WorldWideScience

Sample records for bn-350 fast-breeder reactor

  1. The passive nondestructive assay of the plutonium content of spent-fuel assemblies from the BN-350 fast-breeder reactor in the city of Aqtau, Kazakhstan

    CERN Document Server

    Lestone, J P; Rennie, J A; Sprinkle, J K; Staples, P; Grimm, K N; Hill, R N; Cherradi, I; Islam, N; Koulikov, J; Starovich, Z

    2002-01-01

    The International Atomic Energy Agency is presently interested in developing equipment and techniques to measure the plutonium content of breeder reactor spent-fuel assemblies located in storage ponds before they are relocated to more secure facilities. We present the first quantitative nondestructive assay of the plutonium content of fast-breeder reactor spent-fuel assemblies while still underwater in their facility storage pond. We have calibrated and installed an underwater neutron coincidence counter (Spent Fuel Coincidence Counter (SFCC)) in the BN-350 reactor spent-fuel pond in Aqtau, Kazakhstan. A procedure has been developed to convert singles and doubles (coincidence) neutron rates observed by the SFCC into the total plutonium content of a given BN-350 spent-fuel assembly. The plutonium content has been successfully determined for spent-fuel assemblies with a contact radiation level as high as approx 10 sup 5 Rads/h. Using limited facility information and multiple measurements along the length of spe...

  2. Planning of the BN-350 reactor decommissioning

    International Nuclear Information System (INIS)

    Klepikov, A.Kh.; Tazhibayeva, I.L.; Zhantikin, T.M.; Baldov, A.N.; Nazarenko, P.I.; Koltyshev, S.M.; Wells, P.B.

    2002-01-01

    The experimental and commercial BN-350 NPP equipped with a fast neutron sodium cooled reactor is located in Kazakhstan near the Aktau city on the Caspian Sea coast. It was commissioned in 1973 and intended for weapon-grade plutonium production and as stream supply to a water desalination facility and the turbines of the Mangyshlak Atomic Energy Complex. Taking into account technical, financial and political issues, the Government of Kazakhstan enacted the Decree no. 456 'On Decommissioning of the Reactor BN-350 in the Aktau City of the Mangystau Region'. Because the decision on reactor decommissioning was adopted before the end of scheduled operation (2003), the plan to decommission the BN-350 reactor has not yet been developed. To determine the activities required for ensuring reactor safety and in preparation for decommission in the period prior, the development and ensuring approval by the Republic of Kazakhstan Government of the decommissioning plan, a 'Plan of Priority Actions for BN-350 Reactor Decommissioning' was developed and approved. Actions provided for in the plan include the following: Development of BN-350 Reactor Decommissioning Plan; Accident prevention during the period of transition; Unloading nuclear fuel from reactor and draining the coolant from the heat exchange circuits. Decommission is defined as a complex of administrative and technical actions taken to allow the removal of some or all of regulatory controls over a nuclear facility. These actions involve decontamination, dismantling and removal of radioactive materials, waste, components and structures. They are carried out to achieve a progressive and systematic reduction in radiological hazards and are undertaken on the basis of planning and assessment in order to ensure safety decommissioning operations. In accordance with the decision of Kazakhstan Government, three basic stages for BN-350 reactor decommissioning are envisaged: First stage - Placement of BN-350 into long-term storage

  3. 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

  4. k-eff of the Bn-350 reactor fuel by transportation

    International Nuclear Information System (INIS)

    Lado, A. V.; Romanenko, O. G.; Tazhibaeva, I. L.

    2001-01-01

    There is packaging of nuclear fuel on the BN-350 fast breeder reactor, Actau, now. The analysis of criticality while this procedure was done in the Safety Analysis Report . Keeping in mind the planning displacement of the fuel to a site of long-term storage, the criticality assessment of the fuel packed into transportation cask carried out in this paper

  5. 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

  6. 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.

  7. 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)

  8. 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

  9. 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

  10. 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)

  11. 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)

  12. Fast breeder reactors

    International Nuclear Information System (INIS)

    Ollier, J.L.

    1987-01-01

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

  13. Cobalt-60 production in the BN-350 fast power reactor

    International Nuclear Information System (INIS)

    Zvonarev, A.V.; Korobejnikov, V.V.; Matveenko, I.P.

    1994-01-01

    A possibility of Co-60 isotope production in the BN-350 fast reactor was considered. A special irradiating device, which is an assembly with a central hole, where a container containing cobalt and zirconium hydride is placed. The irradiating device tested permits generating 60 Co with specific activity of 100 Ci/g

  14. Fast breeder reactor research

    International Nuclear Information System (INIS)

    1975-01-01

    , Italy, in April or May 1977. Recognizing the importance of international co-ope ration within the framework of IWGFR for preparing surveys, proposals and recommendations concerning sodium cooled fast breeder reactors, the Working Group prepared a number of joint documents with the help of experts from the participating countries, discussed them at the Eighth Annual Meeting and made recommendations on the preparation of subsequent joint documents. (author)

  15. 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

  16. Nuclear power in Kazakhstan and current status of the BN-350 fast reactor

    International Nuclear Information System (INIS)

    Krechetov, S.

    1998-01-01

    Atomic scientific-industrial complex of Republic of Kazakhstan consist of: Number of natural uranium mines, two plants of U 3 O 8 production at Aktau and Stepnogorsk towns, metallurgical plant producing fuel pellets for RBMK and WWER fuel assemblies. Fast breeder reactor with sodium coolant BN - 350 at Aktau. The average share of BN-350 in total electricity production is 0.7%. Taking into account common condition industrial in Kazakhstan have no significant improvement the total electricity production on goal and oil station stayed on the same level as in 1996. According to government decision in 1998 the following structure of atomic complex have been established. Several rather serious events should be mentioned. In January 1998 the Provision of licensing in nuclear field was signed by Prime Ministry and now Kazakhstan have all necessary acts for starting this process. In April 1998 the General Program of development atomic scientific and industrial complex of Kazakhstan had been reported to Government and got approval in whole. In particular this program are including the design and construction NPP for electricity production on the lake Balhash, and two NPP for heating Almaty and new capital Akmola. In April 1998 the law on Radiation protection had got approval of Parliament and now President should sign it. In January the Nuclear Technologies Safety Center (NTSC) had been established by group of organizations such as KAEA, NNC, University, Nuclear Society of Kazakhstan, Center of standardization and Almaty local administration. NTSC have established as a society independent experts in the field nuclear safety. With cooperation with ANL an expertise on nuclear safety of BN-350 will be done related to long-term spent fuel storage

  17. Fast breeder reactor

    International Nuclear Information System (INIS)

    Ito, Shin-ichi; Maki, Koichi.

    1975-01-01

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

  18. UK contributions to the decommissioning of the BN-350 reactor in Kazakhstan: 2002 – 2011

    International Nuclear Information System (INIS)

    Wells, D.

    2011-01-01

    UK assistance with the decommissioning of BN-350 has cost ~£8.9 million over ten years, ~£4 million spent directly in Kazakhstan. The Programme has immobilised key wastes, contributed to irreversible shutdown of the reactor and addressed issues associated with sodium coolant processing. The Programme funded the operations to load spent fuel canisters into casks at BN-350, together with their despatch from site and receipt at the secure storage facility. The Programme also delivered technical and project management training, assisted in the production of the BN-350 Decommissioning Plan and contributed to the radiation survey effort in the STS

  19. 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

  20. 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

  1. Universal Fast Breeder Reactor Subassembly Counter manual

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-08-01

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

  2. 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,

  3. 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

  4. Fast breeder reactor at Kalkar. Pt. 2

    International Nuclear Information System (INIS)

    Degen, G.

    1979-02-01

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

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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)

  10. An experience of cleaning and decontamination of the BN-350 reactor components

    International Nuclear Information System (INIS)

    Vasilenko, K.T.; Kochetkov, L.A.; Arkhipov, V.M.; Baklushin, R.P.; Gorlov, A.I.; Kiselev, G.V.; Rezinkin, P.S.; Samarkin, A.A.; Tverdovsky, N.D.

    1978-01-01

    In the course of start-up, adjustment and operation of the BN-350 reactor there arose a need for cleaning from sodium and decontamination of primary and secondary equipment components. Design schemes of the systems provided for this purpose as well as those specially designed for cleaning of steam generator evaporators are considered. Technological processes of cleaning and decontamination for some reactor components (removable parts of circulating pumps, evaporators, valves) are described, the results are presented. (author)

  11. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Gatley, J.A.

    1979-01-01

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

  12. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Thatcher, G.; Mitchell, A.J.

    1981-01-01

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

  13. Fast breeder reactor electromagnetic pump

    International Nuclear Information System (INIS)

    Araseki, Hideo; Murakami, Takahiro

    2008-01-01

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

  14. 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

  15. 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)

  16. 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

  17. 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

  18. Chemical surveillance of commercial fast breeder reactors

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  19. Immobilization of Cesium Traps from the BN-350 Fast Reactor (Aktau, Kazakhstan)

    Energy Technology Data Exchange (ETDEWEB)

    J. A. Michelbacher; C. Knight; O. G. Romanenko; I. L. Tazhibaeva; I. L. Yakovlev; A. V. Rovneyko; V. I. Maev; D. Wells; A. Herrick

    2011-03-01

    During BN-350 reactor operations and also during the initial stages of decommissioning, cesium traps were used to decontaminate the reactor’s primary sodium coolant. Two different types of carbon-based trap were used – the MAVR series, low ash granulated graphite adsorber (LAG) contained in a carrier designed to be inserted into the reactor core during shutdown; and a series of ex-reactor trap accumulators(TAs) which used reticulated vitreous carbon (RVC) to reduce Cs-137 levels in the sodium after final reactor shutdown. In total four MAVRs and seven TAs were used at BN-350 to remove an estimated cumulative 755 TBq of cesium. The traps, which also contain residual sodium, need to be immobilized in an appropriate way to allow them to be consigned as waste packages for long term storage and, ultimately, disposal. The present paper reports on the current status of the implementation phase, with particular reference to the work done to date on the trap accumulators, which have the most similarity with the cesium traps used at other reactors.

  20. 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)

  1. 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

  2. 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.)

  3. Superalloy applications in the fast breeder reactor

    International Nuclear Information System (INIS)

    Powell, R.W.

    1976-01-01

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

  4. The economics of fast breeder reactors

    International Nuclear Information System (INIS)

    Rapin, M.

    1990-01-01

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

  5. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Duncombe, E.; Thatcher, G.

    1979-01-01

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

  6. 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.)

  7. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

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

    1984-11-01

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

  8. 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

  9. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1976-01-01

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

  10. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Durston, J.G.

    1976-01-01

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

  11. BN-350 nuclear power plant. Regulatory aspects of decommissioning

    International Nuclear Information System (INIS)

    Shiganakov, S.; Zhantikin, T.; Kim, A.

    2002-01-01

    Full text: The BN-350 reactor is a fast breeder reactor using liquid sodium as a coolant [1]. This reactor was commissioned in 1973 and operated for its design life of 20 years. Thereafter, it was operated on the basis of annual licenses, and the final shutdown was initially planned in 2003. In 1999, however, the Government of the Republic of Kazakhstan adopted Decree on the Decommissioning of BN-350 Reactor. This Decree establishes the conception of the reactor plant decommissioning. The conception envisages three stages of decommissioning. The first stage of decommissioning aims at putting the installation into a state of long term safe enclosure. The main goal is an achievement of nuclear-and radiation-safe condition and industrial safety level. The completion criteria for the stage are as follows: spent fuel is removed and placed in long term storage; radioactive liquid metal coolant is drained from the reactor and processed; liquid and solid radioactive wastes are reprocessed and long-term stored; systems and equipment, that are decommissioned at the moment of reactor safe store, are disassembled; radiation monitoring of the reactor building and environment is provided. The completion criteria of the second stage are as follows: 50 years is up; a decision about beginning of works by realization of dismantling and burial design is accepted. The goal of the third stage is partial or total dismantling of equipment, buildings and structure and burial. Since the decision on the decommissioning of BN-350 Reactor Facility was accepted before end of scheduled service life (2003), to this moment 'The Decommissioning Plan' (which in Kazakhstan is called 'Design of BN-350 reactor Decommission') was not worked out. For realization of the Governmental Decree and for determination of activities by the reactor safety provision and for preparation of its decommission for the period till Design approval the following documents were developed: 1. Special Technical Requirements

  12. 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

  13. 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

  14. Development of devices for handling with BN-350 radioactive waste

    International Nuclear Information System (INIS)

    Iksanov, A.G.; Pustobaev, S.N.; Shirobokov, Yu.P.; Pugachyev, G.P.; Baldov, A.N.; Tikhomirov, L.N.; Tkachenko, V.V.; Tazhibayeva, I.L.; Klepikov, A.Kh.; Romanenko, O.G.; Kenzhin, E.A.; Yakovlev, V.V.; Khametov, S.; Kalinkin, V.L.; Skvortsov, A.I.; Dmitriev, S.A.; Arustamov, A.E.; Zelenski, D.I.; Serebrennikov, Yu.A.

    2010-01-01

    The package of activity performed proves the correctness of the concept accepted by the Government of the Republic of Kazakhstan on the BN-350 decommissioning (three successive steps above) targeted at minimization of cost, exposure and amount of radioactive waste. Decommissioning of the high power fast breeder reactor plant is carried out for the first time and therefore the normative documents and design decisions elaborated, accepted technologies and estimation of capital expenditure and maintenance costs may enrich the database and serve as orientation for decommissioning of similar units. According to the concept accepted the BN-350 decommissioning is the process of top level of complexity that is characterized with the requirement of concurrent execution of a large scope of work by means of international teams from Kazakhstan, Russia, USA, EC, etc. Such approach needs the creation of modern effective organization schemes of interfaces and management of the Projects and will be further used in other complicated Projects

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

    International Nuclear Information System (INIS)

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

    1984-04-01

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

  16. Symposium on key questions about the fast breeder reactor

    International Nuclear Information System (INIS)

    1975-01-01

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

  17. Towards commercial fast breeder reactors the first 1200 MWe unit

    International Nuclear Information System (INIS)

    Banal, M.; Carle, R.

