in general and the Fast Reactor Programme in particular as plutonium produced in these reactors is expected to be the inventory for Prototype Fast Breeder Reactors. It may be recalled that in the last report to this group, a reference was made to initiation of some preliminary design studies for such a reactor

  9. Transitioning nuclear fuel cycles with uncertain fast reactor costs

    Energy Technology Data Exchange (ETDEWEB)

    Phathanapirom, U.B., E-mail: bphathanapirom@utexas.edu; Schneider, E.A.

    2016-06-15

    This paper applies a novel decision making methodology to a case study involving choices leading to the transition from the current once-through light water reactor fuel cycle to one relying on continuous recycle of plutonium and minor actinides in fast reactors in the face of uncertain fast reactor capital costs. Unique to this work is a multi-stage treatment of a range of plausible trajectories for the evolution of fast reactor capital costs over time, characterized by first-of-a-kind penalties as well as time- and unit-based learning. The methodology explicitly incorporates uncertainties in key parameters into the decision-making process by constructing a stochastic model and embedding uncertainties as bifurcations in the decision tree. “Hedging” strategies are found by applying a choice criterion to select courses of action which mitigate “regrets”. These regrets are calculated by evaluating the performance of all possible transition strategies for every feasible outcome of the uncertain parameter. The hedging strategies are those that preserve the most flexibility for adjusting the fuel cycle strategy in response to new information as uncertainties are resolved.

  10. Transitioning nuclear fuel cycles with uncertain fast reactor costs

    International Nuclear Information System (INIS)

    Phathanapirom, U.B.; Schneider, E.A.

    2016-01-01

    This paper applies a novel decision making methodology to a case study involving choices leading to the transition from the current once-through light water reactor fuel cycle to one relying on continuous recycle of plutonium and minor actinides in fast reactors in the face of uncertain fast reactor capital costs. Unique to this work is a multi-stage treatment of a range of plausible trajectories for the evolution of fast reactor capital costs over time, characterized by first-of-a-kind penalties as well as time- and unit-based learning. The methodology explicitly incorporates uncertainties in key parameters into the decision-making process by constructing a stochastic model and embedding uncertainties as bifurcations in the decision tree. “Hedging” strategies are found by applying a choice criterion to select courses of action which mitigate “regrets”. These regrets are calculated by evaluating the performance of all possible transition strategies for every feasible outcome of the uncertain parameter. The hedging strategies are those that preserve the most flexibility for adjusting the fuel cycle strategy in response to new information as uncertainties are resolved.

  11. Breeding gains of sodium-cooled oxide-fueled fast reactors

    International Nuclear Information System (INIS)

    Mougniot, J.C.; Barre, J.Y.; Clauzon, P.; Ciacometti, C.; Neviere, G.; Ravier, J.; Sichard, B.

    Calculated values are presented for the breeding gains of French fast reactors, and the experimental uncertainties are discussed. The effect of various choices of planning on the breeding gains is next analyzed within the framework of classical concepts. In the final part, a new concept involving ''heterogeneous cores'' with a single enrichment zone is presented. This concept permits a significant improvement in the breeding gain and doubling time of fast reactors. (U.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Rosen, S

    1979-07-01

    This document was prepared by the Office of the Program Director for Nuclear Energy, U.S. Department of Energy (USDOE). It sets forth the status and current activities for the development of fast breeder technology in the United States. In April 1977 the United States announced a change in its nuclear energy policy. Concern about the potential for the proliferation of nuclear weapons capability emerged as a major issue in considering whether to proceed with the development, demonstration and eventual deployment of breeder reactor energy systems. Plutonium recycle and the commercialization of the fast breeder were deferred indefinitely. This led to a reorientation of the nuclear fuel cycle program which was previously directed toward the commercialization of fuel reprocessing and plutonium recycle to the investigation of a full range of alternative fuel cycle technologies. Two major system evaluation programs, the Nonproliferation Alternative Systems Assessment Program (NASAP), which is domestic, and the International Nuclear Fuel Cycle Evaluation (INFCE), which is international, are assessing the nonproliferation advantages and other characteristics of advanced reactor concepts and fuel cycles. These evaluations will allow a decision in 1981 on the future direction of the breeder program. In the interim, the technologies of two fast breeder reactor concepts are being developed: the Liquid Metal Fast Breeder Reactor (LMFBR) and the Gas Cooled Fast Reactor (CFR). The principal goals of the fast breeder program are: LMFBR - through a strong R and D program, consistent with US nonproliferation objectives and anticipated national electric energy requirements, maintain the capability to commit to a breeder option; investigate alternative fuels and fuel cycles that might offer nonproliferation advantages; GCFR - provide a viable alternative to the LMFBR that will be consistent with the developing U.S. nonproliferation policy; provide GCFR technology and other needed

  13. Critique of the economic case for the fast reactor. Chapter 12

    International Nuclear Information System (INIS)

    Sweet, C.

    1980-01-01

    The subject is discussed under the headings: introduction; critique of the cost stereotypes; the fast reactor fuel cycle (reprocessing and refabrication costs; out-of-reactor time and economic optimisation; the doubling time and the logistics of the fast breeder programme). (U.K.)

  14. IAEA Activities in the Area of Fast Reactors and Related Fuels and Fuel Cycles

    International Nuclear Information System (INIS)

    Monti, S.; Basak, U.; Dyck, G.; Inozemtsev, V.; Toti, A.; Zeman, A.

    2013-01-01

    Summary: • The IAEA role to support fast reactors and associated fuel cycle development programmes; • Main IAEA activities on fast reactors and related fuel and fuel cycle technology; • Main IAEA deliverables on fast reactors and related fuel and fuel cycle technology

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

    International Nuclear Information System (INIS)

    2011-01-01

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

  16. 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)

  17. Advanced design of fast reactor-membrane reformer (FR-MR)

    International Nuclear Information System (INIS)

    Tashimo, M.; Hori, I.; Yasuda, I.; Shirasaki, Y.; Kobayashi, K.

    2004-01-01

    A new plant concept of nuclear-produced hydrogen is being studied using a Fast Reactor-Membrane Reformer (FR-MR). The conventional steam methane reforming (SMR) system is a three-stage process. The first stage includes the reforming, the second contains a shift reaction and the third is the separation process. The reforming process requires high temperatures of 800 ∼ 900 deg C. The shift process generates heat and is performed at around 200 deg C. The membrane reforming has only one process stage under a nonequilibrium condition by removing H2 selectively through a membrane tube. The steam reforming temperature can be decreased from 800 deg C to 550 deg C, which is a remarkable benefit offered by the non-equilibrium condition. With this new technology, the reactor type can be changed from a High Temperature Gas-cooled Reactor (HTGR) to a Fast Reactor (FR). A Fast Reactor-Membrane Reformer (FR-MR) is composed of a nuclear plant and a hydrogen plant. The nuclear plant is a sodium-cooled Fast Reactor with mixed oxide fuel and with a power of 240 MWt. The heat transport system contains two circuits, the primary circuit and the secondary circuit. The membrane reformer units are set in the secondary circuit. The heat, supplied by the sodium, can produce 200 000 Nm 3 /h by 2 units. There are two types of membranes. One is made of Pd another one (advanced) is made of, for example V, or Nb. The technology for the Pd membrane is already established in a small scale. The non-Pd type is expected to improve the performance. (author)

  18. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

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

    1980-03-01

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  19. The development of fast reactors in France

    International Nuclear Information System (INIS)

    Vautrey, L.

    1982-01-01

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

  20. Status of fast breeder reactor development in Germany

    International Nuclear Information System (INIS)

    Hueber, R.; Kathol, W.; Kempken, M.

    1991-01-01

    The KNK, the sodium cooled compact reactor is an experimental nuclear power plant of 20 MW electric power. Since 1977, it has been operated with fast reactor cores as KNK II. The KNK II/3 core was designed. The core fabrication has been largely completed. In 1990, the KNK II plant achieved a time availability of 56%. On January 8, 1991 KNK II was shut down for inspection. Since pre-nuclear commissioning was completed the Kalkar Nuclear Power Station SNR 300 has been operated in a mode similar to that of a power station. In March 1991 the financing partners decided not to prolong the standby phase because they do not think that the last construction permit and the operation permit will be issued within a definite period of time. The partners were convinced that the lack of progress in the licensing procedure was not caused by basic safety deficiencies of the project but by the way the licensing procedure was executed. The German fast breeder programme is now concentrated on contributions to the European Fast Reactor. (author)

  1. Fast reactor operation in the United States

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  2. Research and development of a super fast reactor (12). Considerations for the reactor characteristics

    International Nuclear Information System (INIS)

    Goto, Shoji; Ishiwatari, Yuki; Oka, Yoshiaki

    2008-01-01

    A research program aimed at developing the Super Fast Reactor (Super FR) has been entrusted by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan since December 2005. It includes the following three projects. (A) Development of the Super Fast Reactor concept. (B)Thermal-hydraulic experiments. (C) Materials development. Tokyo Electric Power Company (TEPCO) has joined this program and works on part (A) together with the University of Tokyo. From the utility's viewpoint, it is important to consider the most desirable characteristics for Super FR to have. Four issues were identified in project (A), (1) Fuel design, (2) Reactor core design, (3) Safety, and (4) Plant characteristics of Super FR. This report describes the desired characteristics of Super FR with respect to item (1) fuel design and item (2) Reactor core design, as compared with a boiling water reactor (BWR) plant. The other two issues will be discussed in this project, and will also be considered in the design process of Super FR. (author)

  3. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

    Parcy, J.P.

    1982-09-01

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

  4. Fuel management codes for fast reactors

    International Nuclear Information System (INIS)

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

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

  5. Substitution models for overlapping technologies - an application to fast reactor deployment

    International Nuclear Information System (INIS)

    Lehtinen, R.; Silvennoinen, P.; Vira, J.

    1982-01-01

    In this paper market penetration models are discussed in the context of interacting technologies. An increased confidence credit is proposed for a technology that can draw on other overlapping technologies. The model is also reduced to a numerically tractable form. As an application, scenarios of fast reactor deployment are derived under different assumptions on the uranium and fast reactor investment costs and by varying model parameters for the penetration of fusion and solar technologies. The market share of fast reactors in electricity generation is expected to lie between zero and 40 per cent in 2050 depending on the market parameters. (orig.) [de

  6. Researches at the University of Tokyo fast neutron sources reactor, YAYOI

    International Nuclear Information System (INIS)

    Koshizuka, S.; Oka, Y.; Saito, I.