    The public probably thinks of these fast breeder reactors in terms of their rising unit capacity: RAPSODIE 20 MW (thermal), raised to 40 MW, PHENIX 25 MWe, and now 1200 MWe. However, the purposes of the project and the framework of construction have been fundamentally different in each case. Design parameters and the development program of the LMFBR are presented. (auth)

  18. 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.)

  19. 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.)

  20. Study of short-time mechanical properties changes for BN-350 reactor spent fuel assemblies jacket material from vacancy swelling

    International Nuclear Information System (INIS)

    Karaulov, V.N.; Blynskij, A.P.; Yakovlev, I.L.; Golovin, S.V.; Lambert, D.

    1999-01-01

    Variations of mechanical properties (ultimate strength and limit of plasticity) for irradiated stainless steels, materials of BN-350 reactor cased fuel assemblies tubes, namely: 12X18H10T MTO, 08X16H11M3 MTO, 10X17H13M2T, 12X13M2BRF from vacancy swelling and neutron damaging doze have been studied. Flat samples cut out from hexagonal fuel assemblies casing were tested. The data on casing profilometry, and also the results from hydrostatic weighing of steel samples, were used to evaluate swelling. All measurements and testing were made at temperature 25 degrees C

  1. 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

  2. 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

  3. 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

  4. 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

  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. 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)

  7. 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

  8. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    1991-09-01

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

  9. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    1990-09-01

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

  10. Seismic analysis of fast breeder reactor block

    International Nuclear Information System (INIS)

    Gantenbein, F.

    1990-01-01

    Seismic analysis of LMFBR reactor block is complex due mainly to the fluid structure interaction and the 3D geometry of the structure. Analytical methods which have been developed for this analysis will be briefly described in the paper and applications to a geometry similar to SPX1 will be shown

  11. 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

  12. 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.

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

    International Nuclear Information System (INIS)

    Stojadinovic, Timm

    1991-08-01

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

  14. Heterogeneous cores for fast breeder reactor

    International Nuclear Information System (INIS)

    Schroeder, R.; Spenke, H.

    1980-01-01

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

  15. 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

  16. 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

  17. 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)

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

    International Nuclear Information System (INIS)

    Martin Deidier, Loick.

    1979-12-01

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

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

    International Nuclear Information System (INIS)

    Gho, C.J.

    1985-01-01

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

  20. The use of waveguide acoustic probes for void fraction measurement in the evaporator of BN-350-Type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Melnikov, V.I.; Nigmatulin, B.I.

    1995-09-01

    The present paper deals with some results of the experimental studies which have been carried out to investigate the steam generation dynamics in the Field tubes of sodium-water evaporators used in the BN-350 reactors. The void fraction measurements have been taken with the aid of waveguide acoustic transducers manufactured in accordance with a specially designed technology (waveguide acoustic transducers-WAT technology). Presented in this paper also the transducer design and calibration methods, as well as the diagram showing transducers arrengment in the evaporator. The transducers under test featured a waveguide of about 4 m in length and a 200-mm long sensitive element (probe). Besides, this paper specifies the void fraction data obtained through measurements in diverse points of the evaporator. The studies revealed that the period of observed fluctuations in the void fraction amounted to few seconds and was largely dependent on the level of water in the evaporator.

  1. Refueling system for the gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Hawke, B.C.

    1980-05-01

    Criteria specifically related to the handling of Gas-Cooled Fast Breeder Reactor (GCFR) fuel are briefly reviewed, and the most significant requirements with which the refueling system must comply are discussed. Each component of the refueling system is identified, and a functional description of the fuel handling machine is presented. An illustrated operating sequence describing the various functions involved in a typical refueling cycle is presented. The design status of components and subsystems selected for conceptual development is reviewed, and anticipated refueling time frames are given

  2. Inherent safe fast breeder reactors and actinide burners, metallic fuel

    International Nuclear Information System (INIS)

    Dorner, S.; Schumacher, G.

    1991-04-01

    Nuclear power without breeder strategy uses the possibilities for the energy supply only to a small extend compared to the possibilities of fast breeder reactors, which offer an energy supply for thousands of years. Moreover, a fast neutron device offers the opportunity to run an actinide-burner that could improve the situation of waste management. Within this concept metallic fuel could play a key role. The present report shows some important aspects of the concept like the pyrometallic reprocessing, the behaviour of metallic fuel during a core meltdown accident and others. The report should contribute to the discussion of these problems and initialize further work

  3. Steam generator tubing development for commercial fast breeder reactors

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  4. Analysis of fuel sodium interaction in a fast breeder reactor

    International Nuclear Information System (INIS)

    Tezuka, M.; Suzuki, K.; Sasanuma, K.; Nagasima, K.; Kawaguchi, O.

    A code ''SUGAR'' has been developed to evaluate molten Fuel Sodium Interaction (FSI) in a fast breeder reactor. This code computes thermohydrodynamic behavior by heat transfer from fuel to sodium and dynamic deformation of reactor structures simultaneously. It was applied to evaluate FSI in local fuel melting accident in a fuel assembly and in core disassembly accident for the 300MWe fast breeder reactor under development in Japan. The analytical methods of the SUGAR code are mainly shown in the following: 1) the thermal and dynamic model of FSI is mainly based on Cho-Wright's model; 2) the axial and radial expansions of surroundings of FSI region are calculated with one-dimensional and compressive hydrodynamics equation; 3) the structure response is calculated with one-dimensional and dynamic stress equation. Our studies show that mass of fuel interacted with sodium, ratio of fuel mass to sodium mass, fuel particle size, heat transfer coefficient from fuel to sodium, and structure's force have great effect on pressure amplitude and deformation of reactor structures

  5. Fabrication of metallic fuel for fast breeder reactor

    International Nuclear Information System (INIS)

    Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arbind; Mittal, R.K.; Prasad, R.S.; Mahule, N.; Kumar, Arun; Prasad, G.J.

    2012-01-01

    Natural uranium oxide fuelled PHWRs comprises of first stage of Indian nuclear power programme. Liquid metal fast breeder reactors fuelled by Pu (from PHWR's) form the second stage. A shorter reactor doubling time is essential in order to accelerate the nuclear power growth in India. Metallic fuels are known to provide shorter doubling times, necessitating to be used as driver fuel for fast breeder reactors. One of the fabrication routes for metallic fuels having random grain orientation, is injection casting technique. The technique finds its basis in an elementary physical concept - the possibility of supporting a liquid column within a tube, by the application of a pressure difference across the liquid interface inside and outside the tube. At AFD, BARC a facility has been set-up for injection casting of uranium rods in quartz tube moulds, demoulding of cast rods, end-shearing of rods and an automated inspection system for inspection of fuel rods with respect to mass, length, diameter and diameter variation along the length and internal and external porosities/voids. All the above facilities have been set-up in glove boxes and have successfully been used for fabrication of uranium bearing fuel rods. The facility has been designed for fabrication and inspection of Pu-bearing metallic fuels also, if required

  6. Investigation of base isolation for fast breeder reactor building

    International Nuclear Information System (INIS)

    Morishita, M.; Kobatake, M.; Ohta, K.; Okada, Y.

    1989-01-01

    Achievement of great rationalization for seismic-resistant design of equipment system is necessary and indispensable from the viewpoints of economical and structural validity for a fast breeder reactor to be made practical. The method of reducing seismic loads on the building and equipment by application of base isolation may be an effective method, but in application to nuclear facilities, it will become necessary to examine the feasibility to actual design considering the severe seismic design requirements in Japan. With these considerations as the background, the authors carried out analytical studies from various viewpoints such as restoring force characteristics of base isolation device, influence of input earthquake motion, soil-structure interaction in base- isolated structure, etc. in case of providing base isolation system for a fast breeder reactor building. Based on these analytical studies, vibration tests on a base-isolated structure using a triaxial shaking table and simulation analyses of the tests were performed attempting to verify the effectiveness of the base isolation system and appropriateness of the analysis method. Results are presented

  7. Nuclear instrumentation systems in prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Vijayakumaran, P.M.; Nagaraj, C.P.; Paramasivan-Pillai, C.; Ramakrishnan, R.; Sivaramakrishna, M.

    2004-01-01

    The nuclear instrumentation systems of the Prototype Fast Breeder Reactor (PFBR) primarily comprise of global Neutron Flux Monitoring, Failed Fuel Detection and Location, Radiation Monitoring and Post-Accident Monitoring. High temperature fission chambers are provided at in-vessel locations for monitoring neutron flux. Failed fuel detection and location is by monitoring the cover gas for fission gases and primary sodium for delayed neutrons. Signals of the core monitoring detectors are used to initiate SCRAM (safety action) to protect the reactor from various postulated initiating events. Radiation levels in all potentially radioactive areas are monitored to act as an early warning system to keep the release of radioactivity to the environment and exposure to personnel well below the permissible limits. Fission Chambers and Gamma Ionisation Chambers are located in the reactor vault concrete for monitoring the neutron flux and gamma radiation levels during and after an accident. (authors)

  8. Characterizing the tribological behaviour of fast breeder reactor materials

    International Nuclear Information System (INIS)

    Depierre, J.; Raffailhac, J.

    1984-04-01

    The object of these tests is to define the behaviour of material couples working in conditions as representative as possible of reactor operation. For this purpose a certain number of test installations have been developed to simulate the most typical cases of friction encountered: plane to plane geometry, rotational bearings, guiding bearings. Endurance tests have also been carried out on ball bearings and ballscrews samples. As said before, the test conditions attempt to reproduce as faithfully as possible the environment of the materials used in fast breeder reactors, particularly in: - using purified liquid sodium, and maintaining it isotherm, respectively at three temperature levels: 180, 400 and 550 0 C; - or using argon containing sodium aerosol particles. Some typical values of friction coefficients and rates of wear obtained during the tests with certain couples of materials are given here as examples. The aims which are currently guiding the direction of the tests are also briefly described

  9. Fast breeder reactors secondary piping potential sodium leakage rate assessment

    International Nuclear Information System (INIS)

    Alicino, F.; Cardini, S.

    1989-01-01

    In the liquid metal fast breeder reactors (LMFBRs) it must always be taken under control any possible air-sodium contact, because of the elevated air-sodium reactivity. This requires that LMFBRs be carefully designed so that over the entire plant life such an event can't occur in an uncontrolled way. For these reactors the operating conditions usually impose that a lot of life be spent in the creep regime and moreover generally severe hot and cold thermal transients are anticipated, which increases the potential of crack propagation. Then, a useful means to ascertain if this event can occur is to adopt a fracture mechanics approach. This paper presents a computer program to perform fracture mechanics calculations

  10. Raw materials problems in connection with fast breeder type reactors

    International Nuclear Information System (INIS)

    Hirsch, H.; Kreusch, J.

    1981-01-01

    The power supply by the FBR type reactors which depends upon the availability of essential raw materials such as Cr and Mo for structural and special steels is supposed to be less ensured than supply by fossil-fueled power plants. This contribution tries to verify this statement by means of estimates of the annual Cr and Mo demand, of the resources, production and consumption as well as by a study of the possibilities of recycling and substituting Cr and Mo. The only realistic alternative to the fast breeder type reactor is supposed to be a soft path of development according to the principle of decentralization, utilization of renewable energy sources regard to environmental protection, and use of less sophisticated technology. (DG) [de

  11. 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.)

  12. Materials requirements for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  13. 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)

  14. 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)

  15. Experience and results of material science research conducted on spent fuel assemblies from the BN-350 fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maksimkin, O.; Gusev, M.; Turubarova, L.G.; Tsai, K.V.; Yarovchuk, A.V. [Institute of Nuclear Physics, Almaty (Kazakhstan)

    2007-07-01

    Full text of publication follows: The BN-350 fast reactor was commissioned in 1973, ran successfully for many years and is now in the decommission stage. Its unique operational parameters (low temperature of sodium at the input, wide range of damage rates, etc. ) allowed the investigation of a number of new radiation effects on both austenitic and ferritic-martensitic steels. The latter class of steel was extensively employed as wrappers for fuel assemblies. Much of the accumulated experience in BN-350 is relevant to development of fusion devices. Results are presented on post-operational research of steels 12Cr18Ni10Ti, 08Cr16Ni11Mo3, and 12Cr13Mo2BFR, all serving as hexagonal shrouds of fuel assemblies. Structural materials in the active core zone operated at temperatures of 280-430 deg. C, and were irradiated the range of 0.25-83 dpa with damage rates of 10{sup -9} - 10{sup -6} dpa/s). Investigations of irradiated hexagonal shroud materials were performed with using traditional techniques of transmission and scanning electron microscopy, metallography, mechanical tests, hydrostatic weighing, magnetometry, etc. Additionally, new techniques have been developed and employed with great success on these highly irradiated materials, such as optical computer extensometry, and magnetization cartography. Typical results to be covered in this presentation are: a) In 12Cr18Ni10Ti steel irradiated at a low dose rate of 0.12 x 10{sup -8} dpa/s voids were found at 281 deg. C after only 0.65 dpa, demonstrating once again the acceleration of swelling at low dpa rates observed in other steels. b) Data on helium release during annealing of highly irradiated sample are presented. c) Differences in deformation-induced hardening between the shroud's corners and faces leads to post-irradiation differences in swelling and mechanical properties. d) During room temperature mechanical tests of 12Cr18Ni10Ti steel at {approx}56 dpa at 350 deg. C it was found that ductility lost at

  16. Decontamination before dismantling a fast breeder reactor primary cooling system

    International Nuclear Information System (INIS)

    Costes, J.R.; Antoine, P.; Gauchon, J.P.

    1997-01-01

    The large-scale decontamination of FBR sodium loops is a novel task, as only a limited number of laboratory-scale results are available to date. The principal objective of this work is to develop a suitable decontamination procedure for application to the primary loops of the RAPSODIE fast breeder reactor as part of decommissioning to Stage 2. After disconnecting the piping from the main vessel, the pipes were treated by circulating chemical solutions and the vessels by spraying. The dose rate in the areas to be dismantled was divided by ten. A decontamination factor of about 300 was obtained, and should allow austenitic steel parts to be melted in special furnaces for unrestricted release. (author)

  17. Radioactive waste management at a Liquid Metal Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Abrams, C.S.; Fryer, R.H.; Witbeck, L.C.

    1979-01-01

    This paper presents the radioactive waste production and management at a Liquid Metal Fast Breeder Reactor-II (EBR-II), which is operated for the US Department of Energy by the Argonne National Laboratory at the Idaho National Engineering Laboratory (INEL). Since this facility, in addition to supplying power has been used to demonstrate the breeder, fuel cycling, and recently operations with defective fuel elements, various categories of waste have been handled safely over some 14 years of operation. Liquid wastes are processed such that the resulting effluent can be discharged to an uncontrolled area. Solid wastes up to 10,000 R/hr are packaged and shipped contamination-free to a disposal site or interim storage with exposures to personnel approximately 10 mrem. Gaseous waste discharges are low such as 143 Ci of noble gases in 1978 and do not have a significant effect on the environment even with operations with breached fuel

  18. IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

    International Nuclear Information System (INIS)

    Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

    2002-01-01

    Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

  19. 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)

  20. Analysis of a sustainable gas cooled fast breeder reactor concept

    International Nuclear Information System (INIS)

    Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

    2014-01-01

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a

  1. 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

  2. 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)

  3. 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

  4. The fast breeder reactor: what is security and freedom worth to us

    International Nuclear Information System (INIS)

    1975-01-01

    Risks for governmental and non-governmental proliferation of nuclear weapons as a consequence of the development of fast breeder reactor technology are discussed. Polemological as well as sociological and political analyses together with ecological consequences lead to the conclusion that the fast breeder is not necessary for energy production and socially unacceptable

  5. Development of high performance core for large fast breeder reactors

    International Nuclear Information System (INIS)

    Inoue, Kotaro; Kawashima, Katsuyuki; Watari, Yoshio.

    1982-01-01

    Subsequently to the fast breeder prototype reactor ''Monju'', the construction of a demonstration reactor with 1000 MWe output is planned. This research aims at the establishment of the concept of a large core with excellent fuel breeding property and safety for a demonstration and commercial reactors. For the purpose, the optimum specification of fuel design as a large core was clarified, and the new construction of a core was examined, in which a disk-shaped blanket with thin peripheral edge is introduced at the center of a core. As the result, such prospect was obtained that the time for fuel doubling would be 1/2, and the energy generated in a core collapse accident would be about 1/5 as compared with a large core using the same fuel as ''Monju''. Generally, as a core is enlarged, the rate of breeding lowers. If a worst core collapse accident occurs, the scale of accident will be very large in the case of a ''Monju'' type large core. In an unhomogeneous core, an internal blanket is provided in the core for the purpose of improving the breeding property and safety. Hitachi Ltd. developed the concept of a large core unhomogeneous in axial direction and proposed it. The research on the fuel design for a large core, an unhomogeneous core and its core collapse accident is reported. (Kako, I.)

  6. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1976-05-01

    Following the completion of the in-argon, high temperature test, the in-sodium functional test of Joyo has set in. The fabrication of the equipments for monitoring the flow rate and temperature in the center channel and the power distribution was finished. The modification of design of the prototype fast breeder reactor Monju came into its phase 3. The interim report on the check-up and review of the Monju project by the Government is now ready. Various calculation codes were developed or are in development stage. The mock-up assembly FCA 7-1 has been built, which consists of the Pu-fueled sector region simulating the Monju core and the U-235-fueled driver region. Various reactor physics experiments have been carried out in this assembly. Also, the calculation methods for reactor physics parameters have been developed, and the detailed calculation on the main shield of Joyo was performed. The situation of the developments of the components for Joyo and Monju and the measuring and control systems is shown. Almost all the existing sodium test facilities in Oarai Engineering Center were in service without any trouble, and the new test facility named ''Carbon transfer test loop'' was commissioned. The progress in the fields of sodium technology, fuel and material, safety and steam generators is reported. (Kako, I.)

  7. Alternative fuels for the French fast breeder reactors programme

    International Nuclear Information System (INIS)

    Bailly, H.; Bernard, H.; Mansard, B.

    1989-01-01

    French fast breeder reactors use mixed oxide as reference fuel. A great deal of experience has been gained in the behaviour and manufacture of oxide fuel, which has proved to be the most suitable fuel for future commercial breeder reactors. However, France is maintaining long-term alternative fuels programme, in order to be in a position to satisfy eventually new future reactor design and operational requirements. Initially, the CEA in France developed a carbide-based, sodium-bonded fuel designed for a high specific power. The new objective of the alternative fuels programme is to define a fuel which could replace the oxide without requiring any significant changes to the operating conditions, fuel cycle processes or facilities. The current program concentrates on a nitride-based, helium-bonded fuel, bearing in mind the carbide solution. The paper describes the main characteristics required, the manufacturing process as developed, the inspection methods, and the results obtained. Present indications are that the industrial manufacture of mixed nitride is feasible and that production costs for nitride and oxide fuels would be not significantly different. (author) 8 refs., 2 figs

  8. Study on hydrogen production using the fast breeder reactors (FBR)

    International Nuclear Information System (INIS)

    Kani, Yoshio

    2003-01-01

    As the fast breeder reactor (FBR) can effectively convert uranium-238 difficult to carry out nuclear fission at thermal neutron reactors to nuclear fissionable plutonium-239 to use it remarkable upgrading of application on uranium can be performed, to be expected for sustainable energy source. And, by reuse minor actinides of long half-life nuclides in reprocessed high level wasted solutions for fuels of nuclear reactors, reduction of radioactive poison based on high level radioactive wastes was enabled. As high temperature of about 800 centigrade was required on conventional hydrogen production, by new hydrogen production technique even at operation temperature of sodium-cooled FBR it can be enabled. Here were described for new hydrogen production methods applicable to FBR on palladium membrane hydrogen separation method carrying out natural gas/steam modification at reaction temperature of about 500 centigrade, low temperature thermo-chemical method expectable simultaneous simplification of production process, and electrolysis method expected on power load balancing. (G.K.)

  9. 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)

  10. Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

    International Nuclear Information System (INIS)

    Porollo, S.I.; Konobeev, Y.V.; Ivanov, A.A.; Shulepin, S.V.; Garner, F.

    2007-01-01

    Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

  11. Plant dynamics and safety analysis of fast breeder reactors

    International Nuclear Information System (INIS)

    Ertel, V.

    1982-01-01

    Some general features of sodium cooled fast breeders which influence the thermohydraulics and differ from LWR'S are discussed. Using the SNR-300 as a reference, some thermohydraulic transients from normal operation and from design accidents are presented. (orig.)

  12. Possible role of fast breeder reactor for global warming mitigation

    International Nuclear Information System (INIS)

    Ikeda, Kazumi; Hatano, Mamoru; Tashimo, Masanori; Sekimoto, Hiroshi

    2003-01-01

    Our optimization model forecasts that Fast Breeder Reactor (FBR) could be introduced as early as 2030 when nuclear power generation becomes price-competitive in the energy market. If such a scenario realizes in the future, nuclear power will be able to supply as much as 30% of total primary energy in 2100. The model also forecasts that nuclear power will contribute to reduce 6 GtC of CO 2 emission, if it completely replaces coal of the same energy. If the option of FBR is abandoned, we expect that carbon tax almost doubles in 2100 from 280 $/tC to 540 $/tC. High-burnup of LWR decreases the initial capacity of FBR plants because FBR and its fuel cycle need about seven tons of plutonium per GWe. Thus, the high-burnup of Light Water Reactor (LWR) would decrease the production of plutonium per watt-hour of electricity generation and reduce the capacity of FBR plants in the near future. However, the introduction of high breeding FBR can solve the above deficiency problem of plutonium. In other words, high breeding FBR can not only work as a catalyst for increasing the FBR plants with improved nuclear energy technology, but also contribute to reduce the greenhouse gas emission since renewable energy may be not enough in Japan. (author)

  13. Instrumentation and control of the fast breeder reactors

    International Nuclear Information System (INIS)

    Juengst, U.

    1982-01-01

    Generally the Liquid Metal Fast Breeder Reactor (LMFBR) in comparison to the Leight Water Reactors hasn't exceptionally different requirements to the instrumentation and control systems. There fore the paper restricts itself to outline the peculiarities of LMFBR's exemplified at the design of the German Breeder SNR-300 beeing under construction. The special sodium instrumentation for temperature level, flow and pressure in the main systems is described. The automatic control system enabled the plant to follow the bad demand of the grid immidiately. The nuclear power production itself is stabilized by inherent coefficients of the core. An exceptional high reliability of the shut down systems is necessary; there fore all LMFBR's have two completely independant and diverse shut down systems. In the SNR-300 also the control of decay heat removal and the active safety systems for enclosure of radioactivity are distributed to two independent and diverse plant protection systems (PPS) so that the first system covered the accidents due to internal events and the second systems managed the external events such as earthquake and aircraft crash. (orig.)

  14. Past and present role of fast breeder reactors in Japan

    International Nuclear Information System (INIS)

    Tomabechi, K.

    1978-01-01

    To meet increasing future energy demand in Japan utilization of fission energy should be promoted. In particular it is of vital importance to develop and utilize FBRs as soon as possible in order to save the natural uranium needed. If one considers the commercial introduction of FBRs in the mid-1990s in Japan, a delay of only one year will eventually result in an additional demand for natural uranium of more than 20,000 tonnes, because several LWRs will have to be installed instead. Ten years have passed since the development of FBRs in Japan was initiated as a national project with the highest priority and now the experimental fast reactor JOYO is successfully being commissioned and the prototype fast breeder reactor MONJU has reached the stage of proceeding to construction. Experience gained from past development work encourages further strong effort towards the commercialization of FBRs. This paper briefly reviews the important role to be played by FBRs, the development programme for FBRs, and the experience gained so far from the programme in Japan. (author)

  15. Flow induced vibrations in liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    1989-01-01

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

  16. 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)

  17. Core disruptive accident analysis in prototype fast breeder reactor

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. Improved alloys for a liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    1981-01-01

    An alloy is specified suitable for use at elevated temperatures and especially in a liquid metal fast breeder reactor consisting essentially of a nickel-chromium steel having a specified range of composition of C, Mn, Si, Zr, V, Ni, Cr, Ti, Al, Mo, B, and the balance iron with incidental impurities, the alloy exhibiting a swelling at peak swelling temperature of less than 10% wherein the matrix composition has after heat treatment at a temperature within the range of 1000 0 C to 1100 0 C for about one half hour followed by aging at a temperature within the range of from 700 0 C to 815 0 C for a time period of between 10 to 24 hours, the longer hours being associated with the lower temperatures and vice-versa, and after the removal of the non-equilibrium gamma prime and other precipitated phases a composition within a specified range of composition of Ni, Cr, Ti, Al, Mo, the balance being essentially iron. (U.K.)

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

    International Nuclear Information System (INIS)

    1988-01-01

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

  20. Method and device for fire extinction of fast breeder reactors

    International Nuclear Information System (INIS)

    Yokota, Norikatsu; Shimoyashiki, Shigehiro; Hikichi, Takayoshi; Sato, Yoshihiko.

    1986-01-01

    Purpose: To effectively restrain fires with coolant in liquid-metal fast breeder reactors. Method: The core material of fire-extinguishing agent is coated with a non-combustible material and capsulated to prevent moisture absorption and at the same time the capsule thus made is coated with a suitable material to restrain a fire with a coolant. A desirable coating material to be used is a material which is little reactive to sodium; for example such a low-melting point metal as Pb or Sn, or paraffin, or sodium-silicate should be used. For the core material, Na 2 CO 3 , NaCl sand are recommendable materials. The core material thus made will never absorb moisture during long-time storage and has no hazard to promote a fire likely to be caused by moisture absorption. Furthermore, the coating material and the core material act to each other, restraining a sodium fire. The fire-extinguishing agent, being granular and capsulated, is easy to transport, thereby reducing a cost required for disposition. (Kamimura, M.)

  1. Status of liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    1985-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Horton, K [U.S. Department of Energy, Washington, DC (United States)

    1981-05-01

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

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

    International Nuclear Information System (INIS)

    Horton, K.

    1981-01-01

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

  4. Multiple recycling of fuel in prototype fast breeder reactor in a closed ...

    Indian Academy of Sciences (India)

    Our previous study in this regard for the prototype fast breeder reactor ... This study aims at finding the feasibility of multiple recycling of PFBR fuel with external ...... maximum allowable Pu content in fuel based on chemistry/metallurgical ...

  5. Dispersion strengthened ferritic alloy for use in liquid-metal fast breeder reactors (LMFBRS)

    International Nuclear Information System (INIS)

    Fischer, J.J.

    1978-01-01

    A dispersion-strengthened ferritic alloy is provided which has high-temperature strength and is readily fabricable at ambient temperatures, and which is useful as structural elements of liquid-metal fast breeder reactors. 4 tables

  6. European in-pile investigations on fast breeder reactor safety

    International Nuclear Information System (INIS)

    Bailly, J.; Penet, F.; Teague, H.J.

    1977-01-01

    Because of the difficulties faced by the various organisations responsible for the design and safety analysis of fast reactors, in the conception and execution of perfectly representative experiments corresponding to the various hypothetical accidents considered, the European experts in this field have chosen to adopt a more fundamental approach. By the coordination of the efforts in the different countries and pooling the results, it appears that sufficient data can be assembled to permit them to: isolate the influence of the principle parameters on the course of an accident and to have an appreciation of the possibilities for detection and propagation; and validate calculational models of the essential phenomena, and determine the values of the adjustable parameters used in accident codes. The present paper does not propose to give an exhaustive description of the European experimental programmes, nor of the possibilities for exploiting the results in safety analysis. The aim is simply to show with the aid of results already obtained or expected from the immediate programme, that it should be possible to answer certain fundamental questions concerning those accidents at present considered most important. Particular consideration is given to local blockages (SCARABEE, DFR and MOL 7C programmes), unprotected loss of pumps leading to cooling failure, and inlet blockage (SCARABEE and CABRI), and reactivity excursions (CABRI and VIPER). In addition it is planned to study, as part of the SCARABEE programme, the long-term post-accident behaviour of fuel in some simple configurations. It is shown that these various results should lead to a great improvement in our understanding of the progress of the hypothetical accidents taken into account in the design of fast breeder reactors [fr

  7. Under sodium ultrasonic viewing for Fast Breeder Reactors: a review

    International Nuclear Information System (INIS)

    Tarpara, Eaglekumar G.; Patankar, V.H.; Vijayan Varier, N.