    1992-01-01

    The Fast neutron source reactor YAYOI was critical in 1971 at the Nuclear Engineering Research Laboratory, the Faculty of Engineering, the University of Tokyo (UTNL). The core is fueled with the enriched uranium surrounded by the depleted uranium. YAYOI is the first fast reactor in Japan. Many types of studies have been carried out by the researchers of the University of Tokyo in these 20 years. It also contributed to the Japan's national project of developing fast breeder reactors. The reactor is opened to the visiting researchers from universities and research institutes. YAYOI has also been utilized for education of undergraduate and graduate students of the Department of Nuclear Engineering of the University of Tokyo. The present paper briefly summerizes past and present researchers. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Arie, Kazuo; Oomori, Takashi; Okita, Takeshi [Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Kawashima, Masatoshi [Toshiba Nuclear Engineering Services Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Kotake, Shoji [The Japan Atomic Power Company, 1-1, Kanda-Mitoshiro-cho, Chiyoda-ku, Tokyo 101-0053 (Japan); Fuji-ie, Yoichi [Nuclear Salon Fuji-ie, 1-11-10, Yushima, Bunkyo-ku, Tokyo 113-0034 (Japan)

    2013-07-01

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

  8. A review of fast reactor program in Japan. April 2000 - March 2001

    International Nuclear Information System (INIS)

    Nagata, Takashi; Yamashita, Hidetoshi

    2001-01-01

    This report describes the development and activities on fast reactors in Japan thru April 2000 to March 2001. During this period, the most important result of the Japanese Fast Reactor Project was that the first phase 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was completed at the end of March 2001, and the second phase study has just started in order to narrow down the candidate concepts selected in the first phase for next stage. In the Experimental Fast Reactor 'Joyo', the 35 th rated power operation was completed by the end of May 2000. The 13 th periodical inspection and reconstruction works for the Joyo upgrading program (MK-III) were started on the beginning of June 2000. The modification of the cooling system is underway. In the Prototype Fast Breeder Reactor 'Monju', countermeasures against sodium leakage have already been drawn up based on 'Monju' comprehensive safety review. The Japan Atomic Energy Commission (JAEC) has issued a new 'Long-term Program for Research, Development and Utilization of Nuclear Energy' in November 2000. (author)

  9. A review of fast reactor progress in Japan, March 1979

    Energy Technology Data Exchange (ETDEWEB)

    Tomabechi, K

    1979-07-01

    The fast reactor development project in Japan will be continued in the next fiscal year, from April 1979 through March 1980, at a similar scale of effort both in budget and personnel, to those of the fiscal year of 1978. The total budget for LMFBR development for the next fiscal year is approximately 24 billion Yen, 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 reactor development in the PNC is approximately 500, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, approval for power increase from presently approved 50 MWt to 75 MWt with the present core and also to 100 MWt with a modified core in the future was granted by the regulatory authority in September 1978. Two operational cycles at 50 MWt have been completed very recently and preparation for power increase to 75 MWt is being made. With respect to the prototype fast breeder reactor MONJU, progress toward construction is being made and an environmental impact statement of MONJU filed last autumn is being reviewed by the concerned authorities. By the new atomic energy law recently made effective in Japan, the tasks of the former Japan Atomic Energy Commission were split into two and the Atomic Energy Safety Commission was newly established on 4th October 1978 in order to deal with nuclear safety problems in the country. All other problems are treated by the Atomic Energy Commission, as before. Highlights and topics of the fast reactor development activities in the past twelve months are summarized in this paper.

  10. Review of Fast Reactor Activities, March 1980

    International Nuclear Information System (INIS)

    Balz, W.

    1980-01-01

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

  11. Fast breeder reactors: can we learn from experience

    International Nuclear Information System (INIS)

    Keck, O.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-10-01

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

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

    International Nuclear Information System (INIS)

    Hammer, P.

    1980-09-01

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

  14. Integral fast reactor safety features

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  15. Status of fast reactor activities in the USSR

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Ding Xiaoming

    2014-01-01

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

  17. New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

    Energy Technology Data Exchange (ETDEWEB)

    Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2017-07-15

    The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

  18. Fast reactor research in Switzerland

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  19. 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

  20. Fast breeder reactors

    International Nuclear Information System (INIS)

    Ollier, J.L.

    1987-01-01

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

  1. 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

  2. Overview of environmental control aspects for the gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Nolan, A.M.

    1981-05-01

    Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included

  3. Mechatronics of fuel handling mechanism for fast experimental reactor 'Joyo'

    Energy Technology Data Exchange (ETDEWEB)

    Fujiwara, Akikazu (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

    1984-01-01

    The outline of the fast experimental reactor ''Joyo'' is introduced, and the fuel handling mechanism peculiar to fast reactors is described. The objectives of the construction of Joyo are to obtain the techniques for the design, construction, manufacture, installation, operation and maintenance of sodium-cooled fast reactors independently, and to use it as an irradiation facility for the development of fuel and materials for fast breeder reactors. At present, the reactor is operated at 100 MW maximum thermal output for the second objective. Since liquid sodium is used as the coolant, the atmosphere of the fuel handling course changes such as liquid sodium at 250 deg C, argon gas at 200 deg C and water, in addition, the spent fuel taken out has the decay heat of 2.1 kW at maximum. The fuel handling works in the reactor and fuel transfer works, and the fuel handling mechanism of a fuel exchanger and that of a cask car for fuel handling are described. Relay sequence control system is used for the fuel handling mechanism of Joyo.

  4. 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

  5. Sodium fast reactor power monitoring using gamma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  7. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  8. Parameter analysis calculation on characteristics of portable FAST reactor

    International Nuclear Information System (INIS)

    Otsubo, Akira; Kowata, Yasuki

    1998-06-01

    In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)

  9. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

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

    1984-08-01

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

  10. 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

  11. High temperature fast reactor for hydrogen production in Brazil

    International Nuclear Information System (INIS)

    Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F.

    2008-01-01

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

  12. Subcritical molten salt reactor with fast/intermediate spectrum for minor actinides transmutation

    International Nuclear Information System (INIS)

    Degtyarev, Alexey M.; Feinberg, Olga S.; Kolyaskin, Oleg E.; Myasnikov, Andrey A.; Karmanov, Fedor I.; Kuznetsov, Andrey Yu.; Ponomarev, Leonid I.; Seregin, Mikhail B.; Sidorkin, Stanislav F.

    2011-01-01

    The subcritical molten-salt reactor for transmutation of Am and Cm with the fast-intermediate neutron spectrum is suggested. It is shown that ∼10 such reactor-burners is enough to support the future nuclear power based on the fast reactors as well as for the transmutation of Am and Cm accumulated in the spent fuel storages. (author)

  13. The status of fast reactor technology development in China

    International Nuclear Information System (INIS)

    Xu, M.

    2002-01-01

    China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basic strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan (2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before 2010 respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor (CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached 16.8m above the ground. Forty seven components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000m2 floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which is only under consideration up to now. Some important technical selections have been settled, but its design has not yet started. (author)

  14. The Progress of Fast Reactor Technology Development in China

    International Nuclear Information System (INIS)

    Yang, Hongyi; Xu, Mi

    1994-01-01

    China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basis strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan(2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor(CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached the top namely 57m above the ground. More than one hundred components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000m 2 floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which in only under consideration up to now. Some important technical selections have been settled, but its design has not yet started

  15. Fast reactors in Russia: State of the art and trends of development

    International Nuclear Information System (INIS)

    Poplavsky, V.M.; Ashurko, Yu.M.; Zverev, K.V.; Oshkanov, N.N.; Korol'kov, A.S.; Filin, A.I.

    2002-01-01

    This status report contains the following: facts on nuclear power in Russia from 2001-2002; plans for further development of nuclear power; state of the art on operation of fast reactors in 2002, namely BN-600, experimental reactors BOR-60 and BR-10; construction of NPP BN-800; participation in activities on BN-350 reactor decommissioning; description of trends of design studies in the field of fast reactors and accelerator driven systems

  16. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  17. Universal Fast Breeder Reactor Subassembly Counter manual

    International Nuclear Information System (INIS)

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

    1984-08-01

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

  18. Kinetic studies on a repetitively pulsed fast reactor

    International Nuclear Information System (INIS)

    Das, S.

    1982-01-01

    Neutronic analysis of an earlier proposed periodically pulsed fast reactor at Kalpakkam (KPFR) has been carried out numerically under equilibrium and transient conditions using the one-point model of reactor kinetics and the experimentally measured total worth of reactivity modulator, the parabolic coefficient of reactivity of the movable reflector and the mean prompt neutron lifetime. Results of steady-state calculations - treated on the basis of delayed neutron precursor and energy balances during a period of operation - have been compared with the analytical formulae of Larrimore for a parabolic reactivity input. Empirical relations for half-width of the fast neutron pulse, the peak pulse power and the power at first crossing of prompt criticality have been obtained and shown to be accurate enough for predicting steady-state power pulse characteristics of a periodically pulsed fast reactor. The concept of a subprompt-critical reactor has been used to calculate the fictitious delayed neutron fraction, β of the KPFR through a numerical experiment. Relative pulse height stability and pulse shape sensitivity to changes of maximum reactivity is discussed. With the aid of new safety concepts, the Power Amplification Factor (PAF) and the Pulse Growth Factor (Rsub(p)), the dynamics KPFR under accidental conditions has been studied for step and ramp reactivity perturbations. All the analysis has been done without taking account of reactivity feedback. (orig.)

  19. Reactivity monitoring for safety purposes on the UK prototype fast reactor

    International Nuclear Information System (INIS)

    Lord, D.J.; Wilkes, D.J.

    1987-01-01

    The small size and high rating of the liquid metal cooled fast breeder reactor (LMFBR) make the provision of safety related instrumentation for individual subassemblies both difficult and expensive. Global monitoring of the core is thus very attractive. Reactivity monitoring is an important part of such global monitoring. Reactivity monitoring on a short timescale (a few seconds) is used on the UK Prototype Fast Reactor (PFR) as a trip parameter and long-term reactivity monitoring is being developed as a means of providing early warning of slowly developing faults. Results are presented from PFR to demonstrate the capabilities of reactivity monitoring in an operational fast reactor power station. (author)

  20. Liquid metal fast reactor transient design

    International Nuclear Information System (INIS)

    Horak, C.; Purvis, E. III

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  2. Statistical estimation of fast-reactor fuel-element lifetime

    International Nuclear Information System (INIS)

    Proshkin, A.A.; Likhachev, Yu.I.; Tuzov, A.N.; Zabud'ko, L.M.

    1980-01-01

    On the basis of a statistical analysis, the main parameters having a significant influence on the theoretical determination of fuel-element lifetimes in the operation of power fast reactors in steady power conditions are isolated. These include the creep and swelling of the fuel and shell materials, prolonged-plasticity lag, shell-material corrosion, gap contact conductivity, and the strain diagrams of the shell and fuel materials obtained for irradiated materials at the corresponding strain rates. By means of deeper investigation of these properties of the materials, it is possible to increase significantly the reliability of fuel-element lifetime predictions in designing fast reactors and to optimize the structure of fuel elements more correctly. The results of such calculations must obviously be taken into account in the cost-benefit analysis of projected new reactors and in choosing the optimal fuel burnup. 9 refs

  3. Systems design of direct-cycle supercritical-water-cooled fast reactors

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

    1995-01-01

    The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

  4. 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

  5. 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

  6. Integral fast reactor safety features

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  7. 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

  8. 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

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1984-01-01

    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: (a) Full power operation of the PFR. (b) Development work supporting the NNC designed CDFR. (c) Demonstration and development of the fast reactor fuel cycle

  10. Contributions of fast breeder test reactor to the advanced technology in India

    International Nuclear Information System (INIS)

    Kapoor, R.P.

    2001-01-01

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe loop type, sodium cooled, plutonium rich mixed carbide fuelled reactor. Its operation at Indira Gandhi Centre for Atomic Research, since first criticality in 1985, has contributed immensely to the advancement of this multidisciplinary and complex fast breeder technology in the country. It has also given a valuable operational feedback for the design of 500 MWe Prototype Fast Breeder Reactor. This paper highlights FBTR's significant contributions to this important technology which has a potential to provide energy security to the country in future. (author)

  11. Design codes for fast reactor steam generators

    International Nuclear Information System (INIS)

    Townley, C.H.A.

    1978-01-01

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

  12. 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.)

  13. Risk-assessment methodology for fast breeder reactors

    International Nuclear Information System (INIS)

    Ott, K.O.

    1976-04-01

    The methods applied or proposed for risk assessment of nuclear reactors are reviewed, particularly with respect to their applicability for risk assessment of future commercial fast breeder reactors. All methods are based on the calculation of accident consequences for relatively few accident scenarios. The role and general impact of uncertainties in fast-reactor accident analysis are discussed. The discussion shows the need for improvement of the methodology. A generalized and improved risk-assessment methodology is outlined and proposed (accident-spectra-progression approach). The generalization consists primarily of an explicit treatment of uncertainties throughout the accident progression. The results of this method are obtained in form of consequence distributions. The width and shape of the distributions depend in part on the superposition of the uncertainties. The first moment of the consequence distribution gives an improved prediction of the ''average'' consequence. The higher-consequence moments can be used for consideration of risk aversion. The assessment of the risk of one or a certain number of nuclear reactors can only provide an ''isolated'' risk assessment. The general problem of safety risk assessment and its relation to public acceptance of certain modes of power production is a much broader problem area, which is also discussed

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

    International Nuclear Information System (INIS)

    Suzuki, Tomoo

    1971-01-01

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

  15. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-07

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

  16. Thermal and fast reactor benchmark testing of ENDF/B-6.4

    International Nuclear Information System (INIS)

    Liu Guisheng

    1999-01-01

    The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

  17. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: • To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; • To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; • To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; • To discuss the results of studies and ongoing R&D activities that address cost reduction and the future economic competitiveness of fast reactors; • To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  18. Universal Fast Breeder Reactor Subassembly Counter manual

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-08-01

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

  19. The plutonium utilization in thermal and fast reactor in Japan

    International Nuclear Information System (INIS)

    Amanuma, T.; Uematsu, K.

    1977-01-01

    The nuclear power development in Japan is rather extensive one, and the installed nuclear power capacity is expected to reach 49,000 MWe by 1985 and possibly to reach 170,000 MWe by 2000 according to a prediction. Currently istalled nuclear power is mainly based on Light Water type Reactor, and this trend is expected to persist for the time-being. The plutonium produced by LWR will be accumulated to 20 tons by 1985 and to more than 200 tons by 2000. If the produced plutonium will simply be stored, it will raise the economic pressure to utilities and the management and physical protection problems associated with plutonium storing. Therefore, it is not too wise simply to store plutonium in a locked vault. In Japan, there are three ways of solving these problems which are currently worked out. There is no doubt that the best solution is to use plutonium in fast reactors. To reach this goal, an Experimental Fast Reactor ''JOYO''has been constructed and it is waiting for criticality in very near future. A prototype fast breeder reactor ''MONJU'', which is designed for about 300 MWe, is nearing to the last stage of the design work. The start of its construction will take place in a few yesars. The domonstration fast breeder reactor will come next to ''MONJU'' and the large scale commercial use of fast breeder reactor is expected to start around 1995. To anwer the near-term need for plutonium utilization, two technologies, which are equally important to Japan, are currently developed. One is the recycle use of plutonium into LWR. This technology has long been jointly developed by research organizations and utilities. Some of fuel irradiation data are already obtained and the physics study has also been extensive. The application of this technology is expected to start about 1987. The other is to burn plutonium in an Advanced Thermal Reactor (D 2 O moderated, Boiling Water Cooled) which shows better characteristics of using plutonium. The 160 MWe ''Fugen'' is a prototype

  20. 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)

  1. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

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

  2. Ageing management practice in Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Srinivasan, G.; Ramanathan, V.; Swaminathan, P.R.; Babu, A.; Rajasekarappa, E.; Rajendran, B.; Ramalingam, P.V.

    2006-01-01

    Fast Breeder Test Reactor is a 40 MWt, sodium cooled, PuC-UC fuelled fast reactor, located at Kalpakkam, India. The reactor went critical in October 85 with Mark I core rated for 10.5 MWt at a peak LHR of 320 W/cm. The reactor core was progressively enlarged and TG was synchronized to the grid in July 97. The present core has 41 fuel subassemblies rated for 15.7 MWt at a peak LHR of 320 W/cm. The reactor has so far been operated for 33000 h and has seen 660 EFPD of operation corresponding to peak LHR of 320 W/cm. The peak burnup reached by the carbide fuel is 127 GWd/t, without any fuel clad failure. The four sodium pumps have been operating satisfactorily for a cumulative time of more than 5,00,000 h. Creep, fatigue and fluence govern the life of the nuclear systems. Because of the reduced power and temperature at which the reactor has so far been operated, there is little ageing of the nuclear systems. The life of the nuclear components is being monitored by periodic surveillance. Periodic assessment of the fluence seen by reactor components is being made. The conventional systems have been in service for the past 19 years. Civil structures are 25 years old. These have been maintained by periodic preventive maintenance and replacement / repair wherever required. This paper details the various ageing management practices in FBTR. (author)

  3. Instrumentation and control of future sodium cooled fast reactors - Design improvements

    International Nuclear Information System (INIS)

    Madhusoodanan, K.; Sakthivel, M.; Chellapandi, P.

    2013-06-01

    India's fast reactor program started with the 40 MWt Fast Breeder Test Reactor. 500 MWe Prototype Fast Breeder Reactor (PFBR) is currently under construction at Kalpakkam. Safety of PFBR is enhanced by improved design features of I and C system. Since the design of Instrumentation and control (I and C) of PFBR, considerable improvements in terms of advancement in technology and indigenization has taken place. Further improvements in I and C is proposed for solving many of the difficulties faced during the design and construction phases of PFBR. Design improvements proposed are covered in this paper which will make the implementation and maintenance of I and C of future SFRs easier. (authors)

  4. The influence of fast reactor emergency conditions upon fuel element performance

    International Nuclear Information System (INIS)

    Bagdasarov, Yu.E.; Buksha, Yu.K.; Zabudko, L.M.; Likhachev, Yu.I.

    1985-01-01

    Fuel-pin cladding is one of the most important protective barriers preventing the release and propagation of radioactive contamination. By now the calculated determination of fast-reactor fuel-element performance under stationary conditions has been considered in detail but the investigation of the influence of emergency conditions has been given less attention. Under emergency conditions of the fast reactor operation there arise short-duration excesses of rated parameters (temperature, energy release, etc.) which are confined within tolerable limits with the use of the safety system. Some features of the sodium-cooled fast reactors (small mean prompt-neutron lifetime, relatively weak reactivity feedback, etc.) complicate the work of safety systems. Therefore, the tolerable deviations of parameters should be carefully validated

  5. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Hyoudou, Hideaki; Tatsumi, Masahiro

    2010-02-01

    Improvement on prediction accuracy for neutronics property of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified constants library as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores, however, improvement of not only static properties but also burnup properties is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup properties using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous research, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for production systems. In the present study, we implemented functions for cell calculations and burnup calculations. With this, whole steps in analysis can be carried out with only this system. In addition, we modified the specification of user input to improve the convenience of this system. Since implementations being done so

  6. 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)

  7. 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)

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

    International Nuclear Information System (INIS)

    Beddingfield, David H.; Lafleur, Adrienne M.

    2009-01-01

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

  9. 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.)

  10. Fabrication of Fast Reactor Fuel Pins for Test Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Karsten, G. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Dippel, T. [Institute for Radiochemistry, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Laue, H. J. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

    1967-09-15

    An extended irradiation programme is being carried out for the fuel element development of the Karlsruhe fast breeder project. A very important task within the programme is the testing of plutonium-containing fuel pins in a fast-reactor environment. This paper deals with fabrication of such pins by our laboratories at Karlsruhe. For the fast reactor test positions at present envisaged a fuel with 15% plutonium and the uranium fully enriched is appropriate. Hie mixed oxide is both pelletized and vibro-compacted with smeared densities between 80 and 88% theoretical. The pin design is, for example, such that there are two gas plena at the top and bottom, and one blanket above the fuel with the fuel zone fitting to the test reactor core length. The specifications both for fuel and cladding have been adapted to the special purpose of a fast-breeder reactor - the outer dimensions, the choice of cladding and fuel types, the data used and the kind of tests outline the targets of the development. The fuel fabrication is described in detail, and also the powder line used for vibro-compaction. The source materials for the fuel are oxalate PuO{sub 2} and UO{sub 2} from the UF{sub 6} process. The special problems of mechanical mixing and of plutonium homogeneity have been studied. The development of the sintering technique and grain characteristics for vibratory compactive fuel had to overcome serious problems in order to reach 82-83% theoretical. The performance of the pin fabrication needed a major effort in welding, manufacturing of fits and decontamination of the pin surfaces. This was a stimulation for the development of some very subtle control techniques, for example taking clear X-ray photographs and the tube testing. In general the selection of tests was a special task of the production routine. In conclusion the fabrication of the pins resulted in valuable experiences for the further development of fast reactor fuel elements. (author)

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

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

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

  12. Effect of neutron anisotropic scattering in fast reactor analysis

    International Nuclear Information System (INIS)

    Chiba, Gou

    2004-01-01

    Numerical tests were performed about an effect of a neutron anisotropic scattering on criticality in the Sn transport calculation. The simplest approximation, the consistent P approximation and the extended transport approximation were compared with each other in one-dimensional slab fast reactor models. JAERI fast set which has been used for fast reactor analyses is inadequate to evaluate the effect because it doesn't include the scattering matrices and the self-shielding factors to calculate the group-averaged cross sections weighted by the higher-order moment of angular flux. In the present study, the sub-group method was used to evaluate the group-averaged cross sections. Results showed that the simplest approximation is inadequate and the transport approximation is effective for evaluating the anisotropic scattering. (author)

  13. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Yun, Sunghwan; Kim, Sang Ji

    2015-01-01

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

  14. Assessment of residual life of fast breeder test reactor

    International Nuclear Information System (INIS)

    Srinivasan, G.

    2016-01-01

    The Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor and has been in operation since 1985. As a part of regulatory requirement for relicensing, residual life assessment had to be carried out. The systems are made of SS 316, and designed for creep and fatigue. The design life for creep is 100,000 h at 550°C. The design fatigue cycle for operation from shutdown to full power varies from component to component. In general, most of the components are designed for 2000 cycles. The reactor has operated mostly below the design temperatures. It is seen that enough creep-fatigue life is available for the non-replaceable, permanent components. The residual life was found to be governed by the residual ductility of the Grid Plate supporting the core after neutron irradiation. Fast flux measurements were carried out at the grid plate location. Samples were irradiated and tensile tested. Results indicate the allowable dpa for a 10% residual ductility criterion as 4.37. This gave a residual life of ~ 6 Effective Full Power Years for the reactor as of Feb 2012. Measures to reduce the neutron dose on the grid plate are being taken. (author)

  15. Fuel, structural material and coolant for an advanced fast micro-reactor

    International Nuclear Information System (INIS)

    Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

    2011-01-01

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

  16. Experimental Methods Related to Coupled Fast-Thermal Systems at the RB Reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2002-01-01

    In addition to the review of RB reactor characteristics this presentation is focused on the coupled fast-thermal systems achieved at the reactor. The following experimental methods are presented: neutron spectra measurements; steady state experiments and kinetic measurements ( β eff ) related to the coupled fast-thermal cores

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

    Science.gov (United States)

    Qvist, Staffan Alexander

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

  18. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Johnson, R.N.

    1984-04-01

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

  19. 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

  20. 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)

  1. 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)

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

    International Nuclear Information System (INIS)

    Righini, R.

    1988-01-01

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

  3. Vessel core seismic interaction for a fast reactor

    International Nuclear Information System (INIS)

    Martelli, A.; Maresca, G.

    1984-01-01

    This report deals with the analysis carried out in collaboration between ENEA and NIRA for optimizing the iterative procedure applied for the evaluation of the effects of the vessel core dynamic interaction for a fast reactor in the case of a earthquake. In fact, as shown in a previous report the convergence of such procedure was very slow for the design solution adopted for the PEC reactor, i.e. with a core restraint plate located close to the top of the core elements. This study, although performed making use of preliminary data (the same of the cited previous report) demonstrates that the convergence is fast if a suitable linear core model is applied in the first iteration linear calculations carried out by NIRA, with an intermediate stiffness with respect to those corresponding to the two limit models previously assumed and increased damping coefficients. Thus, the optimized iterative procedures is now applied in the PEC reactor block seismic verification analysis

  4. Management of European fast reactor R and D

    International Nuclear Information System (INIS)

    Judd, A.M.; Sheriff, N.

    1993-01-01

    Since 1984 government-funded fast reactor R and D in France, Germany and the UK has been run as a collaborative activity, and since 1988 as a unified programme in support of the design and construction of the advanced European Fast Reactor. This paper describes the international management structure which has been set up, and the means used to control the work. It is written from the point of view of those engaged in the project, and makes no attempt at a formal analysis of the structure. The main difficulty is that control of funding remains with the three governments. The R and D programme has to be managed so that it meets the needs of each government separately as well as the designers' requirements. To start with the management structure was excessively bureaucratic, but it has become more flexible and efficient. This has happened as the initial nationalistic suspicions have broken down, and the staff engaged in the work have learnt more about each others' ways of working so that an atmosphere of trust and inter-dependence has grown up. (This paper was written before the changes in UK policy on fast reactor development were announced in November 1992). (Author)

  5. Development of fast reactor metal fuels containing minor actinides

    International Nuclear Information System (INIS)

    Ohta, Hirokazu; Ogata, Takanari; Kurata, Masaki; Koyama, Tadafumi; Papaioannou, Dimitrios; Glatz, Jean-Paul; Rondinella, Vincenzo V.