    2016-09-01

    Liquid Metal Fast Breeder Reactors (LMFBR/FBR) are of two types: Loop type and Pool type. Many countries like USA, Japan, UK, Russia, China, France, Lithuania, Belgium, Korea, and India have worked extensively on these types of FBRs. FBRs are capable of breeding more fissionable fuel than it consumes like breeding of Plutonium-239 from non-fissionable Uranium-238. In FBR, heat is released by fission process and it must be captured and transferred to the electric generator by the liquid metal coolant (i.e. Sodium). Due to continuous operation and for safety and licensing reasons, periodic inspection and maintenance is required for reactor fuel assemblies which carry nuclear fuels. For this reason, under sodium ultrasonic imaging technique is adopted as in-service inspection activity for viewing of core of FBRs. Since liquid sodium is optically opaque, ultrasonic technique is the only method which can be employed for imaging in liquid sodium. In harsh conditions like high temperature and high radiation, there is a restriction on the development of possible ultrasonic visualization systems and selection of the transducer materials which can operate in the core region of reactor at around 200 °C during shutdown of reactor. This report provides a review of works related to ultrasonic imaging in sodium, different materials used in high temperature transducer assemblies and their different coupling/bonding techniques to achieve maximum transmission efficiency in high temperature sodium environment. The report also provides the overview of different architectures and imaging methods of transducer array elements which were used in LMFBRs for inspection and visualization of the reactor core sub-assemblies. The report is focused on a review of some possible beam forming techniques which may be used for nuclear applications for high temperature environment. Published information of the different simulation models are also reviewed which can be adopted to simulate the

  8. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder ...

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

    International Nuclear Information System (INIS)

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

    1980-06-01

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

  10. Y2K issues for real time computer systems for fast breeder test reactor

    International Nuclear Information System (INIS)

    Swaminathan, P.

    1999-01-01

    Presentation shows the classification of real time systems related to operation, control and monitoring of the fast breeder test reactor. Software life cycle includes software requirement specification, software design description, coding, commissioning, operation and management. A software scheme in supervisory computer of fast breeder test rector is described with the twenty years of experience in design, development, installation, commissioning, operation and maintenance of computer based supervision control system for nuclear installation with a particular emphasis on solving the Y2K problem

  11. The French liquid metal fast breeder reactor programme

    International Nuclear Information System (INIS)

    Rapin, M.

    1980-01-01

    The strong French LMFBR development and the corresponding success obtained up to now show that there is no technical insuperable barrier to fast breeder construction. This satisfactory evolution is in fact the conjunction of a strong incentive due to the lack of other resources, a firm and permanent stand of the government, and an obstinate effort of all the teams involved in the LMFBR field. The changeover to industrial level should be helped by the simplicity of the French organization for fast breeder. Finally, the development of LMFBR on a larger scale is helped by international agreements through which the present French know-how can be put at the disposal of other partners, and the general knowledge can be improved by setting common R and D programmes. A quite successful example of such agreements is given by the German-French agreement, and we hope that new partners will join us soon. (orig.) [de

  12. Study on advanced structural design for commercialized fast breeder reactors

    International Nuclear Information System (INIS)

    Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai

    2002-01-01

    Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R and D activities for the feasibility studies on commercialized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study. This report contains the results of the research on the structural design technology. The research scope was identified as (1) System Based Code for Integrity, (2) FDS (FBR Design Standard), and (3) Standardization of new material, and the results of this year's studies are summarized as follows. (1) System Based Code for Integrity. Requirements that a structural design standard must fulfill for enhanced reliability and economy were clarified. Based on this, the authors began to develop the system-based code for integrity. A structural reliability based design approach was proposed as a basic concept for an integrated evaluation of structural integrity. A system consisting of a supreme code and partial codes was proposed. Technologies and engineering tools that are necessary to materialize this code were clarified and research and development was begun. (2) FDS(FBR Design Standard). A rational design approach against thermal loads was proposed. Applicable area of inelastic analysis methods was investigated to develop inelastic analysis guide. A new design system which realizes feedback from structural to thermal hydraulic designs was proposed with a total analysis method of thermal hydraulic and mechanical behaviors. (3) Standardization of new material. Current status of development of high-chromium ferritic steels was investigated. Those steels have excellent high temperature strength and thermal properties. The authors proposed material specifications to apply those steels to structures

  13. Study on advanced structural design for commercialized fast breeder reactors

    International Nuclear Information System (INIS)

    Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai

    2003-05-01

    Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R and D activities for the feasibility studies on commercialized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study. This report contains the results of the research on the structural design technology. The research scope was identified as (1) FDS (FBR Design Standard), (2) Standardization of new material, and (3) System Based Code for Integrity, and the results of this year's studies are summarized as follows. (1) FDS (FBR Design Standard). R and D policy of Phase II study considering to the newest needs demanded for structural design were clarified, and R and D items were settled concretely. As for failure criteria, preliminary ratcheting-fatigue tests were conducted with rational methods, and Negligible Creep curves that settle creep designing region rationally were expanded their applying condition. R and D policy and items of guideline for inelastic analysis were clarified, and analyzing methodologies were studied. Summering up exemplification of thermal load prediction methodologies were progressed. To predict thermal striping loads, advanced frequency response function of thermal stress, and fatigue evaluation methodology were studied. (2) Standardization of new material. As for candidate 12-chromium stainless steel (added tungsten, non-added tungsten), that is expected to improve strength of components of commercialized FR, short and medium-term material tests, and creep-fatigue tests at small strain range were conducted. As for above candidate steel, TIG (Tungsten Inert Gas) welding method was recommended. Requirements to standardize new materials and index to select new materials were studied

  14. Study on advanced structural design for commercialized fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Sagayama, Yutaka; Dozaki, Koji; Tanaka, Yoshihiko [Japan Atomic Power Co., Research and Development Department, Tokyo (Japan)

    2002-01-01

    Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R and D activities for the feasibility studies on commercialized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study. This report contains the results of the research on the structural design technology. The research scope was identified as (1) System Based Code for Integrity, (2) FDS (FBR Design Standard), and (3) Standardization of new material, and the results of this year's studies are summarized as follows. (1) System Based Code for Integrity. Requirements that a structural design standard must fulfill for enhanced reliability and economy were clarified. Based on this, the authors began to develop the system-based code for integrity. A structural reliability based design approach was proposed as a basic concept for an integrated evaluation of structural integrity. A system consisting of a supreme code and partial codes was proposed. Technologies and engineering tools that are necessary to materialize this code were clarified and research and development was begun. (2) FDS(FBR Design Standard). A rational design approach against thermal loads was proposed. Applicable area of inelastic analysis methods was investigated to develop inelastic analysis guide. A new design system which realizes feedback from structural to thermal hydraulic designs was proposed with a total analysis method of thermal hydraulic and mechanical behaviors. (3) Standardization of new material. Current status of development of high-chromium ferritic steels was investigated. Those steels have excellent high temperature strength and thermal properties. The authors proposed material specifications to apply those steels to

  15. 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)

  16. 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.)

  17. Reflections on the introduction of fast breeder reactors in the DeBeNeLux states

    International Nuclear Information System (INIS)

    Schroeder, R.; Wagner, J.

    1975-06-01

    This report gives a survey of the impact of introducing sodium-cooled fast breeder reactors in the Federal Republic of Germany and the BeNeLux countries (DeBeNeLux region). The supply situation with respect to electric and thermal energy is studied in particular, together with aspects of economy and environmental impact. The potential and consequences of a breeder economy, the present status and future r+d work are discussed. In addition to sodium-cooled fast breeder reactors with oxide or carbide fuel, alternative solutions are touched: 1) light water and high temperature reactors, 2) helium-cooled fast breeder reactors, 3) geothermal energy, solar energy and fusion energy. (orig.) [de

  18. 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

  19. Econometric modelling of certain nuclear power systems based on thermal and fast breeder reactors

    International Nuclear Information System (INIS)

    Pavelescu, M.; Pioaru, C.; Ursu, I.

    1988-01-01

    Certain known economic analysis models for a LMFBR fast breeder and CANDU thermal solitary reactors are presented, based on the concepts of discounting and levelization. These models are subsequently utilized as a basis for establishing an original model for the econometric analysis of certain thermal reactor systems or/and fast breeder reactors. Case studies are subsequently conducted with the systems: 1-CANDU, 2-LMFBR, 3-CANDU + LMFBR which enables us to draw certain interesting conclusions for a long range nuclear power policy. (author)

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

    International Nuclear Information System (INIS)

    Gho, C.J.

    1985-01-01

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

  1. Current status of feasibility studies on commercialized fuel cycle system for Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Ojima, Hisao; Nagaoki, Yoshihiro

    2000-01-01

    A 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' is underway at the Japan Nuclear Cycle Development Institute (JNC). The study will select the promising concepts with their R and D tasks in order to commercialize the fast breeder reactor (FBR) cycle system. The feasibility studies (F/S) have to present surveyed and screened various relevant technologies, and defined the design requirement of the commercialized fuel cycle system for FBR. The promising technical options are being evaluated and conceptual designs are being examined. At the end of JFY2000, several candidate concepts of the commercialized FBR cycle system will be proposed. (author)

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

    International Nuclear Information System (INIS)

    Ishiguro, Y.

    1986-10-01

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

  3. Approaches to measurement of thermal-hydraulic parameters in liquid-metal-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1983-01-01

    This lecture considers instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, and sodium purity. It is divided into three major parts: (1) measurement requirements for sodium cooled reactor systems, (2) in-core and out-of-core measurements in liquid metal systems, and (3) performance measurements of water steam generators

  4. Risk/benefits of fast breeder reactors societal risks and accident modelling

    International Nuclear Information System (INIS)

    Farmer, F.R.

    1979-01-01

    A special feature of the Fast Breeder Reactor is the possibility of fuel vaporisation, hence accidents may have more severe consequences than thermal reactors. This article discusses the process of accident modelling, the identification and assessment of risk, not yet incurred. 10 refs

  5. Reactor shutdown system of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Govindarajan, S.; Singh, Om Pal; Kasinathan, N.; Paramasivan Pillai, C.; Arul, A.J.; Chetal, S.C.

    2002-01-01

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

  6. 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

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

    International Nuclear Information System (INIS)

    1986-02-01

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

  8. 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

  9. Comparison of material property specifications of ferritic steels in fast-breeder reactor technology

    International Nuclear Information System (INIS)

    Delporte, E.; Vanderborck, Y.

    1988-01-01

    The component fabrications for the fast breeder reactors request the use of ferritic steels specially appropriated for the construction of the equipments sustaining pressure and high temperature. The Activity Group nr 3 Materials of the FRCC has decided to make a study to compare the different norms related to the properties of somme ferritic steels used in the different European fast breeder projects. In particular, this study should allow in the different countries of the Community, to identify the designation of a specific steel and to compare its properties. Deviations between the different norms of a same material are mentioned to facilitate European standardization of this type of material

  10. Correlation of yield stress and microhardness in 08Cr16Ni11Mo3 stainless steel irradiated to high dose in the BN-350 fast reactor

    International Nuclear Information System (INIS)

    Gusev, M.N.; Maksimkin, O.P.; Tivanova, O.V.; Silnaygina, N.S.; Garner, F.A.

    2006-01-01

    The relationship between the microhardness and the engineering yield stress in 08Cr16Ni11Mo3 steel after irradiation in the BN-350 reactor has been experimentally derived and agrees with a previously published correlation developed by Toloczko for unirradiated 316 in a variety of cold-work conditions. Even more importantly, when the correlation is derived in the K Δ format where the correlation involves changes in the two properties, excellent agreement is found with a universal K Δ correlation developed by Busby and coworkers. Additionally, this report points out that microhardness measurements must take into account that sodium exposure at high temperature and neutron fluence alters the metal surface to produce ferrite, and therefore the altered layers should be removed prior to testing

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

    International Nuclear Information System (INIS)

    Abramson, R.

    1984-04-01

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

  12. The European Fast Breeder Reactor project (EFR). Should Germany withdraw from the project?

    International Nuclear Information System (INIS)

    1993-01-01

    The Fed. Min. of Research and Technology has discontinued its financial aid to industry for the development of fast breeder reactor technology at the end of 1991, and a termination of the EFR project is in sight. There is, however, no need for the German government to withdraw from the EFR agreement concluded in 1984. (orig.) [de

  13. French experience and prospects in the reprocessing of fast breeder reactor fuels

    International Nuclear Information System (INIS)

    Megy, J.

    1983-06-01

    Experience acquired in France in the field of reprocessing spent fuels from fast breeder reactors is recalled. Emphasis is put on characteristics and quantities of spent fuels reprocessed in La Hague and Marcoule facilities. Then reprocessing developments with the realisation of the new pilot plant TOR at Marcoule, new equipments and study of industrial reprocessing units are reviewed [fr

  14. Compilation of data and descriptions for United States and foreign liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Appleby, E.R.

    1975-08-01

    This document is a compilation of design and engineering information pertaining to liquid metal cooled fast breeder reactors which have operated, are operating, or are currently under construction, in the United States and abroad. All data has been taken from publicly available documents, journals, and books

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

    International Nuclear Information System (INIS)

    1978-09-01

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

  16. 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

  17. Fuel cycles and advanced core designs for the Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Simon, R.H.; Hamilton, C.J.; Hunter, R.S.

    1982-01-01

    Studies indicate that a 1200 MW(e) Gas-Cooled Fast Breeder Reactor could achieve compound system doubling times of under ten years when using advanced oxide or carbide fuels. In addition, when thorium is used in the breeding blankets, enough U-233 can be generated in each GCFR to supply several advanced converter reactors with fissionable material and this symbiotic relationship could provide energy for the world for centuries. (author)

  18. Recent developments in the analysis of coolant sodium for fast breeder reactors

    International Nuclear Information System (INIS)

    Keough, R.F.; McCown, J.J.