    2011-01-01

    Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU. (author)

  6. Benchmark tests of JENDL-3.2 for thermal and fast reactors

    International Nuclear Information System (INIS)

    Takano, Hideki

    1995-01-01

    Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

  7. 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)

  8. IAEA fast reactor knowledge preservation initiative. Project focus: KNK-II reactor, Karlsruhe, Germany

    International Nuclear Information System (INIS)

    2004-08-01

    This Working Material (including the attached CD-ROM) documents progress made in the IAEA's initiative to preserve knowledge in the fast reactor domain. The brochure describes briefly the context of the initiative and gives an introduction to the contents of the CD-ROM. In 2003/2004 a first focus of activity was concentrated on the preservation of knowledge related to the KNK-II experimental fast reactor in Karlsruhe, Germany. The urgency of this project was given by the impending physical destruction of the installation, including the office buildings. Important KNK-II documentation was brought to safety and preserved just in time. The CD-ROM contains the full texts of 264 technical and scientific documents describing research, development and operating experience gained with the KNK-II installation over a period of time from 1965 to 2002, extending through initial investigations, 17 years of rich operating experience, and final shutdown and decommissioning. The index to the documents on the CD-ROM is printed at the end of this booklet in chronological order and is accessible on the CD by subject index and chronological index. The CD-ROM contains in its root directory also the document 'fr c lassification.pdf' which describes the classification system used for the present collection of documents on the fast reactor KNK-II

  9. High-burn-up fuels for fast reactors. Past experience and novel applications

    International Nuclear Information System (INIS)

    Weaver, Kevan D.; Gilleland, John; Whitmer, Charles; Zimmerman, George

    2009-01-01

    Fast reactors in the U.S. routinely achieved fuel burn-ups of 10%, with some fuel able to reach peak burn-ups of 20%, notably in the Experimental Breeder Reactor II and the Fast Flux Test Facility. Maximum burn-up has historically been constrained by chemical and mechanical interactions between the fuel and its cladding, and to some extent by radiation damage and thermal effects (e.g., radiation-induced creep, thermal creep, and radiation embrittlement) that cause the cladding to weaken. Although fast reactors have used several kinds of fuel - including oxide, metal alloy, carbide, and nitride - the vast majority of experience with fast reactors has been using oxide (including mixed oxide) and metal-alloy fuels based on uranium. Our understanding of high-burn-up operation is also limited by the fact that breeder reactor programs have historically assumed that their fuel would eventually undergo reprocessing; the programs thus have not made high burn-up a top priority. Recently a set of novel designs have emerged for fast reactors that require little initial enrichment and no reprocessing. These reactors exploit a concept known as a traveling wave (sometimes referred to as a breed-and-burn wave, fission wave, or nuclear-burning wave). By breeding and using its own fuel in place as it operates, a traveling-wave reactor can obtain burn-ups that approach 50%, well beyond the current base of knowledge and experience. Our computational work on the physics of traveling-wave reactors shows that they require metal-alloy fuel to provide the margins of reactivity necessary to sustain a breed-and-burn wave. This paper reviews operating experience with high-burn-up fuels and the technical feasibility of moving to a qualitatively new burn-up regime. We discuss our calculations on traveling-wave reactors, including those concerning the possible use of thorium. The challenges associated with high burn-up and fluence in fuels and materials are also discussed. (author)

  10. Fast-acting nuclear reactor control device

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  11. Research and development of super light water reactors and super fast reactors in Japan

    International Nuclear Information System (INIS)

    Oka, Y.; Morooka, S.; Yamakawa, M.; Ishiwatari, Y.; Ikejiri, S.; Katsumura, Y.; Muroya, Y.; Terai, T.; Sasaki, K.; Mori, H.; Hamamoto, Y.; Okumura, K.; Kugo, T.; Nakatsuka, T.; Ezato, K.; Akasaka, N.; Hotta, A.

    2011-01-01

    Super Light Water Reactors (Super LWR) and Super Fast Reactors (Super FR) are the supercritical- pressure light water cooled reactors (SCWR) that are developed by the research group of University of Tokyo since 1989 and now jointly under development with the researchers of Waseda University, University of Tokyo and other organizations in Japan. The principle of the reactor concept development, the results of the past Super LWR and Super FR R&D as well as the R&D program of the Super FR second phase project are described. (author)

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

    International Nuclear Information System (INIS)

    Gregg, R.; Hesketh, K.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    1987-07-01

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

  14. Utilization of fast reactor excess neutrons for burning long-lived fission products

    International Nuclear Information System (INIS)

    Kawashima, K.; Kobayashi, K.; Kaneto, K.

    1995-01-01

    An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc-99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction. The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs. (author)

  15. BN-800 as a new stage in development of fast neutron sodium cooled reactors

    International Nuclear Information System (INIS)

    Poplavskij, V.M.; Chebeskov, A.N.; Matveev, V.I.

    2004-01-01

    The role of fast reactors in the strategy of evolution of the nuclear power of Russia is discussed, BN-800 under construction, where unique technical and construction decisions are used, is viewed. Economical estimations of expenses with regard for all life cycle demonstrate that fast reactors may be no higher-priced than the most popular in the world water moderated reactors. Closing of nuclear fuel cycle of BN-800 makes possible decision of the problem of plutonium and actinide utilization, that makes the fast reactor more safety for the environment [ru

  16. Active acoustic leak detection in steam generator units of fast reactors

    International Nuclear Information System (INIS)

    Oriol, L.; Journeau, Ch.

    1996-01-01

    Steam generators (SG) of Fast Reactors can be subject to water leakage into the sodium secondary circuit, causing an exothermic chemical reaction with potential serious damage to plant. Within the framework of the European Fast Reactor project, the CEA has developed an active acoustic detection technique which, when used in parallel with passive acoustic detection, will lead to effective leak detection results in terms of reliability and false alarm rates. Whilst the passive method is based on the increase in acoustic noise generated by the reaction, the active method takes advantage of the acoustic attenuation by the hydrogen bubbles produced. The method has been validated: in water, during laboratory testing at the Centre d'Etudes de Cadarache; in sodium, at the ASB loop at Bensberg (Germany) and at AEA Dounreay (Scotland). Full analysis of the tests carried out on the SG of the Prototype Fast Reactor in 1994 during end-of-life testing should lead to reactor validation on the method. (authors)

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

    International Nuclear Information System (INIS)

    Merk, B.; Kliem, S.

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  19. Small liquid metal reactor for an initial phase of fast breeder reactor introduction

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Nascimento, J.A. do.

    1985-01-01

    Safety and burnup characteristics of a 1000 MWth liquid metal reactor have been examined for various fuel types. With metallic Pu/Th fuel containing a small amount of zirconium hydride, low sodium-void reactivity, a high Doppler coefficient, and small burnup reactivity swings can be achieved. A conservative design is considered for an initial phase of fast breeder reactor development and possible modifications are discussed. (Author) [pt

  20. IAEA Technical Meeting on Status of IAEA Fast Reactor Knowledge Preservation Initiative. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    In response to needs expressed by Member States and within a broader IAEA-wide effort in nuclear knowledge preservation, the IAEA has been carrying out a dedicated initiative on Fast Reactor Data Knowledge Preservation (FRKP). The main objectives of the FRKP initiative are to: • Halt the on-going loss of information related to Fast Reactors (FR); • Collect, retrieve, preserve and make accessible already existing data and information on FR. These objectives require the implementation of activities supporting digital document archival, exchange, search and retrieval and facilitating, by developing and using suitable standards and IT tools, the knowledge preservation over the next decades. To this purpose the IAEA has developed the Fast Reactor Knowledge Organization System (FRKOS), a web-based application employing IAEA methodology and approach for categorization of FR knowledge domain, which allows creating a comprehensive and well-structured international inventory of fast reactor data and information provided by different Member States. The resulting Web Portal is established and maintained by the IAEA. The IAEA knowledge preservation initiatives and tools in the field of fast neutron systems - which were presented and very well received during the recent IAEA Fast Reactor and Related Fuel Cycles Conference (FR13) - are supposed to be of interest for national nuclear authorities, regulators, scientific and research organizations, commercial companies and all other stakeholders involved in fast reactor activities at national or international level. The objectives of the technical meeting were to: • Exchange information between the member states/international organizations on national and international initiatives addressing knowledge preservation and data retrieval/collection in the field of fast neutron systems; • Present and discuss the member states’/international organizations’ policies and conditions for releasing to the IAEA both publicly

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

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: - To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; - To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; - To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; - To discuss the results of studies and on-going R&D activities that address cost reduction and the future economic competitiveness of fast reactors; and - To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    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 desirably essential to have this technology available for introduction. The recycling of plutonium into LMFRs would allow 'burning' of the associated extremely long-life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. This additional important mission for the LMFR is gaining worldwide interest. In the framework of disarmament of nuclear weapons and the utilization of the nuclear material or peaceful purposes a role for fast reactors can be also considered. Over the past 29 years, the IAEA has actively encouraged and advocated international co-operation in Fast Breeder Reactor Technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1995, as reported at the 29th Annual

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

    International Nuclear Information System (INIS)

    1996-01-01

    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 desirably essential to have this technology available for introduction. The recycling of plutonium into LMFRs would allow 'burning' of the associated extremely long-life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. This additional important mission for the LMFR is gaining worldwide interest. In the framework of disarmament of nuclear weapons and the utilization of the nuclear material or peaceful purposes a role for fast reactors can be also considered. Over the past 29 years, the IAEA has actively encouraged and advocated international co-operation in Fast Breeder Reactor Technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1995, as reported at the 29th Annual

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

    International Nuclear Information System (INIS)

    2005-01-01

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

  5. Past and present role of fast breeder reactors in Italy

    International Nuclear Information System (INIS)

    Castelli, G.; Cicognani, G.; Ghilardotti, G.; Musso, B.

    1978-01-01

    The paper describes the programme that is under development in Italy for fast breeder reactors. The Italian engagement in the construction of the Creys-Malville plant is discussed as well as the work at present under way in R and D activities and on the PEC reactor. The roles of the different organizations involved in the fast breeder reactor programme are also considered with particular attention to the activities carried out by AGIP NUCLEARE, CNEN, ENEL, NIRA and by the manufacturing companies. An overall picture of the different agreements between European countries is also given with reference to the construction of the Creys-Malville plant and to research and industrial development activities. (author)

  6. Application of synthesis methods to two-dimensional fast reactor transient study

    International Nuclear Information System (INIS)

    Izutsu, Sadayuki; Hirakawa, Naohiro

    1978-01-01

    Space time synthesis and time synthesis codes were developed and applied to the space-dependent kinetics benchmark problem of a two-dimensional fast reactor model, and it was found both methods are accurate and economical for the fast reactor kinetics study. Comparison between the space time synthesis and the time synthesis was made. Also, in space time synthesis, the influence of the number of trial functions on the error and on the computing time and the effect of degeneration of expansion coefficients are investigated. The matrix factorization method is applied to the inversion of the matrix equation derived from the synthesis equation, and it is indicated that by the use of this scheme space-dependent kinetics problem of a fast reactor can be solved efficiently by space time synthesis. (auth.)

  7. Methods for reactor physics calculations for control rods in fast reactors

    International Nuclear Information System (INIS)

    Grimstone, M.J.; Rowlands, J.L.

    1988-12-01

    The IAEA Specialists' Meeting on ''Methods for Reactor Physics Calculations for Control Rods in Fast Reactors'' was held in Winfrith, United Kingdom, on 6-8 December, 1988. The meeting was attended by 23 participants from nine countries. The purpose of the meeting was to review the current calculational methods and their accuracy as assessed by theoretical studies and comparisons with measurements, and then to identify the requirements for improved methods or additional studies and comparisons. The control rod properties or effects to be considered were their reactivity worths, their effect on the power distribution through the core, and the reaction rates and energy deposition both within and adjacent to the rods. The meeting was divided into five sessions, in the first of which each national delegation presented a brief overview of their programme of work on calculational methods for fast reactor control rods. In the next three sessions a total of seventeen papers were presented describing calculational methods and assessments of their accuracy. The final session was a discussion to draw conclusions regarding the current status of methods and the further developments and validation work required. A separate abstract was prepared for each of the 23 papers presented at the meeting. Refs, figs and tabs

  8. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-01-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

  9. Neutronics methods for transient and safety analysis of fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marchetti, Marco

    2017-07-01

    Modeling the evolution of possible or postulated accidents in nuclear reactors is fundamental in designing safe systems. For the next generation of reactors, in particular fast reactors, fuel movement during an accident can, in principle, drive an energetic event. Such is the issue of recriticality. The thermal energy produced during these events will, possibly, be converted into mechanical energy by some mechanisms. For example, the nuclear heat deposited in the fuel could cause fuel vaporization and its subsequent expansion. This movement would accelerate the surrounding sodium: part of the initial energy in the fuel is thus converted into sodium kinetic energy. This mechanical energy will finally be absorbed, in some way or another, by the reactor vessel. Providing an accurate estimate for the maximum mechanical work that any accidental sequence can do onto the reactor vessel is an essential step in designing a reactor containment that would withstand any load generated by any accident. That would assure accident containment, without consequences for the general public. Fast reactor accident modeling is a complicated task. The outcome of an accident is determined by different physical phenomena, all acting at almost the same time. Safety analysts must track all these different phenomena. Multi-physics codes have been developed for this task. They must contain accurate models for fluid-dynamics, neutronics, and structures. This work has to do with neutronics modeling of such accidents. Past and recent analyses have been limited to the approximate description of the neutronic field, for example by using a rough description of the energy and/or of the angular dependence of the neutron flux. In this work, different neutronic solvers are selected and coupled into a general multi-physics code for fast reactor accident analysis. Performances of each of them is then assessed. Some emphasis has been put also in assessing the speed of these solvers for determining the

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1981-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-05-01

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

  12. Fast reactors and nonproliferation

    International Nuclear Information System (INIS)

    Orlov, V.V.

    1997-01-01

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

  13. Thermal baffle for fast-breeder reactor

    International Nuclear Information System (INIS)

    Rylatt, J.A.

    1977-01-01

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

  14. A prospect of fast reactor and related fuel cycle in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Nagata, Takashi [Japan Atomic Energy Agency, Ibaraki (Japan)

    2009-04-15

    JAEA has launched a new project 'Fast Reactor Cycle Technology Development'(FaCT) in cooperation with electric utilities. In this FaCT project, a combination of 'the Japanese sodium cooled loop type fast reactor with oxide fuel, the advanced aqueous reprocessing, and the simplified palletizing fuel fabrication systems' is adopted, where many innovative technologies with technical challenging issues are actively used in order to provide significant improvements in economic competitiveness, and enhancement of safety and reliability, sustainability, and nonproliferation. Fast reactor cycle technology will provide harmonic solutions for global issues of energy resources and environments, and is expected to contribute to sustainable development of the future society. Therefore, it was selected as one of key technologies of national importance in the third term (JPY2006-2010) 'Science and Technology Basic Plan' in March 2006 in Japan. The 'Nuclear Energy National Plan' in August 2006 states start up of a demonstration FR by around 2025 and deployment of a commercial FR before 2050, and start operating fuel cycle facilities when these reactors achieve consistency. Accordingly, we will decide about the adoption of innovative technologies by judging their applicability by 2010, and present the conceptual designs of commercial and demonstration FR cycle facilities by 2015 with the R and D plans to realize. In developing the FR cycle, 5 Party council, which consists of MEXt, MITI, electricity utilities, manufacturers, and JAEA, was established in July 2006 for moving forward on the commercialization smoothly. In this framework, users' requirements for the future R and D, a scenario of transition from light water reactor cycle to sodium cooled FR cycle, international collaboration, development schedule, demonstration steps, and so on are discussed. In this presentation, a prospect concerning the system design features of JSFR and a

  15. A prospect of fast reactor and related fuel cycle in Japan

    International Nuclear Information System (INIS)

    Nagata, Takashi

    2009-01-01

    JAEA has launched a new project 'Fast Reactor Cycle Technology Development'(FaCT) in cooperation with electric utilities. In this FaCT project, a combination of 'the Japanese sodium cooled loop type fast reactor with oxide fuel, the advanced aqueous reprocessing, and the simplified palletizing fuel fabrication systems' is adopted, where many innovative technologies with technical challenging issues are actively used in order to provide significant improvements in economic competitiveness, and enhancement of safety and reliability, sustainability, and nonproliferation. Fast reactor cycle technology will provide harmonic solutions for global issues of energy resources and environments, and is expected to contribute to sustainable development of the future society. Therefore, it was selected as one of key technologies of national importance in the third term (JPY2006-2010) 'Science and Technology Basic Plan' in March 2006 in Japan. The 'Nuclear Energy National Plan' in August 2006 states start up of a demonstration FR by around 2025 and deployment of a commercial FR before 2050, and start operating fuel cycle facilities when these reactors achieve consistency. Accordingly, we will decide about the adoption of innovative technologies by judging their applicability by 2010, and present the conceptual designs of commercial and demonstration FR cycle facilities by 2015 with the R and D plans to realize. In developing the FR cycle, 5 Party council, which consists of MEXt, MITI, electricity utilities, manufacturers, and JAEA, was established in July 2006 for moving forward on the commercialization smoothly. In this framework, users' requirements for the future R and D, a scenario of transition from light water reactor cycle to sodium cooled FR cycle, international collaboration, development schedule, demonstration steps, and so on are discussed. In this presentation, a prospect concerning the system design features of JSFR and a summary of the above R and D progresses for

  16. Thermophysical properties of fast reactor fuel

    International Nuclear Information System (INIS)

    Fink, J.K.

    1984-01-01

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

  17. 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

  18. 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)

  19. The behaviour of materials in fast reactors

    International Nuclear Information System (INIS)

    Matthews, J.R.

    1977-01-01

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

  20. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-11-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

  1. Commission of the European Communities - Review of fast reactor activities, March 1979

    International Nuclear Information System (INIS)

    Balz, W.

    1979-01-01

    The Commission of the European Communities pursued its activities on the lines described earlier. Besides the execution of a research programme in its Joint Research Centre the Commission endeavoured to support the development and demonstration of fast reactors. Most of the latter activities were performed in the frame of the Fast Reactor Coordinating Committee

  2. Tests for validation of fast neutron reactors safety

    International Nuclear Information System (INIS)

    Nagata, T.; Yamashita, H.

    2001-01-01

    Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

  3. Scenario for commercialization of fast breeder reactors

    International Nuclear Information System (INIS)

    Kumaoka, Yoshio; Sato, Morihiko

    1989-01-01

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

  4. Development of a fresh plutonium fuel container for a prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Ohtake, T.; Takahashi, S.; Mishima, T.; Kurakami, J.; Yamamoto, Y.; Ohuchi, Y.

    1989-01-01

    Japan gives a good deal of encouragement to development of a fast breeder reactor (which is considered as the most likely candidate for nuclear power generation) to secure long-term energy source. And, following an experimental fast breeder reactor Joyo, a prototype fast breeder reactor Monju is now under vigorous construction. Related to development of the prototype fast breeder reactor, it is necessary and important to develop transport container which is used for transporting fresh fuel assemblies from Plutonium Fuel Production Facility to the Monju power plant. Therefore, the container is now being developed by Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, shipment and vibration tests, handling performance tests, shielding performance tests and prototype container tests are executed with prototype containers fabricated according to a final design, in order to experimentally confirm soundness of transport container and its contents, and propriety of design technique. This paper describes the summary of general specifications and structures of this container and the summary of preliminary safety analysis of package

  5. Invitation to the IAEA FR17 Conference on Fast Reactors, Russia, 2017

    International Nuclear Information System (INIS)

    Rachkov, V.

    2013-01-01

    Russian Federation has a large successful experience of peaceful application of nuclear power and it is a leader in development of fast reactors that permit to create sustainable and safe base for future world nuclear power. Russian Government approved in 2010 a Federal Target Program “Nuclear power technologies of a new generation for period of 2010-2015 and with outlook to 2020” aimed at creation of new technological platform for future nuclear power based on fast reactors. State Atomic Energy Corporation “Rosatom” supports activities of the IAEA on strengthening international collaboration on fast reactors and invites IAEA to host next FR17 Conference on FRs in Russia in 2017

  6. Status of sodium cooled fast reactors with closed fuel cycle in India

    International Nuclear Information System (INIS)

    Raj, B.

    2007-01-01

    Fast reactors form the second stage of India's 3-stage nuclear power programme. The seed for India's fast reactor programme was sown through the construction of the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam, that was commissioned in 1985. FBTR has operated with an unique, indigenously developed plutonium rich mixed carbide fuel, which has reached a burn up as high as 155 GWd/t without any fuel failure in the core. The sodium systems in the reactor have performed excellently. The availability of the reactor has been as high as 92% in the recent campaigns. The fuel discharged from FBTR up to 100 GWd/t has been reprocessed successfully. The experience gained in the construction, commissioning and operation of FBTR has provided the necessary confidence to launch a Prototype FBR of 500 MWe capacity (PFBR). This reactor will be fuelled by uranium, plutonium mixed oxide. The reactor construction started in 2003 and the reactor is scheduled to be commissioned by 2010. The design of the reactor has incorporated the worldwide operating experience from the FBRs and has addressed various safety issues reported in literature, besides introducing a number of innovative features which have reduced the unit energy cost and contributed to its enhanced safety. Simultaneous with the construction of the reactor, the fuel cycle of the reactor has been addressed in a comprehensive manner and construction of a fuel cycle facility has been initiated. Subsequent to the PFBR, 4 more reactors with identical design are proposed to be constructed. Various elements of reactor design are being carefully analysed with the aim of introducing innovative features towards further reduction in unit energy cost and enhancing safety in these reactors

  7. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  8. Charging machine for a fast production reactor

    International Nuclear Information System (INIS)

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

    1971-01-01

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

  9. The feasibility study on commercialized fast reactor cycle system

    International Nuclear Information System (INIS)

    Noda, Hiroshi

    2002-01-01

    The feasibility study on commercialized Fast Reactor cycle system (FS) has been carried out by a joint team with the participation of all parties concerned in Japan since July, 1999. It aims to clarify various perspectives for commercialized fast reactor cycle system and also suggest development strategies to diverse social needs in the 21 st century. The FS consists of several phases. The phase 1 has progressed as planned and the highly feasible candidate concepts with innovative technologies have been screened out among a wide variety of concepts. During the phase 2, approximately five years after the phase 1, the in-depth design studies and engineering scale tests of key technologies are being conducted to verify and validate the feasibility of screened candidate concepts. At the end of the phase 2, a few promising concepts will be selected with their R and D tasks. The paper describes the results of the phase 1, the activities of the phase 2 and the new concept related to the fast reactor fuel cycle focusing on the reduction in environmental burden, which is one of key factors to sustain the nuclear power generation in the 21 st century

  10. Evaluation of transmutation performance of long-lived fission products with a super fast reactor

    International Nuclear Information System (INIS)

    Lu, Haoliang; Han, Chiyoung; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    The performance of the Super Fast Reactor for transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super Fast Reactor. First is in the blanket assembly due to the ZrH 1.7 layer which can slow down the fast neutrons. Second is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected of transmutation analysis include 99 Tc, 129 I and 135 Cs discharged from LWR. Their isotopes, such as 127 I, 133 Cs, 134 Cs and 137 Cs were also considered. By loading the isotopes ( 99 Tc or 127 I and 129 I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe·y and 2.79%/GWe.y can be obtained for 99 Tc and 129 I, respectively. The transmuted amounts of 99 Tc and 129 I are equal to the outputs from 11.8 and 6.2 1000MWe-class PWRs. Because of the very low capture cross section of 135 Cs and the effect of other cesium isotopes, 135 Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000MWe-class PWRs. Based on these results, 99 Tc and 129 I can be transmuted conveniently and higher transmutation performance can be obtained by the Super Fast Reactor. However, the transmutation of 135 Cs is very difficult and the transmuted amount is less than that produced by the Super Fast Reactor. It turns out that the 135 Cs transmutation is a challenge not only for the Super Fast Reactor but also for other commercial fast reactors. (author)

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

    International Nuclear Information System (INIS)

    Azpitarte, O.; Ramilo, L.