    1976-01-01

    The measurement of impurities in sodium is an important part of the sodium technology for the fast breeder reactor program. A knowledge of impurity levels in sodium provides an important source of information on such things as corrosion rates, air and water leak location and detection, and the effectiveness of the sodium purification systems. A discussion is presented of some of the analytical techniques for sodium characterization in reactor heat transfer systems. Emphasis is placed on some recently developed alternatives to laboratory analysis

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  20. PRISM: An innovative liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Kruger, G.B.; Boardman, C.E.; Olich, E.E.; Switick, D.M.

    1986-01-01

    This paper describes an innovative sodium-cooled reactor concept employing small certified reactor modules coupled with a standardized steam generator system. The total plant employs nine PRISM reactors (power reactor inherently safe module) in three 415 MWe power blocks. The PRISM design concept utilizes inherent safety characteristics and modularity to improve licensability, reduce owner's risk, and reduce costs. The relatively small size of each reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. It is proposed that a single PRISM module be used in a full-scale integrated reactor safety test. Results from the test would be used to obtain NRC certification of the standard design

  1. Tube sheet structural analysis of intermediate heat exchanger for fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    Nakagawa, Y.; Ueno, T.; Fukuda, Y.; Ichimiya, M.

    1983-01-01

    The Prototype Fast Breeder Reactor 'Monju' is the first power generating fast breeder reactor in Japan. We have been designing the components of the plant for manufacturing. Among these is the intermediate heat exchanger (IHX) which exchanges heat between primary and secondary sodium loop. The tube sheet of IHX (shell to ligament junction) is a difficult area from the view point of structural strength design under elevated temperature. To validate the structural integrity of tube sheet we performed the series of inelastic analysis and tube sheet thermal shock test using test pieces and half scale model of actual design. The results of inelastic analyses showed there is little progressive deformation around shell to ligament structural discontinuous junction. Furthermore, thermal shock tests showed no increase of an accumulative deformation. By these analyses and experiments, structural reliability of tube sheet could be shown. (author)

  2. Irradiation Creep and Swelling of Russian Ferritic-Martensitic Steels Irradiated to Very High Exposures in the BN-350 Fast Reactor at 305-335 degrees C

    International Nuclear Information System (INIS)

    Konobeev, Yury V.; Dvoriashin, Alexander M.; Porollo, S.I.; Shulepin, S.V.; Budylkin, N.I.; Mironova, Elena G.; Garner, Francis A.

    2003-01-01

    Russian ferritic/martensitic (F/M) steels EP-450, EP-852 and EP-823 were irradiated in the BN-350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb-Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP-450 and EP-823 at temperatures between 390 and 520C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP-450 and EP-852 at temperatures between 305 and 335C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures <420C, but may be camouflaged somewhat by precipitation-related densification. These irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels.

  3. Mathematical modeling of a fast-breeder-reactor generating unit

    International Nuclear Information System (INIS)

    Kim, V.E.; Golovach, E.A.; Senkin, V.I.

    1984-01-01

    Dynamics equations are given for a reactor, intermediate heat exchanger, steam generator, and turbogenerator. The dynamic characteristics of the generating unit are described when perturbations occur in grid frequency, turbine valves, and feedwater consumption

  4. Exploding the myths about the fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, S.

    1979-01-01

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

  5. Analysis of the formation of local cooling disturbances in sodium-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1976-09-01

    The aim of this analysis of the formation of local cooling disturbances in sodium-cooled fast breeder reactors is to get results on the possible extent of blockages and the time necessary for growth which may be used for a safety evaluation. After an introduction where the thermohydraulic and physical/chemical aspects of the problems are considered, the causes for the local cooling disturbances and the phenomena arising with it are freated in more detail. (orig./TK) [de

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  7. Experience in quality assurance of alloy D9 clad tubes for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Kapoor, K.; Prahlad, B.

    2012-01-01

    Stainless Steel Alloy D9 is the material for cladding in various sub-assemblies of Prototype Fast Breeder Reactor (PFBR). The fabrication, inspection, testing and supply of the clad tubes for the first core of PFBR is nearly completed. The paper also compares the specification requirements and the achieved results for some of the critical aspects which is arrived after completing supply against the first core requirement

  8. Transformation of sodium from the Rapsodie fast breeder reactor into sodium hydroxide

    International Nuclear Information System (INIS)

    Roger, J.; Latge, C.; Rodriguez, G.

    1994-01-01

    One of the major problems raised by decommissioning a fast breeder reactor (FBR) concerns the disposal of the sodium coolant. The Desora operation was undertaken to eliminate the Rapsodie primary sodium as part of the partial decommissioning program, and to develop an operational sodium treatment unit for other needs. The process involves reacting small quantities of sodium in water inside a closed vessel, producing aqueous sodium hydroxide and hydrogen gas. It is described in this work. (O.L.). 4 figs

  9. Research and developments on nondestructive testing in fabrications of fast breeder reactor structural components in Japan

    International Nuclear Information System (INIS)

    Ebata, M.; Ooka, K.; Miyoshi, S.; Senda, T.

    1985-01-01

    Research and developments (R and D) have been conducted on the nondestructive testing techniques necessary for the construction of fast breeder reactor (FBR). Radiographic tests have been made on tube-tube plate welds and small-diameter tube welds, etc. Ultrasonic tests have been conducted on austenitic stainless steel welds. In the penetrant tests and magnetic particle tests, the investigations have been performed on the effects of various test factors on the test results

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

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1983-01-01

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

  11. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1979-01-01

    The experimental fast reactor ''Joyo'' will be tested at 75 MW output, starting in April, 1980. In connection with the accident in the Three Mile Island plant, the reexamination of the plant safety and the rechecking-up of the maintenance control system were carried out, and the special inspection by the Science and Technology Agency was executed from May 21 to 23, 1979. Thereafter, the preparation for raising the power output was completed. The periodical inspection after the completion of 50 MW operation is being carried out. The state of progress of various equipments and the codes for core characteristic analysis is reported. The construction preliminary design (2) of the prototype reactor ''Monju'' is examined, and the same design (3) is prepared. The analysis of the decay heat in the prototype reactor is carried on for the safety licensing. The technological investigation of LMFBRs in foreign countries is under way. The preliminary design (4) of the demonstration reactor is under examination, and the technical specifications of the conceptual design (1) are prepared. The researches and developments of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structures and materials, safety and steam generators are reported. (Kako, I.)

  12. 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.)

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

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Solomon, K.A.

    1979-07-01

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

  14. Survey of creep data on structural materials of fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, S.

    1977-11-01

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

  15. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Daeunert, U [Bundesministerium fuer Forschung und Technologie, Bonn (Germany); Kessler, G [Kernforschungszentrum Karlsruhe, Karlsruhe (Germany)

    1978-07-01

    Problems of public acceptance of nuclear power have affected the development of fast breeder reactors in the Federal Republic of Germany. Besides cooperation with UK, USA and Japan, the most outstanding event in the field of international fast breeder cooperation was a set of the agreements between Germany and France. These agreements opened the possibility of joint fast breeder development by Germany together with Belgium and the Netherlands. Most activities on the site of Compact Sodium Cooled Nuclear Reactor KNK-II were concerned with commissioning of the plant and final construction work. Criticality was achieved in Oct. 1977 and low-power tests performed. This paper includes a description of the status of construction of SNR-300 reactor and the results of research and development programmes performed. These were concerned with fuel elements development and results of irradiation experiments; development of cladding materials and core element structural materials; interaction between fuel and cladding; sodium tests; development and verification of computer codes; experiments in fast critical assemblies; fast rector safety; core disruptive accidents; development of instrumentation; thermodynamics od fuel assemblies; fluid dynamics.

  16. Liquid Metal Fast Breeder Reactor plant maintenance and equipment design

    International Nuclear Information System (INIS)

    Swannack, D.L.

    1982-01-01

    This paper provides a summary of maintenance equipment considerations and actual plant handling experiences from operation of a sodium-cooled reactor, the Fast Flux Test Facility (FFTF). Equipment areas relating to design, repair techniques, in-cell handling, logistics and facility services are discussed. Plant design must make provisions for handling and replacement of components within containment or allow for transport to an ex-containment area for repair. The modular cask assemblies and transporter systems developed for FFTF can service major plant components as well as smaller units. The plant and equipment designs for the Clinch River Breeder Reactor (CRBR) plant have been patterned after successful FFTF equipment

  17. Status of the fast breeder reactor technology in China

    International Nuclear Information System (INIS)

    Xu Mi

    1990-01-01

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

  18. Present status and problems of development of fast breeder reactors

    International Nuclear Information System (INIS)

    Ikeda, Hiroshi

    1984-11-01

    The development of FBRs in Japan now reached the stage to conclude on the development organization for a demonstration reactor positioning one step before a practical reactor. FBRs can be operated while converting uranium-238 existing in natural uranium by 99.3% to fissile plutonium-239, as the result, the nuclear fuel more than that consumed can be produced. However, there are various technical difficulties in FBRs, and the construction cost is estimated to be considerably higher as compared with that of LWRs. Also the plutonium obtained by reprocessing spent fuel is used for FBRs, accordingly, the development of FBRs is inseparable from the establishment of nuclear fuel cycle. In order to get rid of the burden of enormous development cost for FBRs, the trend of international joint development is conspicuous. The Superphenix with 1200 MWe output under construction centering around SERENA is expected to attain the criticality in the spring of 1985. For the development of a demonstration reactor, it is necessary to increase the role of private businesses, and the smooth transfer of know-how accumulated in Power Reactor and Nuclear Fuel Development Corp. to civilian side is an important problem. (Kako, I.)

  19. Fast Breeder Reactor Development in France During 1987

    International Nuclear Information System (INIS)

    Asty, M.; Cambillard, E.

    1988-01-01

    On March 8, 1987, a ''sodium leak'' alarm signal was received in the Creys-Malville control room. By the end of March, it had been established beyond all doubt that sodium was, in fact, leaking into the fuel storage drum inter-vessel gap. The reactor has been shut down since May 26. The origin of the leak was located on September 5, after complete drainage of the main tank. Despite the fact that the leak was confined, had had no radiological consequences and cast no doubts on the safe operation of the reactor, the impact of this incident on public opinion, both in France and in the neighbouring European states, was considerable. Two facts would appear to have been decisive. The first was that the reactor had not been shut down immediately, the second was that the leak was only detected and localized in September: it was difficult for people to understand that before its exact position could be determined, certain operations (transfer of a few subassemblies to the reactor core, unloading of the fuel storage drum) had to be performed

  20. World energy resources, demand and supply of energy, and the prospects for the fast breeder reactor

    International Nuclear Information System (INIS)

    Haefele, W.

    1978-01-01

    In the past it was taken for granted that the prime role of fast breeder reactors was to complement light water reactors, mainly because of their similar and compatible fuel cycles. In particular, the plutonium converted in LWRs is most intelligently disposed of and used in FBRs. Evaluation of the time horizon of such reactor strategies generally extended only to the year 2000. It is important to realize, however, that the salient task in the breeder field after 2000 - besides electricity generation - will be to substitute for conventional ''cheap'' oil. Electricity today makes up only 10% to 12% of the total secondary energy, while liquids essentially command up to about 50%. Thus the future application of the FBR technology will have to be geared more to the production of a liquid secondary energy carrier than to electricity. A new yardstick for all these considerations is the strongly rising energy prices. They may double, for example, leading to an oil price of US 24/bbl. Under these circumstances it is prudent to generalize the scope for future fast breeders. The key element of such a new fast breeder strategy would be the production of hydrogen by electrolysis or thermolysis or a combination of both. For example, methanol synthesized from hydrogen and residual fossil fuels would thus become economically attractive. The FBR breeding gain, on the other hand, would be used for the continued supply of LWRs generating electricity. The paper identifies order-of-magnitude considerations most important for such a fast breeder application against a global energy demand scenario for the year 2030. (author)

  1. The development of the gas cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Kernforschungszentrum Karlsruhe

    1975-01-01

    A survey of the present technological state is given on the basis of the developments made so far. Some milestones of development - e.g. the German gas breeder memorandum, the Gas Breeder Reactor Association the results of the BR-2 radiation experiments and of GfK-KWU design and safety studies - are described. The problems connected with a large store of plutonium are also discussed. (UA/AK) [de

  2. Integrated fuel-cycle models for fast breeder reactors

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  3. Correlation of Yield Stress And Microhardness in 08cr16ni11mo3 Irradiated To High Dose In The Bn-350 Fast Reactor

    International Nuclear Information System (INIS)

    Maksimkin, O.P.; Gusev, M.N.; Tivanova, O.S.; Silnaygina, N.S.; Garner, Francis A.

    2006-01-01

    The relationship between values of the microhardness and the engineering yield stress in steel 08Cr16Ni11Mo3 (Russian analog of AISI 316) heavily irradiated in the BN-350 reactor has been experimentally derived. It agrees very well with the previously published correlation developed by Toloczko for unirradiated 316 in a variety of cold-work conditions. Even more importantly, when the correlation is derived in the K δ format where the correlation involves changes in the two properties, we find excellent agreement with a universal K δ correlation developed by Busby and coworkers. With this K δ correlation, one can predict the value of yield stress in irradiated material based on measured values of microhardness. The technique is particularly suitable when the material of interest is in an inconvenient location or configuration, or when significant gradients in mechanical properties are anticipated over small dimensions. This approach makes it possible to reduce the labor input and risk when conducting such work. It appears that the derived correlation is equally applicable to both Russian and Western austenitic steel, and also in both irradiated and unirradiated conditions. Additionally, this report points out that microhardness measurements must take into account that high temperature sodium exposure alters the metal surface to produce ferrite, and therefore the altered layers should be removed prior to testing

  4. Acoustic control system BN-350. Explanatory note

    International Nuclear Information System (INIS)

    1982-02-01

    A description of the acoustic system developed to control boiling in the active zone of the BN 350 reactor is given together with the corresponding technical characteristics. The results of experiments and calculations which confirm the validity of the solutions adopted are discussed. Theoretical calculations on the boiling process in the duct are reported together with details on the fast diagnostic system. A means for localizing the onset of boiling is also given, possible error being taken into consideration. The special features of the passive acoustic diagnostic method used to study boiling are described and schemas of the anciliary equipment presented [fr

  5. Seismic analysis of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Gilbert, R.J.; Martelli, A.

    1989-06-01

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

  6. Liquid metal cooled fast breeder nuclear reactor constructions

    International Nuclear Information System (INIS)

    Chesworth, G.; Hind, J.R.; Hodgson, D.; Seed, G.

    1981-01-01

    In a nuclear reactor of the pool kind the primary vessel and fuel assembly are carried from the roof of the containment vault by tie straps. The primary vessel incorporates an annular yoke of 'k' cross-section the tie straps being attached to the upwardly directed vertical leg and the downwardly directed inclined leg. The upper and lower strakes of the primary vessel are extensions of the remaining legs. Load supporting welds therefore are of intermittent nature thereby limiting the effects of weld crack propagation

  7. Status of fast breeder reactor development in India

    International Nuclear Information System (INIS)

    Bhoje, S.B.

    1996-01-01

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

  8. Status of fast breeder reactor development in India

    Energy Technology Data Exchange (ETDEWEB)

    Bhoje, S B [Reactor Group, IGCAR, Kalpakkam (India)

    1996-07-01

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

  9. Construction work for prototype fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Suzuki, Yasuyuki; Tsuji, Koichi; Shimizu, Hisashi

    1991-01-01

    The main construction work of MONJU was started from the excavation for building foundation in October 1985, the containment vessel was prepared in April 1987, the reactor vessel was installed in October 1988, and the installation of the whole equipment was completed in April 1991. Fuji Electric made preparations for construction matching the above master schedule in consideration of construction schedule quality assurance and safety and accomplished the work within the scheduled time without personal injury as long as 2.35 million hours. (author)

  10. Nuclear data needs for fast breeder reactor shielding

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  11. Liquid Metal Fast Breeder Reactor Program: Argonne facilities

    International Nuclear Information System (INIS)

    Stephens, S.V.

    1976-09-01

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

  12. The early realization of fast breeder reactor cycle

    International Nuclear Information System (INIS)

    Mukai, Kazuo

    2007-01-01

    The Japanese government assessed the result of 'Feasibility Study on Commercialized FR Cycle System 2nd Phase' and a major concept that is the combination of a sodium cooled FBR (oxide fuel), an advanced aqueous reprocessing and a simplified pelletizing was selected. From now on, the JAEA invests the development resource to a major concept intensively, and aims to put it to practical use by new project 'FaCT'. With 'FaCT' project, the adoption of innovative technology is going to be judged in 2010. The result of 'Fact' are going to be presented in 2015 that is the systematized technologies based on R and D and the conceptual designs to demonstration facility and commercial facility. In order to start the demonstration reactor operation in 2025 by the result of 'FaCT', the project is carried with international cooperation. (author)

  13. Alternative fuels for the French fast breeder reactors programme

    International Nuclear Information System (INIS)

    Bailly, H.; Bernard, H.; Mansard, B.

    1988-01-01

    At the present time, due to the very competitive cost per kWh produced in France by the PWRs, it appears clear that, despite the improved use of uranium by FBRs, they will only be developed if the cost of the fuel cycle is sufficiently lower than that of the PWRs to compensate for the additional investment. The current economic programme has fixed the following fuel related objectives: - burn-up as high possible, the value of 150 000 MWd/t being considered as a minimum, and not a final target to be achieved, - extension of the duration of reactor operation cycles, leading to high in-pile times for fuel. Reaching the latter objective depends on obtaining high internal breeding gain performances, so that the total reactivity drop related to fuel impoverishment can be minimized. In this respect, a large diameter oxide fuel and/or an axial heterogeneous core concept can be envisaged. Dense fuels could form another solution. The feasibility of the fabrication of carbide and nitride fuels has been demonstrated in several countries and there is currently convergence towards a single type of process based on a carbothermic reaction. The optimization of fabrication procedures for these fuels must be continued to satisfy economic requirements and to obtain a fabrication cost of the same order or magnitude as that of oxide, although higher. If this target is achieved, fabrication will not be the major criterion for the selection of the FBR fuel, which will then be a function of the cost of reprocessing, performances under irradiation and reactor operating requirements

  14. Fusion-fission hybrid as an alternative to the fast breeder reactor

    International Nuclear Information System (INIS)

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

    1980-09-01

    This report compares the fusion-fission hybrid on the plutonium cycle with the classical fast breeder reactor (FBR) cycle as a long-term nuclear energy source. For the purpose of comparison, the current light-water reactor once-through (LWR-OT) cycle was also analyzed. The methods and models used in this study were developed for use in a comparative analysis of conventional nuclear fuel cycles. Assessment areas considered in this study include economics, energy balance, proliferation resistance, technological status, public safety, and commercial viability. In every case the characteristics of all fuel cycle facilities were accounted for, rather than just those of the reactor

  15. Status of fast breeder reactor development in the United States of America

    International Nuclear Information System (INIS)

    Horton, K.E.

    1983-01-01

    The goal of the United States Liquid Metal Fast Breeder Reactor (LMFBR) program is to develop the technology to the point that the private sector can deploy a safe, economic breeder reactor. The LMFBR will provide virtually inexhaustible supplies of electrical energy for the long term and will provide additional confidence to LWR nuclear deployment in the near term. The LMFBR program consists of a streamlined research and development effort focussing on those actions needed to enable private sector financing of industrial deployment including plant demonstration and technology efforts in reactor fuels, components, materials, physics, and safety

  16. Investigation of zero-release cycle using fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The task force was organized for the main purpose of offering quantitative basic data to the study group on nuclear fuel cycle in February, 1997. The effect of so-called frontier technologies such as the isotope separation by laser method, the FP annihilation with electron beam accelerators and so on in the FBR cycle based on MOX fuel and PUREX reprocessing method was expected. It is aimed at to recycle the total amount of minor actinides. The object of recycling is the nuclides which contribute largely to toxicity, namely 11 elements, 12 nuclides. The preconditions and the target to be attained of the investigation are explained. As the results of investigation, the amount of reloading MA and FP into a reactor, squeezing the recycling scenario, the effect of reducing toxicity and the subject of the countermeasures to the nuclides with long half-life which cannot be reloaded are reported. As the technical evaluation required for realizing the concept, the concept of the core which excludes recriticality, the advance of reprocessing technology, isotope separation, the fabrication into the optimal form for recycling and so on are discussed. The economical efficiency of the recycling based on MOX and PUREX and the proposal of the development scenario are described. (K.I.)

  17. Decay heat removal for the liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Zemanick, P.P.; Brown, N.W.

    1975-01-01

    The functional and reliability requirements of the decay heat removal systems are described. The reliability requirement and its rationale as adequate assurance that public health and safety are safeguarded are presented. The means by which the reliability of the decay heat removal systems are established to meet their requirement are identified. The heat removal systems and their operating characteristics are described. The discussion includes the overflow heat removal service and its role in decay heat removal if needed. The details of the systems are described to demonstrate the elements of redundancy and diversity in the systems design. The quantitative reliability assessment is presented, including the reliability model, the most important assumptions on which the analysis is based, sources of failure data, and the preliminary numerical results. Finally, the qualitative analyses and administrative controls will be discussed which ensure reliability attainment in design, fabrication, and operation, including minimization of common mode failures. A component test program is planned to provide reliability data on selected critical heat removal system equipment. This test plan is described including a definition of the test parameters of greatest interest and the motivation for the test article selection. A long range plan is also in place to collect plant operational data and the broad outlines of this plan are described. A statement of the high reliability of the Clinch River Breeder reactor Plant decay heat removal systems and a summary of the supporting arguments is presented. (U.S.)

  18. Post-accident monitoring systems in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Suriya Murthy, N.; Sivasailanathan, Vidhya; Ananth, Allu; Roy, Kallol

    2018-01-01

    PFBR is a 500 MW(e) MOX fueled and sodium cooled fast reactor (SFR) under advanced stage of commissioning at Kalpakkam. Currently, the main vessel is preheated and sodium has been charged into two secondary loops that are operated in recirculation mode. In order to characterize the radiation field and contamination, the workplace monitoring is undertaken using installed monitors that are commissioned and made operational. This helps to ensure radiological protection during normal operating conditions. On the other hand, radiological monitoring in emergency conditions is quite different. For undertaking the mitigative accident management, a set of specialized nuclear instruments called post-accident monitoring systems (PAMS) which include radiation monitors are stipulated. The Fukushima Daiichi accident emphasized the importance and need for reliable accident monitoring instrumentation to indicate the safety functions during the progression and aftermath of accident in NPP. In PFBR, the PAMS are integrated with other monitoring systems in design stage itself to manage the measurements and indicating the safety functions for implementing EOP and SAMG

  19. Decay Heat Removal for the Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zemanick, P. P.; Brown, N. W.

    1975-10-15

    The functional and reliability requirements of the decay heat removal systems are described. The reliability requirement and its rationale as adequate assurance that public health and safety are safeguarded are presented. The means by which the reliability of the decay heat removal systems are established to meet their requirement are identified. The heat removal systems and their operating characteristics are described. The discussion includes the overflow heat removal service and its role in decay heat removal if needed. The details of the systems are described to demonstrate the elements of redundancy and diversity in the systems design. The quantitative reliability assessment is presented, including the reliability model, the most important assumptions on which the analysis is based, sources of failure data, and the preliminary numerical results. Finally, the qualitative analyses and administrative controls will be discussed which ensure reliability attainment in design, fabrication, and operation, including minimization of common mode failures. A component test program is planned to provide reliability data on selected critical heat removal system equipment. This test plan is described including a definition of the test parameters of greatest interest and the motivation for the test article selection. A long range plan is also in place to collect plant operational data and the broad outlines of this plan are described. The paper closes with a statement of the high reliability of the Clinch River Breeder Reactor Plant decay heat removal systems and a summary of the supporting arguments. (author)

  20. Water storage of liquid-metal fast-breeder-reactor fuel

    International Nuclear Information System (INIS)

    Meacham, S.A.

    1982-01-01

    The purpose of this paper is to present a general overview of a concept proposed for receiving and storing liquid metal fast breeder reactor (LMFBR) spent fuel. This work was done as part of the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL). The CFRP has as its major objective the development of technology for reprocessing advanced nuclear reactor fuels. The program plans that research and development will be carried through to a sufficient scale, using irradiated spent fuel under plant operating conditions, to establish a basis for confident projection of reprocessing capability to support a breeder industry

  1. Status and programme for the fast breeder reactor in the UK

    International Nuclear Information System (INIS)

    Franklin, N.L; Hill, J.

    1977-01-01

    The paper briefly reviews the long standing objectives and strategy for the introduction of the Fast Breeder Reactor in the U.K. and goes on to comment upon the early operation of the prototype PFR and on the status of the design of the first commercial demonstration plant DFR 1. The arrangements for the supporting technology and component development are discussed together with the contributions that can follow from collaboration. The out-of-pile fuel cycle, so critical to the success of the fast reactor power programme, has its principal objectives and timescales identified and these form the bases of the several U.K. papers that follow in this Conference

  2. Technology development program for safe shipment of spent fuel from liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Freedman, J.M.; Humphreys, J.R.

    1975-10-01

    A comprehensive plan to develop shipping cask technology is described. Technical programs in the disciplines of heat transfer, structures and containment, spent fuel characterization, hot laboratory verification, shielding, and hazards analysis are discussed. Both short- and long-term goals in each discipline are delineated and how the disciplines interrelate is shown. The technologies developed will be used in the design, fabrication, and testing of truck-mounted and rail-car casks. These casks will be used for safely transporting short-cooled, high-burnup Liquid Metal Fast Breeder Reactor (LMFBR) spent fuel from reactors to reprocessing plants

  3. Fast breeder reactor fuel reprocessing R and D: technological development for a commercial plant

    International Nuclear Information System (INIS)

    Colas, J.; Saudray, D.; Coste, J.A.; Roux, J.P.; Jouan, A.

    1987-01-01

    The technological developments undertaken by the CEA are applied to a plant project of a 50 t/y capacity, having to reprocess in particular the SUPERPHENIX 1 reactor fuel. French experience on fast breeder reactor fuel reprocessing is presented, then the 50 t/y capacity plant project and the research and development installations. The R and D programs are described, concerning: head-end operations, solvent extractions, Pu02 conversion and storage, out-of-specification Pu02 redissolution, fission products solution vitrification, conditioning of stainless steel hulls by melting, development of remote operation equipments, study of corrosion and analytical problems

  4. The role of materials in the analysis of fast breeder reactor components

    International Nuclear Information System (INIS)

    Aubert, Michel; Petrequin, Pierre.

    1982-09-01

    The analysis of fast breeder reactor components involves the knowledge of certain properties of the materials used. The latter consist of the following: - a body of data required for calculations, including allowable stresses and fatigue strength, as well as the rules applicable to these data, - a number of qualitative requirements serving to guarantee that the quality of the material fully justifies the use of the previously established elements. This duality of concerns is illustrated by some recent examples which occured during the construction of the Super Phenix reactor [fr

  5. Safety of the liquid-metal cooled fast breeder reactor and aspects of its fuel cycle

    International Nuclear Information System (INIS)

    Kessler, G.; Papp, R.; Huebel, D.

    1977-01-01

    Design and construction of the sodium-cooled fast reactors KNK-II (20MW(e)) and SNR-300 (300MW(e)) determine the status of safety engineering and safety R and D of LMFBRs in the Federal Republic of Germany. Both prototype fast power reactors have to go through a civil licensing process similar to that applied to present LWRs. A multilevel safety - or defence in depth - approach is applied to the design and construction of fast power reactors. All design data of the fast reactor plant are confirmed by extensive experimental programmes. Design limits of the plant are thoroughly discussed during the licensing process. Important safety R and D programmes have been and are still being performed. A very conservative safety analysis for hypothetical core and other plant accidents is used for present prototype fast reactors. The paper reviews the future trend of development of theoretical methods for accident analysis and the application of experimental results, especially in view of large commercial-type LMFBRs. The safety approach applied to the LMFBR plant is safe operation under normal operating conditions and safe shutdown under off-normal conditions. The consequences of releases of radioactivity to the environment meet the given standards. No chemical reprocessing plant for fast breeder fuel is in operation in the FRG at present; however, R and D work on investigation of all aspects and problem areas of the fast breeder fuel cycle are under way. Systems studies on safety aspects of the fast breeder fuel cycle (transport, reprocessing, fuel fabrication) and its impact on the environment have been performed and the main consequences of these studies are presented in the paper. (author)

  6. Progress Report on Sodium Cooled Fast Breeder Reactor Development in Japan, April 1975

    International Nuclear Information System (INIS)

    Tomabechi, K.