    2013-01-01

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

  12. 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.)

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

    International Nuclear Information System (INIS)

    Ninokata, H.; Kamide, H.

    2011-01-01

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

  14. Modeling delayed neutron monitoring systems for fast breeder reactors

    International Nuclear Information System (INIS)

    Bunch, W.L.; Tang, E.L.

    1983-10-01

    The purpose of the present work was to develop a general expression relating the count rate of a delayed neutron monitoring system to the introduction rate of fission fragments into the sodium coolant of a fast breeder reactor. Most fast breeder reactors include a system for detecting the presence of breached fuel that permits contact between the sodium coolant and the mixed oxide fuel. These systems monitor for the presence of fission fragments in the sodium that emit delayed neutrons. For operational reasons, the goal is to relate the count rate of the delayed neutron monitor to the condition of the breach in order that appropriate action might be taken

  15. Critical experiment and analysis for nitride fuel fast reactor using FCA

    International Nuclear Information System (INIS)

    Andoh, Masaki; Iijima, Susumu; Okajima, Shigeaki; Sakurai, Takeshi; Oigawa, Hiroyuki

    2000-03-01

    As a research on FBR with new types of fuel, a series of experiments on a nitride fuel fast reactor was carried out at Fast Critical Assembly (FCA) to evaluate the calculation accuracy on the neutronic characteristics of the reactor. In this study, criticality, sample reactivity worth and sodium void reactivity worth were measured in the FCA XIX-2 core simulating a nitride fuel fast reactor and were analyzed using the standard analysis method for FCA fast reactor cores. The accuracy of the analysis on the effective multiplication factor was the same as those of the other FCA cores. For the plate sample reactivity worth, the calculation on the radial distribution of plutonium plate reactivity worth overestimated the measurement depending on the distance from the center of the core. For the sodium void reactivity worth, the calculation overestimated the experimental value 10 to 20% at the core center, while the overestimation was improved as the voided position was located at the core boundary. It was found that the transport effect was considerable even at the center of the core. It was considered that the calculation accuracy on the non-leakage term of the void reactivity worth and transport correction should be improved. (author)

  16. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors

    International Nuclear Information System (INIS)

    Vega R, A. K.; Espinosa P, G.; Gomez T, A. M.

    2016-09-01

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  17. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  18. Development, operational experience and implications for future design of fast reactors in Western Europe

    International Nuclear Information System (INIS)

    Brandstetter, A.; Broomfield, A.M.; Saitcevsky, B.

    1990-01-01

    Over the past 30 years, the partners now collaborating in Europe on fast-reactor development have taken the technology from a theoretical possibility to an engineering reality. In that time there has been a progression from experimental zero-energy facilities followed by small power-producing engineering test reactors, to prototype reactors and a large commercial-size reactor. The paper describes the highlights of the reactor programmes in the partner countries and by example illustrates the experience gained from reactor operation and some of the principal activities in the supporting development programme. These include such topics as fuel performance, fast-neutron physics, liquid-metal thermal hydraulics, sodium chemistry, instrumentation and safety aspects. The paper concludes by summarizing some of the main objectives of the current development programme, which is directed to the support of the European Fast Reactor design being prepared by the European design and construction companies. (author)

  19. SSTAR: The US lead-cooled fast reactor (LFR)

    International Nuclear Information System (INIS)

    Smith, Craig F.; Halsey, William G.; Brown, Neil W.; Sienicki, James J.; Moisseytsev, Anton; Wade, David C.

    2008-01-01

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

  20. Liquid metal cooled experimental fast reactor simulator

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  1. 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

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

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

    2010-10-01

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

  3. Construction of fast experimental reactor 'Joyo' from start of construction to criticality

    International Nuclear Information System (INIS)

    Sakata, Hajime

    1977-01-01

    The fast experimental reactor ''Joyo'' is a sodium-cooled, fast neutron reactor using mixed oxide of uranium and plutonium, the first in Japan. The purposes of its construction are to experience and solve the various technical problems expected in the constructions of the prototype reactor ''Monju'' and future practical reactors, and to use as the irradiation facility for developing the fuel and material for fast breeder reactors in Japan after the completion. The construction finished by the end of 1974, and the synthetic functional test was carried out for about two years thereafter. The whole installation was handed over to PNC on March 8, 1977. The reactor attained the criticality on April 24, 1977. The outline of the construction works is described. ''Guidance to the structural design of sodium machinery for Joyo'' was compiled, and the analysis was made according to it. Moreover, various inspection standards regarding welding, electrical machinery, fuel and others were made. The revision of the design for improving the safety and performance was made during the construction at all times. The synthetic functional test was carried out for about two years on 266 items, and subsequently, the criticality test was completed satisfactorily. (Kako, I.)

  4. Core design of long life-cycle fast reactors operating without reactivity margin

    International Nuclear Information System (INIS)

    Aristova, E. N.; Baydin, D. F.; Gol'din, V. Y.; Pestryakova, G. A.; Stoynov, M. I.

    2012-01-01

    In this paper we consider a possibility of designing a fast reactor core that operates without reactivity margin for a long time. This study is based on the physical principle of fast reactor operating in a self-adjustable neutron-nuclear regime (SANNR-1) introduced by L.P. Feoktistov (1988-1993) and improved by V. Ya. Gol'din SANNR-2 (1995). The mathematical modeling of active zones of fast reactors in SANNR modes is held by authors since 1992. The numerical simulation is based on solving the neutron transport equation coupled with quasi-diffusion equations. The calculations have been performed using standard 26 energy groups. We use a hierarchy of spatial models of 1D, 1.5D, 2D, and 3D geometries. The spatial models of higher dimensionality are used for verification of results. The calculations showed that operation of the reactor in this mode increases its efficiency, safety and simplifies management. It is possible to achieve continuous work of the reactor in SANNR-2 during 7-10 years without fuel overloads by means of further optimization of the mode. Small reactivity margin is used only for the reactor start up. After first 10-15 days the reactor in SANNR-2 operates without reactivity margin. (authors)

  5. Real time simulation method for fast breeder reactors dynamics

    International Nuclear Information System (INIS)

    Miki, Tetsushi; Mineo, Yoshiyuki; Ogino, Takamichi; Kishida, Koji; Furuichi, Kenji.

    1985-01-01

    The development of multi-purpose real time simulator models with suitable plant dynamics was made; these models can be used not only in training operators but also in designing control systems, operation sequences and many other items which must be studied for the development of new type reactors. The prototype fast breeder reactor ''Monju'' is taken as an example. Analysis is made on various factors affecting the accuracy and computer load of its dynamic simulation. A method is presented which determines the optimum number of nodes in distributed systems and time steps. The oscillations due to the numerical instability are observed in the dynamic simulation of evaporators with a small number of nodes, and a method to cancel these oscillations is proposed. It has been verified through the development of plant dynamics simulation codes that these methods can provide efficient real time dynamics models of fast breeder reactors. (author)

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

    International Nuclear Information System (INIS)

    Nagata, T.; Ieda, Y.

    2002-01-01

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

  7. Lead-based Fast Reactor Development Plan and R&D Status in China

    International Nuclear Information System (INIS)

    Wu Yican

    2013-01-01

    • Lead-based fast reactors have good potential for waste transmutation, fuel breeding and energy production, which has been selected by CAS as the advanced reactor development emphasis with the support of ADS program and MFE program. Sharing of technologies R&D is possible among GIF/ADS/Fusion. • The concepts and test strategy of series China lead-based fast reactors (CLEAR) have been developed. The preliminary engineering design and safety analysis of CLEAR-I are underway. • Technology R&D on CLEAR with series lead alloy loops and accelerator-based neutron generator have been constructed or under construction. • CLEAR series reactor design and construction have big challenges, widely international cooperation on reactor design and technology R&D is welcome

  8. European Union: Review of fast reactor related activities

    International Nuclear Information System (INIS)

    Goethem, G. van; Hugon, M.

    1998-01-01

    The European Commission (EC) continued its fast reactor research activities on the same lines as in the past, but with the main emphasis on partitioning and transmutation (P and T) of long-lived radionuclides. The work was carried out by research institutions in the Member States and by the EC Joint Research Centre (JRC) as cost shared actions. The JRC has also been performing its own programme through institutional and competitive research activities. The JRC institutes involved in these studies are the Institute of Systems, Informatics and Safety (ISIS) in Ispra (I), the Institute for Transuranium Elements (ITU) in Karlsruhe (D) and the Institute for Advanced Materials in Petten. This paper summarizes the main activities performed in the field of (i) fast reactor safety and of (ii) partitioning and transmutation. (author)

  9. A study of reactor neutrino monitoring at the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

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

    2012-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3 m from the JOYO reactor core of 140 MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11±1.24(stat.)±0.46(syst.) events/day. Although the statistical significance of the measurement was not enough, backgrounds in such a compact detector at the ground level were studied in detail and MC simulations were found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

  10. 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

  11. Design characteristics of research zero power fast reactor Lasta

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  12. Progress report on fast breeder reactor development at PNC, Japan, October - December, 1974

    International Nuclear Information System (INIS)

    1975-03-01

    Following the completion of building construction and equipment installation for the experimental fast breeder reactor ''Joyo'' at PNC's Oarai Engineering Center, hydrostatic pressure and leak tests were conducted on the reactor vessel. For the prototype fast breeder reactor ''Monju'', specification was finalized after the design adjustment. For the period from October to December, 1974, the following matters are described: construction of the Joyo, design of the Monju, reactor physics, components and equipments, instruments and control, sodium technology, fuel and material research and development, safety research and development, and steam generator. (Mori, K.)

  13. Study on effects of development of reactor constant in fast reactor analysis

    International Nuclear Information System (INIS)

    Chiba, Gou

    2002-12-01

    Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

  14. 4. generation sodium-cooled fast reactors. The ASTRID technological demonstrator

    International Nuclear Information System (INIS)

    2012-12-01

    The sodium-cooled fast reactor (SFR) concept is one of the four fast neutron concepts selected by the Generation IV International Forum (GIF). SFRs have favourable technical characteristics and they are the sole type of reactor for which significant industrial experience feedback is available. After a discussion of the past experience gained on fast breeder reactors in the world (benefits, difficulties and problematics), the authors discuss the main improvement domains and the associated R and D advances (reactor safety, prevention and mitigation of severe accidents, the sodium-water risk, detection of sodium leaks, increased availability, instrumentation and inspection, control and repairability, assembly handling and washing). Then, they describe the technical requirements and safety objectives of the ASTRID experimental project, notably with its reactivity management, cooling management, and radiological containment management functions. They describe and discuss requirements to be met and choices made for Astrid, and the design options for its various components (core and fuels, nuclear heater, energy conversion system, fuel assembly handling, instrumentation and in-service inspection, control and command). They present the installations which are associated with the ASTRID cycle, evoke the development and use of simulations and codes, describe the industrial organization and the international collaboration about the ASTRID project, present the planning and cost definition

  15. A review of fast reactor activities in Switzerland - April 1985

    International Nuclear Information System (INIS)

    Wydler, P.