    1975-01-01

    The progress of the sodium cooled fast Breeder Reactor development in Japan in the past 12 months can be summarized as follows. Installation of all the components of the Experimental Fast Reactor, ''JOYO'', was completed in the end of the last year and various commissioning tests of the reactor began in January 1975. It is planned to charge sodium into the reactor in coming fall and the first criticality experiment is currently planned in the summer 1976. Most of the research and development works for ''JOYO'' are nearing completion. These include an endurance test of 3 prototype primary sodium pump for 12,000 hours. 86 core fuel subassemblies and 220 blanket subassemblies, a sufficient number for composing the initial core, have already been fabricated. Concerning the Prototype Fast Breeder Reactor, ''MONJU'', design activity as well as relevant research and development works are continued. A siting problem exists and it is hoped to be resolved soon. Of the research and development works, a significant achievement in the past 12 months can be a successful operation at full power of the 50 MW Steam Generator Test Facility. This facility was put into operation at full power in June 1974. No leak of water into sodium has been experienced with operation of the steam generator tested. The steam generator is being dismantled for a detailed inspection originally planned

  7. Pu utilization in fast-breeder and in light-water reactors in Italy

    International Nuclear Information System (INIS)

    Mangiagalli, D.; Cicognani, F.; Pistella, F.; Testa, G.; Villani, A.; Ariemma, A.; Castelli, G.F.; Linari, A.; Paoletti Gualandi, M.; Musso, B.

    1977-01-01

    The paper illustrates the most important activities carried out in Italy for the development of fast breeder reactors and its fuel as well as for plutonium recycle in light water reactors. The Italian strategy is based, on one hand, on the short-term commercialization of fast breeder reactors, and on the other, on the adoption of the technology of the Phenix prototype whose further development will be ensured by the joint Italian and French efforts as insured by the important agreements signed by CNEN, NIRA (Nucleare Italiana Reattori Avanzati) and Italian manufacturing industries with CEA and the main French industries. The paper also includes the main results of the ENEL Demonstration Program on Pu prototypes introduced in the Garigliano BWR in 1968 and 1970, and of the destructive and non-destructive analyses on said fuel, as well as of the analyses carried out by CNEN on prototypical fuel fabricated by CNEN and irradiated in various reactors. Furthermore, the paper deals with design and licensing aspects of the 46 Pu-island assembly reload introduced in the Garigliano reactors in 1975 and of a batch of 8 all Pu assemblies loaded in the Trino Vercellese PWR in 1976. Subsequently, the experimental activities planned for the near future both on high burn-up prototypes and on industrial fuel after one cycle of operation are examined [fr

  8. 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

  9. Progress report on fast breeder reactor development in Japan, July - September 1977

    International Nuclear Information System (INIS)

    1978-06-01

    As for the experimental fast breeder reactor ''Joyo'', the low power performance tests have been continued, and the measurements of reactor noise, the reactivity of fuel assemblies, power distribution and Na-void effect have been made. Efforts have been exerted to develop the required maintenance equipments, to manufacture the transfer rotor maintenance facilities, and to construct the spent fuel storing and cooling facilities. The analysis and calculation of the core characteristics have been in progress. The design work on the prototype fast breeder reactor ''Monju'' has been continued, and the development of the computer codes for the design has progressed. Informations have been gathered regarding the technological developments of LMFBRs overseas. The surveys on the site for ''Monju'' have been carried out. The design and research works on the demonstration reactor were started, and the general design factors such as the steam condition and the plant layout have been studied. As for the research and development of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structural materials, safety, and steam generators, the progresses are reported in detail. High performance neutron detectors for nuclear instrumentation have been under development, and the tagging gas method for fuel failure detection and location system has been tested. (Kako, I.)

  10. Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

    2003-01-01

    Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

  11. 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

  12. Comparison of diffusion and transport theory analysis with experimental results in fast breeder test reactor

    International Nuclear Information System (INIS)

    Sathyabama, N.; Mohanakrishnan, P.; Lee, S.M.

    1994-01-01

    A systematic analysis has been performed by 3 dimensional diffusion and transport methods to calculate the measured control rod worths and subassembly wise power distribution in fast breeder test reactor. Geometry corrections (rectangular to hexagonal and diffusion to transport corrections are estimated for multiplication factors and control rod worths. Calculated control rod worths by diffusion and transport theory are nearly the same and 10% above measured values. Power distribution in the core periphery is over predicted (15%) by diffusion theory. But, this over prediction reduces to 8% by use of the S N method. (authors). 9 refs., 4 tabs., 3 fig

  13. Study of the containment system of the planned SNR-2 Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Bunz, H.; Scholle, U.

    1979-01-01

    Research is presented concerning the behaviour of four possible containment concepts for the planned Fast Breeder Reactor SNR-2 under the conditions of a HCDA. Detailed descriptions and calculations are given dealing with the acceptable leakages and the filter loads reached. The transport of the aerosol-type as well as the gaseous activity through the different compartments into the environment is calculated and from this the resulting accident doses. To cover the uncertainties of the accident course a wide range of accident parameters is examined

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  15. On the actual controlling of standards concerning the 'fast breeder reactor'

    International Nuclear Information System (INIS)

    Rengeling, H.W.

    1978-01-01

    If the decision of the OVG Muenster to present the case to the Federal Constitutional Court asking whether Article 7 of the atomic energy law corresponds to the constitution or not, as far as the article allows the licensing of a fast breeder reactor, two problems arise: The legal question in how far the actual controlling of standards is to be preceeded by a statement of facts given by the court of first instance, and the problem behind concerning the responsibility of decision. The Federal Constitutional Court should accept the responsibility of decision to be borne by it. (orig.) [de

  16. Assessment of weld joints of steam generator of prototype fast breeder reactor by microfocal radiography

    International Nuclear Information System (INIS)

    Venkatraman, B.; Saravanan, T.; Jayakumar, T.; Kalyanasundaram, P.; Raj, B.

    2004-01-01

    The tube to tubesheet (TTS) welds of steam generator of Prototype Fast Breeder Reactor (PFBR) are quite critical. Sodium flows on shell side and water on tube side. Any failure would thus be catastrophic. Apart from defects such as porosities, wall thinning due to concavity is endemic in such joints and needs to be detected. This paper presents the methodologies developed for quantitative evaluation of defects including wall thinning due to concavity in the TTS welds by micro focal radiography. The method has been successfully adopted in the shop floor for the evaluation of TTS welds of steam generator and evaporator. (author)

  17. Dynamical analysis on carbon transfer in liquid metal cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Kataoka, Tadayuki; Matsumoto, Keishi

    1979-01-01

    The dynamical analysis was undertaken on the exchange of carbon taking place between the structural steels and sodium for the case of a bi-metallic secondary system constituted of type 304 stainless and 2 1/4Cr-1Mo steels, representing the secondary system of a liquid sodium cooled fast breeder reactor. The analysis brought to light the effects to be expected on the long terms carbon transfer behavior of: (a) the surface areas of structural steels in contact with flowing sodium, (b) the thickness of the sodium-boundary layer, (c) the initial carbon concentration in the sodium, and (d) the rate of carbon contamination of the sodium. (author)

  18. Main boiler feed pump for fast breeder test reactor. Failure analysis and remedial measures

    International Nuclear Information System (INIS)

    Iyer, M.A.K.; Chande, S.K.; Raghuvir, A.D.; Baskar, S.; Kale, R.D.

    1994-01-01

    A small capacity ten stage 670 kw feed water pump is used for supplying feed water at a temperature of 190 deg C to a once through steam generator in the Fast Breeder Test Reactor at Kalpakkam. During preparatory heating up stage to commission the steam generator the pump suffered a severe loss of suction which resulted in failure of hydrostatic journal bearings and extensive damage to pump internals. This paper discusses the detailed mechanism of loss of suction, details of damage to the pump and various modifications carried out to prevent recurrence of the problem. (author). 4 refs., 3 figs., 2 tabs

  19. Compendium of computer codes for the safety analysis of fast breeder reactors

    International Nuclear Information System (INIS)

    1977-10-01

    The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available

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

    Energy Technology Data Exchange (ETDEWEB)

    Scaller, K; Vrillon, B

    1980-02-01

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

  1. Stabilization of magnet assemblies of permanent magnet sodium flowmeters used in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rajan, K.K., E-mail: kkrajan@igcar.gov.in; Vijayakumar, G.

    2014-08-15

    Highlights: • Stabilization procedure for ALNICO-5 permanent magnet material is evolved. • Effect of time and temperature on ALNICO-5 assembly is determined. • Suitability of ALNICO-5 flowmeters at high temperatures is established. • Temperature coefficient of flux density is determined. - Abstract: Permanent magnet flow meters (PMFMs) are used to measure the sodium flow in sodium cooled Fast Breeder Reactor Circuits. Prototype fast breeder reactor (PFBR) which is under construction at Kalpakkam is a 500 MWe, sodium cooled, pool type reactor. Sodium flow measurement in various loops of the reactor is of prime importance from operational and safety point of view. To measure the flow of electrically conducting liquid sodium, in primary and secondary circuit pipe lines of PFBR, permanent magnet flow meters are used. PMFM is a non-invasive device, which works on the principle of generation of motional EMF by magnetic forces exerted on the charges in a moving conductor. Flowmeters of different pipe sizes ranging from 10 mm to 200 mm pipe diameter are required for PFBR. Long term performance of the flowmeters mainly depends on stability of permanent magnets used in flowmeters to generate constant magnetic field in stainless steel (SS) pipes. This paper describes the effects of time and temperature on permanent magnet assemblies made of ALNICO-V used in PFBR flowmeters. The stabilization methodology for ALNICO-V permanent magnet assemblies is evolved and established. Loss of magnetic field strength with respect to time and temperatures is determined by experiments and found negligible.

  2. Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

    2003-01-01

    The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

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

    International Nuclear Information System (INIS)

    Scaller, K.; Vrillon, B.

    1980-01-01

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

  4. Progress report on fast breeder reactor development in Japan, October - December 1976

    International Nuclear Information System (INIS)

    1977-03-01

    As for the fast breeder experimental reactor ''Joyo'', the remodeling works for the bearings of primary main circulation pump and the support bands of hanger support have finished. Preparatory arrangement was made for the second performance test, such as preheating the primary cooling system and filling sodium in it. Construction, installation and adjustment of test and inspection apparatuses required after the criticality experiment are in progress. The lining work for the temporary storage pool for solid wastes, the enlarging work for the operation control building, and construction of the brake equipment for cooler blower were over. As for the operation control, operation, inspection and maintenance of each system required for the modeling work are being carried out, and new parts to take the place of old ones were provided. The analysis of the core characteristics and the development of operation-monitoring codes are in progress. The coordination design work (4) for the fast breeder prototype reactor ''Monju'' has been started, and the preliminary design work (2) for the demonstration reactor is going on. Geological, meteorological and seismic researches carried out at the proposed construction site for ''Monju''. The researches and developments on reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, safety and steam generators are in progress. (Kako, I.)

  5. Cold trap dismantling and sodium removal at a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Graf, Anja; Petrick, Holger; Stutz, Uwe [WAK GmbH, Eggenstein-Leopoldshafen (Germany). Hauptabt. Dekontaminationsbetriebe Rueckbau Kompakte Natriumgekuehlte Kernreaktoranlage (KNK); Hosking, Paul [Nuclear Decommissioning Services Limited (NDSL), Sutherland, Dornoch (United Kingdom)

    2013-11-15

    The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. As this reactor type used sodium as a coolant in its primary and secondary circuit, 7 cold traps containing various amounts of partially activated sodium needed to be disposed of as part of the dismantling. The resulting combined difficulties of radioactive contamination and high chemical reactivity were handled by treating the cold traps differently depending on their size and the amount of sodium contained inside. Six small cold traps were processed on-site by cutting them up into small parts using a band saw under a protective atmosphere. The sodium was then converted to sodium hydroxide by using water. The remaining large cold trap could not be handled in the same way due to its dimensions (2.9 m x 1.1 m) and the declared amount of sodium inside (1,700 kg). It was therefore manually dismantled inside a large box filled with a protective atmosphere, while the resulting pieces were packaged for later burning in a special facility. The experiences gained by KNK during this process may be advantageous for future dismantling projects in similar sodium-cooled reactors worldwide. (orig.)

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

    International Nuclear Information System (INIS)

    Cunningham, G.W.

    1977-01-01

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

  7. 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)

  8. Present state of inspection robot technology in nuclear power facilities. Case of fast breeder reactors

    International Nuclear Information System (INIS)

    Ara, Kuniaki

    1995-01-01

    In the maintenance works in nuclear power facilities such as checkup, inspection and repair, for the main purpose of radiation protection, remote operation technology was introduced since relatively early stage, and at present, the robots that carry out the inspection works for confirming the soundness of main equipment have been developed and put to practical use. At the time of introducing these technologies, in addition to the research and development of robots proper, the coordination with the design of plant machinery and equipment facilities as the premise of introducing robots is an important requirement. In this report, the present state of the development of remote inspection technology for fast breeder reactors is introduced, and the matters to which attention is paid in the plant design for introducing robots are explained. First, fast breeder reactors are described. The needs of robotizing and adopting remote operation in nuclear power facilities are explained, using the examples of the inspection system for a reactor vessel and the inspection system for steam generator heat transfer tubes. (K.I.)