    1986-01-01

    In the nuclear fission field, there are activities related to many different reactor concepts, including the Light Water Reactor, the Light Water High Converter Reactor, the High Temperature Reactor, the Liquid Metal Fast Breeder Reactor and the recently proposed new concept of a small heating reactor. In 1984 the total expenditure for fast reactor activities remained the same as that in the previous year, but the budget for 1985 has declined. The 6.0 million Swiss Francs expended in 1984 have been allocated to an LMFBR safety progamme (46%) and a fuel development programme (54%). All activities reported below are carried out at the Federal Institute for Reactor Research (EIR). In the natural convection studies described in Section 5, the Nuclear Engineering Laboratory (LKT) of the Federal Institute of Technology at Zuerich is actively participating. In the past twelve months collaboration with foreign research organizations in the Federal Republic of Germany, France, Italy (JRC Ispra) and the U.K. for the LMFBR safety programme, and the Federal Republic of Germany and the U.S.A. for the fuel development programme has proved to be very fruitful. In this context an attachment agreement with CEA-DERS at Cadarache is worth mentioning, since it enabled an EIR staff member to participate in the prediction and analysis of the SCARABEE-APL in-pile tests

  16. 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)

  17. Physics of plutonium recycling: volume V. Plutonium recycling in fast reactors

    International Nuclear Information System (INIS)

    1996-01-01

    As part of a programme proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed. In this report, the multi-recycle performance of the metal-fuelled benchmark is evaluated. Benchmark results assess the reactor performance and toxicity behaviour in a closed nuclear fuel cycle for a parametric variation of the conversion ratio between 0.5 and 1.0. Results indicate that a fast burner reactor closed fuel cycle can be utilised to significantly reduce the radiotoxicity originating in the LWR cycle which would otherwise be destined for burial. (Author). tabs., figs., refs

  18. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1980-01-01

    The performance test on the reactor power increase to 75 MW was started on July 3, and the target of 75 MW was reached on July 16, in the experimental fast reactor Joyo. The tests on the heat transport characteristics, power coefficient, the response to the change of outlet temperature, the loss of external power supply and so on were carried out, and the performance test was finished on August 23, except the test of 75 MW continuous operation. The annual inspection of the systems is being carried out in parallel with the regular inspection. The design to prepare for the manufacture of the prototype fast reactor Monju is being prepared. The analysis of decay heat removal is being carried out, and various calculation codes were developed. The technological survey on overseas LMFBRs is being made. The conceptual design of the demonstration reactor is being prepared. The research and development of reactor physics, structural components for Joyo and Monju, instrumentation and control, sodium technology, fuel materials, structural materials, safety problems and steam generators are reported. The tests on the transient boiling of sodium, fuel failure propagation, heat transfer between molten materials, post-accident decay heat removal and so on have been carried out. (Kako, I.)

  19. Consultancy on 'Knowledge preservation in the area of fast reactor technology'. Working material

    International Nuclear Information System (INIS)

    2002-01-01

    The fast reactor, which can generate electricity and breed additional fissile material for future fuel stocks is a resource that will be needed when economic uranium supplies for the advanced light water reactors or other thermal-spectrum options diminish. Further, the fast-fission fuel cycle in which material is recycled offers the flexibility needed to contribute decisively towards solving the problem of growing spent fuel inventories by greatly reducing the volume of high-level waste that must be disposed of in long-term repositories. This is a waste management option that also should be retained for future generations. The fast reactor has been the subject of research and development programs in a number of countries for upwards of 40 years. Now, despite early sharing and innovative worldwide research and development, ongoing work is confined to China, India, Japan, the Republic of Korea, and Russia. Information generated worldwide will be needed in the future. Presently, it is in danger of being lost even in those countries continuing the work. Some countries have already taken the issue of knowledge preservation seriously: Japan, France, Britain, and Russia, in particular. At worst, valuable contributory information elsewhere will be lost and would have to be regenerated when needed. The IAEA initiative seeks to establish a comprehensive, international inventory of fast reactor data and knowledge, which would be sufficient to form the basis for fast reactor development in 20 to 40 years from now. The Agency is in a good position to provide the framework for knowledge preservation efforts. Under Article III of its Statute, the IAEA is mandated to encourage and assist research on, and development and practical application of atomic energy for peaceful uses throughout the world. Obviously, an important aspect of this mandate is maintaining and increasing the knowledge that is necessary for the technological development. The main objectives of the consultancy

  20. Possible future roles for fast breeder reactors Part 1 and 2

    International Nuclear Information System (INIS)

    1978-06-01

    Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

  1. 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)

  2. Benchmark test of evaluated nuclear data files for fast reactor neutronics application

    International Nuclear Information System (INIS)

    Chiba, Go; Hazama, Taira; Iwai, Takehiko; Numata, Kazuyuki

    2007-07-01

    A benchmark test of the latest evaluated nuclear data files, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0, has been carried out for fast reactor neutronics application. For this benchmark test, experimental data obtained at fast critical assemblies and fast power reactors are utilized. In addition to comparing of numerical solutions with the experimental data, we have extracted several cross sections, in which differences between three nuclear data files affect significantly numerical solutions, by virtue of sensitivity analyses. This benchmark test concludes that ENDF/B-VII.0 predicts well the neutronics characteristics of fast neutron systems rather than the other nuclear data files. (author)

  3. Energy-averaged neutron cross sections of fast-reactor structural materials

    International Nuclear Information System (INIS)

    Smith, A.; McKnight, R.; Smith, D.

    1978-02-01

    The status of energy-averaged cross sections of fast-reactor structural materials is outlined with emphasis on U.S. data programs in the neutron-energy range 1-10 MeV. Areas of outstanding accomplishment and significant uncertainty are noted with recommendations for future efforts. Attention is primarily given to the main constituents of stainless steel (e.g., Fe, Ni, and Cr) and, secondarily, to alternate structural materials (e.g., V, Ti, Nb, Mo, Zr). Generally, the mass regions of interest are A approximately 50 to 60 and A approximately 90 to 100. Neutron total and elastic-scattering cross sections are discussed with the implication on the non-elastic-cross sections. Cross sections governing discrete-inelastic-neutron-energy transfers are examined in detail. Cross sections for the reactions (n;p), (n;n',p), (n;α), (n;n',α) and (n;2n') are reviewed in the context of fast-reactor performance and/or diagnostics. The primary orientation of the discussion is experimental with some additional attention to the applications of theory, the problems of evaluation and the data sensitivity of representative fast-reactor systems

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

    International Nuclear Information System (INIS)

    Motta, M.; Giovannini, R.

    1985-07-01

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

  5. 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

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

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Tentner, A. M.

    2007-01-01

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

  7. Three-dimensional Core Design of a Super Fast Reactor with a High Power Density

    International Nuclear Information System (INIS)

    Cao, Liangzhi; Oka, Yoshiaki; Ishiwatari, Yuki; Ikejiri, Satoshi; Ju, Haitao

    2010-01-01

    The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/cm 3 . The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied

  8. Gas cooled fast breeder reactors using mixed carbide fuel

    International Nuclear Information System (INIS)

    Kypreos, S.

    1976-09-01

    The fast reactors being developed at the present time use mixed oxide fuel, stainless-steel cladding and liquid sodium as coolant (LMFBR). Theoretical and experimental designing work has also been done in the field of gas-cooled fast breeder reactors. The more advanced carbide fuel offers greater potential for developing fuel systems with doubling times in the range of ten years. The thermohydraulic and physics performance of a GCFR utilising this fuel is assessed. One question to be answered is whether helium is an efficient coolant to be coupled with the carbide fuel while preserving its superior neutronic performance. Also, an assessment of the fuel cycle cost in comparison to oxide fuel is presented. (Auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

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

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

    International Nuclear Information System (INIS)

    1999-01-01

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

  11. Utility industry evaluation of the Sodium Advanced Fast Reactor

    International Nuclear Information System (INIS)

    Burstein, S.; DelGeorge, L.O.; Tramm, T.R.; Gibbons, J.P.; High, M.D.; Neils, G.H.; Pilmer, D.F.; Tomonto, J.R.; Wells, J.T.

    1990-02-01

    A team of utility industry representatives evaluated the Sodium Advanced Fast Reactor plant design, a current liquid metal reactor design created by an industrial team led by Rockwell International under Department of Energy sponsorship. The utility industry team concluded that the plant design offers several attractive characteristics, especially in the safety arena, as well as preserving the traditional attraction of liquid metal reactors, very high fuel utilization. Specific comments and recommendations are provided as a contribution towards improving an already attractive plant design. 18 refs

  12. Safety design study of fast breeder reactors in Japan

    International Nuclear Information System (INIS)

    Miura, M.; Inagaki, T.

    1992-01-01

    This paper reports on two fast breeder reactor (FBR) concepts, the tank type and the loop type, that have been studied as possible reactor designs to be used for a demonstration FBR (DFBR). The basic principle fo the DFBR design is to ensure plant safety through a defense-in-depth methodology. Improvements in the seismic and thermal stress designs have been attempted for both reactor concepts. The system design study strives to maximize the reliability of the safety-related systems and to rationalize commercialization of the plant

  13. Safe Management Of Fast Reactors: Towards Sustainability

    International Nuclear Information System (INIS)

    Dreimanis, Andrejs

    2015-01-01

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

  14. Evolution of the technical concept of fast reactors. The concept of BREST

    International Nuclear Information System (INIS)

    Orlov, V.V.

    2001-01-01

    Having understood that conventional power was limited by available fuel resources, as well as the environmental concern, and willing to use the advantages of defense nuclear power achievements, the development of civil nuclear power was initiated. Scarce supply of uranium has been a matter of concern from the very beginning of nuclear power development, but plutonium produced in the thermal reactors was supposed to be used as fuel for the fast reactors which would not be limited by fuel resources. In order to attain high breeding ratio and high power density, the first generation of fast reactors were designed with sodium coolant, uranium blanket to make up for a decrease in breeding ratio if uranium oxides were used as fuel. Development of nuclear power in the sixties and seventies was followed by stagnation. Lessons learned from a 50-year experience and new conditions set for power industry demand a new concept of fast reactor which would meet a variety of cost-efficiency and safety requirements in their present understanding. Development of fast breeders in Russia began after commissioning of BN-350 and completion of BN-600 design. According to present demands BREST reactors should be designed so as to implement consistently the principles of natural safety without deviation from materials and technology which was proven in defense and civil nuclear power facilities

  15. TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

    1983-01-01

    TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

  16. Challenges and achievements - Prototype Fast Breeder Reactor construction

    International Nuclear Information System (INIS)

    Subramani, V.A.; Dhere, S.S.; Manoharan, V.; Subbaraman, P.

    2010-01-01

    Prototype fast breeder reactor presently under construction poses several challenges in materials, design and construction. The civil structure and equipment are of very large size and complex in nature. This paper presents the features of the design and construction of the PFBR excavation, raft, civil structure of the nuclear island connected buildings and reactor vault. This paper also brings out the details of the large size equipment of special stainless steel and handling structure for their lifting and placement inside the reactor vault. The paper is divided into three parts viz. introduction, challenges and achievements during construction of civil structures and erection of large size components. (author)

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

    International Nuclear Information System (INIS)

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

    1980-05-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  19. Reprocessing of fast neutron reactor fuel

    International Nuclear Information System (INIS)

    Bourgeois, M.

    1981-05-01

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

  20. Lead- or Lead-bismuth-cooled fast reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Courouau, J.L.; Dufour, P.; Guidez, J.; Latge, C.; Martinelli, L.; Renault, C.; Rimpault, G.

    2014-01-01

    Lead-cooled fast reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. So far no lead-cooled reactors have existed in the world except lead-bismuth-cooled reactors in soviet submarines. Some problems linked to the use of the lead-bismuth eutectic appeared but were satisfactorily solved by a more rigorous monitoring of the chemistry of the lead-bismuth coolant. Lead presents various advantages as a coolant: no reactivity with water and the air,a high boiling temperature and low contamination when irradiated. The main asset of the lead-bismuth alloy is the drop of the fusion temperature from 327 C degrees to 125 C degrees. The main drawback of using lead (or lead-bismuth) is its high corrosiveness with metals like iron, chromium and nickel. The high corrosiveness of the coolant implies low flow velocities which means a bigger core and consequently a bigger reactor containment. Different research programs in the world (in Europe, Russia and the USA) are reviewed in the article but it appears that the development of this type of reactor requires technological breakthroughs concerning materials and the resistance to corrosion. Furthermore the concept of lead-cooled reactors seems to be associated to a range of low output power because of the compromise between the size of the reactor and its resistance to earthquakes. (A.C.)

  1. 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)

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

    International Nuclear Information System (INIS)

    1986-01-01

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

  3. Japan: The Experimental Fast Reactor JOYO. Profile 12

    International Nuclear Information System (INIS)

    2017-01-01

    The experimental fast reactor JOYO of the Japan Atomic Energy Agency (JAEA) is the first sodium-cooled fast reactor (SFR) in Japan. JOYO attained its initial criticality as a breeder core (MK-I core) in 1977. During the MK-I operation, which consisted of two 50 MWt and six 75 MWt duty cycles, the basic characteristics of plutonium (Pu) and uranium (U) mixed oxide (MOX) fuel core and sodium cooling system were investigated and the breeding performance was verified. In 1983, the reactor increased its thermal output up to 100 MWt in order to start the irradiation tests of fuels and materials to be used mainly for other SFRs. Thirty-five duty cycle operations and many irradiation tests were successfully carried out using the MK-II core by 2000. The core was then modified to the MK-III core in 2003. In order to obtain higher fast neutron flux, the core was modified from one region core to two region core with different Pu fissile contents. Accordingly, the reactor power increased up to 140 MWt together with a renewal of intermediate heat exchangers (IHXs) and dump heat exchangers (DHXs). The rated power operation of the MK-III core started in 2004. The MK-III core has been used for the irradiation tests of fuels and materials for future SFRs and other R&D fields like innovative nuclear energy systems and technologies as well. This powerful neutron irradiation flux has an advantage especially for high burn-up fuel irradiation and material irradiation with high neutron dose. This paper shows the outline of the irradiation irradiation irradiation irradiation irradiation capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop innovative nuclear energy systems and technologies.

  4. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    Scott, D.

    1979-01-01

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

  5. Status of fast reactor activities in the Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-07-01

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

  6. Status of fast reactor activities in the Russian Federation

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  7. The seismic assessment of fast reactor cores in the UK

    International Nuclear Information System (INIS)

    Duthie, J.C.; Dostal, M.

    1988-01-01

    The design of the UK Commercial Demonstration Fast Reactor (CDFR) has evolved over a number of years. The design has to meet two seismic requirements: (i) the reactor must cause no hazard to the public during or after the Safe Shutdown Earthquake (SSE); (ii) there must be no sudden reduction in safety for an earthquake exceeding the SSE. The core is a complicated component in the whole reactor. It is usually represented in a very simplified manner in the seismic assessment of the whole reactor station. From this calculation, a time history or response spectrum can be generated for the diagrid, which supports the core, and for the above core structure, which supports the main absorber rods. These data may then be used to perform a detailed assessment of the reactor core. A new simplified model of the core response may then be made and used in a further calculation of the whole reactor. The calculation of the core response only, is considered in the remainder of this paper. One important feature of the fast reactor core, compared with other reactors, is that the components are relatively thin and flexible to promote neutron economy and heat transfer. A further important feature is that there are very small gaps between the wrapper tubes. This leads to very strong fluid-coupling effects. These effects are likely to be beneficial, but adequate techniques to calculate them are only just being developed. 9 refs, figs

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

    International Nuclear Information System (INIS)

    Perez-Navarro Gomez, A.

    1973-01-01

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

  9. A multi-group neutron noise simulator for fast reactors

    International Nuclear Information System (INIS)

    Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe

    2013-01-01

    Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring

  10. Control rod studies in small and medium sized fast reactors

    International Nuclear Information System (INIS)

    John, T.M.; Mohanakrishnan, P.; Mahalakshmi, B.; Singh, R.S.

    1988-01-01

    Control rods are the primary safety mechanism in the operation of fast reactors. Neutronic parameters associated with the control rods have to be evaluated precisely for studying the behaviour of the reactor under various operating conditions. Control rods are strong neutron absorbers discretely distributed in the reactor core. Accurate estimation of control rod parameters demand, in principle transport theory solutions in exact geometry. But computer codes for such evaluations usually consume exorbitantly large computer time and memory for even a single parameter evaluation. During the design of reactors, evaluation of these parameters will be required for many configurations of control rods. In this paper, the method used at Indira Gandhi Centre for Atomic Research for estimating the parameters associated with control rods is presented. Diffusion theory solutions were used for computations. A scheme using three dimensional geometry represented by triangular meshes and diffusion theory solutions in few energy groups for control rod parameter evaluation is presented. This scheme was employed in estimating the control rod parameters in a 500 Mw(e) fast reactor. Error due to group collapsing is estimated by comparing with 25 group calculations in three dimensions for typical cases. (author). 5 refs, 4 figs, 3 tabs

  11. A review of fast reactor program in Japan - April 1984

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1984-01-01

    The fast breeder reactor development project in PNC has been in progress steadily in these eighteen years. Concerning the experimental fast reactor, JOYO, the MK-II core attained criticality on November 22, 1982 with 51 fuel assemblies, and received the ''Certificate of Inspection before Operation'' from Government Authority on March 31, 1983, after 100 hours operation with the rated output of 100 MW. Since then, the core has been utilized to implement irradiation bed characteristics test, and to irradiate fuels and structural materials especially for the prototype reactor MONJU. With respect to the prototype reactor MONJU, the installation permit was issued on May 27, 1983, from the prime minister, and the contracts of the first stage between PNC and fabricators were made recently. At the same time, almost all the licenses of preparatory construction works were issued by March 1983, and preparatory construction works were started in April 1983. On the other hand, conceptual design of a demonstration reactor is now under way in a close cooperation with concerned authorities and utilities, as well as investigations of the way of conducting necessary research and development

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

    International Nuclear Information System (INIS)

    Balz, W.

    1977-01-01

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

  13. Seismic responses of a pool-type fast reactor with different core support designs

    International Nuclear Information System (INIS)

    Wu, Ting-shu; Seidensticker, R.W.

    1989-01-01

    In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs

  14. Some aspects of fast reactor economics

    International Nuclear Information System (INIS)

    Kazachkovskij, O.D.

    1996-01-01

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

  15. Basic principles of accounting and control of nuclear materials in the BOR-60 experimental fast reactor

    International Nuclear Information System (INIS)

    Gryazev, V.M.; Gadzhiev, G.I.; Alekseev, I.N.

    1979-01-01

    Under a contract with the International Atomic Energy Agency, the V.I. Lenin Atomic Reactor Research Institute is currently carrying out a study of ways of organizing a nuclear materials accounting and control system for the BOR-60 fast reactor. Some results of this study are presented in the paper. The special physical and technological features of fast reactors create additional difficulties in safeguards systems and give rise to a number of new possibilities for the illicit removal of nuclear materials. These questions are discussed with reference to the BOR-60 reactor but the conclusions are probably applicable to all fast reactors. The proposed accounting and control system is based on non-destructive measurements of the amount of fissile materials in the operating fuel assemblies and screened bundles of the reactor, on the independent control of the principal facility parameters (a list of which is given) and on an automated information collection and evaluation system. Visual means of inspection can be very effective in fast reactor safeguards systems, especially for controlling storage, but they are not used with the BOR-60 reactor. (author)

  16. Education & Training in Support to Sodium Fast Reactors Around the World

    International Nuclear Information System (INIS)

    Latgé, C.; Soucille, M.; Grandy, C.; Xu Mi; Garbil, R.; Monti, S.; Sai Baba, M.; Chellapandi, P.; Kitabata, T.; Kim, Y-G

    2013-01-01

    The results of these ambitious and long term strategies are: - first the creation of a new generation of skilled nuclear engineers in the field. - secondly a share of knowledge gained through experimental studies carried out in research laboratories as well as feedback from fast reactors operation, - thirdly a standardized information on safety, - and finally the creation of a “Sodium Fast Reactor community” is promoted, able to debate, share the knowledge and suggest new tracks for a better definition of design and operating rules

  17. Immediate relation of ING to fast breeder reactor programs

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, W B

    1969-07-01

    The future large-scale use of nuclear energy is linked in the United States and other major countries to their fast breeder reactor development. Very serious basic problems have been discovered within the last two years, limiting the life in the high fast neutron flux at appropriate temperatures of materials, in particular of metals suitable for fuel cladding in sodium coolant. There is therefore a most urgent need for materials testing facilities under controlled conditions of temperature and neutron flux at sufficiently high ratings to match or surpass those required in commercially competitive fast breeder reactors. None of the test facilities yet planned for 1976 or sooner in the western world appears to match these conditions. The problem is mainly the difficulty of providing the high neutron flux effectively continuously. The spallation reaction in heavy elements was chosen as the basis of ING - the intense neutron generator, because it is the only known reaction that promises a fast neutron source density that is higher than can be controlled from the fission process. It is suggested that several countries will wish to consider urgently whether they should also explore the spallation reaction for the purpose of a fast neutron irradiation test facility. In view of the discontinuance of the ING project in Canada a favourable opportunity will exist over the next few months 10 obtain from Canada by direct personal contact details of the significant study that has been carried on for ING over the last five years. In the event that satisfactory materials are established within the lifetime of the spallation facilities they may continue to be used for the production of selected isotopes more profitably produced in high neutron fluxes. The facilities may be also used for the desirable preirradiation of thorium reactor fuel. The other research purposes planned for ING could also be served. (author)

  18. Immediate relation of ING to fast breeder reactor programs

    International Nuclear Information System (INIS)

    Lewis, W.B.

    1969-01-01

    The future large-scale use of nuclear energy is linked in the United States and other major countries to their fast breeder reactor development. Very serious basic problems have been discovered within the last two years, limiting the life in the high fast neutron flux at appropriate temperatures of materials, in particular of metals suitable for fuel cladding in sodium coolant. There is therefore a most urgent need for materials testing facilities under controlled conditions of temperature and neutron flux at sufficiently high ratings to match or surpass those required in commercially competitive fast breeder reactors. None of the test facilities yet planned for 1976 or sooner in the western world appears to match these conditions. The problem is mainly the difficulty of providing the high neutron flux effectively continuously. The spallation reaction in heavy elements was chosen as the basis of ING - the intense neutron generator, because it is the only known reaction that promises a fast neutron source density that is higher than can be controlled from the fission process. It is suggested that several countries will wish to consider urgently whether they should also explore the spallation reaction for the purpose of a fast neutron irradiation test facility. In view of the discontinuance of the ING project in Canada a favourable opportunity will exist over the next few months 10 obtain from Canada by direct personal contact details of the significant study that has been carried on for ING over the last five years. In the event that satisfactory materials are established within the lifetime of the spallation facilities they may continue to be used for the production of selected isotopes more profitably produced in high neutron fluxes. The facilities may be also used for the desirable preirradiation of thorium reactor fuel. The other research purposes planned for ING could also be served. (author)

  19. Technical Meeting on Passive Shutdown Systems for Liquid Metal-Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2015-01-01

    A major focus of the design of modern fast reactor systems is on inherent and passive safety. Specific systems to improve reactor safety performance during accidental transients have been developed in nearly all fast reactor programs, and a large number of proposed systems have reached various stages of maturity. This Technical Meeting on Passive Shutdown Systems for Fast Reactors, which was recommended by the Technical Working Group on Fast Reactors (TWG-FR), addressed Member States’ expressed need for information exchange on projects and programs in the field, as well as for the identification of priorities based on the analysis of technology gaps to be covered through R&D activities. This meeting was limited to shutdown systems only, and did not include other passive features such as natural circulation decay heat removal systems etc.; however the meeting catered to passive shutdown safety devices applicable to all types of fast neutron systems. It was agreed to initiate a new study and produce a Nuclear Energy Series (NES) Technical Report to collect information about the existing operational systems as well as innovative concepts under development. This will be a useful source for member states interested in gaining technical expertise to develop passive shutdown systems as well as to highlight the importance and development in this area

  20. Risk Management for Sodium Fast Reactors.

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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