  9. Safety requirements expected to the prototype fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    2014-11-01

    In July 2013, Nuclear Regulation Authority (NRA) has enforced new regulatory requirements in consideration of severe accidents for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFR) of 'Monju' based on TEPCO Fukushima Daiichi nuclear power plant accident (hereinafter referred to as '1F accident') occurred in March 2011. Although the regulatory requirements for SFR will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency (JAEA) set up 'Advisory Committee on Monju Safety Requirements' consisting of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected to the prototype FBR 'Monju' considering severe accidents with knowledge from JAEA as well as scientific and technical insights from the experts. This report summarizes the safety requirements expected to Monju discussed by the committee. (author)

  10. A capital cost reduction study on the fast breeder reactor plant

    International Nuclear Information System (INIS)

    Taniyama, H.; Kamei, M.; Moriyama, M.

    1991-01-01

    A capital cost reduction study has been performed for large fast breeder reactor designs. The primary objective of this study is to show a trend of capital cost reduction between FBR plants at the prototype stage, the demonstration stage, and the future commercialization stage. For the FBR plant at the demonstration stage a construction cost comparison with a light water reactor has also been performed, and the target cost of FBR of below 1.5 times that of the light water reactor cost was achieved. To extend the capital cost reduction study, a feasibility study was made to achieve a capital cost of an FBR less than that of a light water reactor. The recommended design is shown as a future commercialization FBR design concept. (author)

  11. Status of fast breeder reactor development in the United States of America - April 1985

    International Nuclear Information System (INIS)

    Horton, K.E.

    1986-01-01

    In order to provide a continuum of development for liquid metal reactors, the U.S. program has been reshaped into two portions -- advanced converter reactor technology including the Liquid Metal Reactor for the near and intermediate term, and the Liquid Metal Fast Breeder Reactor for deployment in the twenty-first century. The focused research and development program is directed at innovative means to improve economics, provide inherent safety, and to meet the needs of the ultimate user, the utilities. Work in the fuels and materials, and reprocessing areas is being continued to support eventual deployment. A major factor in successful deployment will be the effectiveness of international collaboration in reducing costs and duplication of efforts

  12. Development of sputter ion pump based SG leak detection system for Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Babu, B.; Sureshkumar, K.V.; Srinivasan, G.

    2013-01-01

    Highlights: ► Development and commissioning of SG leak detection system for FBTR. ► Development of Robust method of using sputter ion pump based system. ► Modifications for improving reliability and availability. ► On line injection of hydrogen in sodium during reactor operation. ► Triplication of the SG leak detection system. - Abstract: The Fast Breeder Test Reactor (FBTR) is a 40 MWt, loop type sodium cooled fast reactor built at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam as a fore-runner to the second stage of Indian nuclear power programme. The reactor design is based on the French reactor Rapsodie with several modifications which include the provision of a steam-water circuit and turbo-generator. FBTR uses sodium as the coolant in the main heat transport medium to transfer heat from the reactor core to the feed water in the tertiary loop for producing superheated steam, which drives the turbo-generator. Sodium and water flow in shell and tube side respectively, separated by thin-walls of the ferritic steel tubes of the once-through steam generator (SG). Material defects in these tubes can lead to leakage of water into sodium, resulting in sodium water reactions leading to undesirable consequences. Early detection of water or steam leaks into sodium in the steam generator units of liquid metal fast breeder reactors (LMFBR) is an important requirement from safety and economic considerations. The SG leak in FBTR is detected by Sputter Ion Pump (SIP) based Steam Generator Leak Detection (SGLD) system and Thermal Conductivity Detector (TCD) based Hydrogen in Argon Detection (HAD) system. Many modifications were carried out in the SGLD system for the reactor operation to improve the reliability and availability. This paper details the development and the acquired experience of SIP based SGLD system instrumentation for real time hydrogen detection in sodium for FBTR.

  13. Annual technical report of the prototype fast breeder reactor Monju. 2012

    International Nuclear Information System (INIS)

    2013-09-01

    The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the primary achievements and the data related to the plant management in Monju during fiscal 2012. From the aspect of the design evaluation, the following items are summarized: 1) Comprehensive safety assessments of Monju taking into account the accident at Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company, 2) Evaluation of nuclear characteristics based on the data of core confirmation test, 3) Evaluation of hydrogen flux from steam generator tubes, 4) Construction of the advanced safeguards system, 5) Development of a plant dynamics analytical model for the Monju ex-vessel fuel storage system. Then, from the aspect of the maintenance technology, the following items are summarized: 1) Response to the administrative order to the defect of maintenance management, 2) Recovery of in vessel transfer machine dropping accident, 3) Work management by introduction of packaged isolation, 4) Evaluation of result of vibration control of RID sampling blower for secondary sodium loop. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

  14. Fast breeder reactor expected to soon overtake LWRs in terms of economic efficiency

    International Nuclear Information System (INIS)

    Eitz, A.W.; Theisen, G.; Wolff, U.; Riethmueller, R.

    1986-01-01

    On the basis of current knowledge, the authors try to comparatively assess the fuel cycle costs of the SNR and the LWR, and to establish a prognosis of the competibility of the SNR in terms of fuel expenditure. The result is that already today, the fast breeder would achieve the economic efficiency of the LWR-type reactor, and in future would offer clear advantages. For their assessments, the authors made a few cost estimates, as e.g. with regard to the fuel reprocessing plant, which deserve close attention, and power discussion. As to the next project to follow, a large-scale European fast breeder reactor power plant, the authors state that the fuel cycle cost will not present a special economic risk, if an option for interim fuel storage is ensured. Determination of the right and adequate time to start construction of a commercial fuel reprocessing plant is no point in current discussions, and should anyhow be made with great care and the least time pressure possible. Cooperation among European states, however, is indispensable for this project. (orig.) [de

  15. Comparison of material property specifications of austenitic steels in fast breeder reactor technology

    International Nuclear Information System (INIS)

    Vanderborck, Y.; Van Mulders, E.

    1985-01-01

    Austenitic stainless steels are very widely used in components for European Fast Breeder Reactors. The Activity Group Nr.3 ''Materials'', within Working Group ''Codes and Standards'' of the Fast Reactor Co-Ordination Committee of the European Communities, has decided to initiate a study to compare the material property specifications of the austenitic stainless steel used in the European Fast Breeder Technology. Hence, this study would allow one to view rapidly the designation of a particular steel grade in different European countries and to compare given property values for a same grade. There were dissimilarities, differences or voids appear, it could lead to an attempt to complete and/or to uniformize the nationally given values, so that on a practical level interchangeability, availability and use ease design and construction work. A selection of the materials and of their properties has been made by the Working Group. Materials examined are Stainless Steel AISI 304, 304 L, 304 LN, 316, 316 L, 316 LN, 316''Ti stab.'', 316''Nb stab''., 321, 347

  16. Annual technical report of the prototype fast breeder reactor Monju. 2013

    International Nuclear Information System (INIS)

    2014-08-01

    The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the principal achievements and the data related to the plant management in Monju in fiscal 2013. From the aspect of the management and design evaluation, the following items are summarized: 1) Current status of coping with the order from NRA to alter the safety regulations. 2) Implementation status of reformation of Monju. 3) Current status of the additional geological surveys of crush zones at the Monju site. 4) Development of core seismic assessment method for FBR. Then, from the aspect of the operation and maintenance technology, the following items are summarized: 1) Response to the administrative order to the defect of maintenance management (Part 2). 2) Performance confirmation of the failed fuel detection and location system. 3) Deviation from the limiting conditions for operation in the emergency diesel generator periodic test and so on. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

  17. Study on enhancement of heat transfer of reactor vessel auxiliary cooling system of fast breeder reactor

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Kinoshita, Izumi; Ueda, Nobuyuki; Furuya, Masahiro

    1996-01-01

    A reactor vessel auxiliary cooling system (RVACS), which is one of the decay heat removal systems of the fast breeder reactor (FBR), has passive safety as well as high reliability. However, the heat removal capability is relatively small, because its heat exchange is dependent on the natural convection of the air. The objectives of this report are to propose a heat transfer medium to enhance the heat transfer and to confirm the heat transfer performance of this system by experimental and analytical studies. From these studies, the following main results were obtained. (1) A porous plate with 5 mm thickness, 5 mm pore diameter, 92% porosity, was found to have the highest enhancement of heat transfer. (2) The heat transfer enhancement was demonstrated by large scale heat transfer experiments. Also, the heat transfer correlations, which can be used in the plant transient analyses, were derived from the experimental results. (3) Analysing the transient conditions of conventional pool-type FBR by means of the system analysis code, the applicable range of this system was assumed from the capability of the RVACS with porous plates. As a result, this type of RVACS was found to be applicable to conventional pool-type FBRs with capacity of about 500 MWe or less. (author)

  18. Status and program of development of the fast breeder reactor system in the U.S

    International Nuclear Information System (INIS)

    Roberts, R.W.

    1977-01-01

    The U.S. Energy Research and Development Administration's highest priority breeder concept is the Liquid Metal Fast Breeder Reactor (LMFBR). Its objective is the development of a broad technological and engineering base with extensive utility and industrial involvement which will establish a timely capability for a competitive and commercial breeder industry. In addition to technological development, an integral part of the LMFBR program is the transfer of LMFBR technology to the nuclear industry and the parallel development of utility capability. An important indicator of the effectiveness of this program is the successful construction and operation of a number of LMFBR experimental and test reactors and associated non-nuclear test facilities such as those located at Santa Susana, California, the Argonne (III). National Laboratory and the Hanford Engineering Development Laboratory at Richland, Washington. A principal element of the U.S. fast breeder reactor program is the Fast Flux Test Facility (FFTF). The largest test reactor facility in the world, the FFTF will focus on the testing of fuels and material. The FFTF, being built at the Hanford Engineering Development Laboratory, is scheduled for completion in the late 1970's. The next step toward timely commercial breeder capability is to implement a program for a large-scale demonstration of this concept. A 380 megawatt electrical demonstration plant - designated the Clinch River Breeder Reactor Plant (CRBRP) - will be constructed and operated near Oak Ridge, Tennessee, under a cooperative arrangement of industrial contractors, utilities, and the Government. The completion date is 1983. The estimated cost of the CRBRP project is 1.95 billion dollars. This includes the cost of design, construction, related research and development, and five years of operation - the full timespan from the project's beginning in 1972 through completion. The Nation's principal electric utilities have strongly endorsed the

  19. The case for developing the liquid metal fast breeder reactor in France

    International Nuclear Information System (INIS)

    Baumier, J.; Duchatelle, L.

    1984-01-01

    An analysis is presented of the economic aspects of electricity generation by fast breeder reactor (FBR) plants, with particular reference to FBRs planned in France during the 1980s, and in comparison with French pressurized water reactor (PWR) units. Individual sections deal with the following: an outline of the bases of comparison between FBRs and PWRs, describing methods used to calculate costs; a description of the FBRs and PWRs concerned, defining and explaining the concepts of breeding gain and doubling time for FBRs, and calculating the energy extracted from uranium by the two types of reactor; the results of analysis and FBR/PWR comparison for capital, operating, fuel cycle and total generating costs, drawing conclusions for the competitivity of FBRs; general conclusions on the advisability of large scale FBR deployment. (U.K.)

  20. Diversion analysis and safeguards measures for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Persiani, P.J.

    1981-10-01

    The general objective of the study is to perform a diversion analysis and an assessment of the available safeguards methods and systems for verifying inventory and flow of nuclear material in accessible and inaccessible areas of liquid-metal fast breeder reactor, LMFBR, systems. The study focuses primarily on the assembly-handling operations, assembly storage facilities, and reactor operations facilities relating to existing and/or near-term planned experimental, demonstration and prototypal reactor plants. The safeguards systems and methods presented are considered to be feasible for development and for implementation within the resource limitation of the IAEA and are considered to be consistent with the objectives, requirements, and constraints of the IAEA as outlined in the IAEA documents INFCIRC/153 and INFCIRC/66-Rev-2

  1. Compatibility of steels for fast breeder reactor in high temperature sodium

    International Nuclear Information System (INIS)

    Yuhara, Shunichi

    1981-01-01

    In recent years, considerable progress has been made and experience has been obtained for material applicability in sodium-cooled fast breeder reactors. In this report, materials, principal dimensions and sodium conditions for the reactor system components, which include fuel pin cladding, intermediate heat exchangers, steam generators and pipings, are reviewed with emphasis on the thin-walled, high temperature and high strength components. The corrosion, mechanical and tribological behavior in sodium of important materials used for the reactor components, such as Types 304 and 316 stainless steel and 2 1/4Cr-1Mo steel, are discussed on the basis of characteristic testing results. Furthermore, material requirements concerned with compatibility in sodium are summarized from this review and discussion. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

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

  3. Risk-oriented analysis on the German prototype fast breeder reactor SNR-300

    International Nuclear Information System (INIS)

    Bayer, A.; Koeberlein, K.; Gesellschaft fuer Reaktorsicherheit, Garching, Germany)

    1984-01-01

    On request of a fact-finding committee of the German Federal Parliament, a risk-oriented analysis on the SNR-300, the German prototype fast breeder reactor, has been performed to allow a pragmatic safety comparison of the SNR-300 and a modern light-water reactor. Results of the technical plant analysis have been summarized in seven release categories. Accident consequences have been calculated for the actual site at Kalkar/Rhine. The results indicate that for the SNR-300 both the frequency of major accidents and the consequences of accidents are smaller than for the pressurized-water reactor analyzed in the German Risk Study. This article summarizes the methods and main results of the analysis of the SNR-300

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  5. The fast breeder reactor Rapsodie (1962); Le reacteur rapide surregenerateur rapsodie (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Vautrey, L; Zaleski, C P [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1962-07-01

    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) [French] Dans ce rapport, les auteurs font le point du projet RAPSODIE (reacteur francais surregenerateur a neutrons rapides), au moment du debut effectif de sa construction. On y trouvera decrits: les principales caracteristiques neutroniques et thermiques, le bloc pile et les circuits de refroidissement, les principaux moyens de manutention des ensembles actifs ou contamines, les principes et les moyens qui regissent la conduite du reacteur, les fonctions et l'implantation des divers batiments. La divergence de RAPSODIE est prevue pour 1964. (auteurs)

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

    International Nuclear Information System (INIS)

    Holmes, J.A.G.

    1982-01-01

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

  7. Experience of secondary cooling system modification at fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Ito, Keisuke; Nakatsuji, M.; Matsuno, Hiroki; Matsui, K.; Tone, T.

    2007-01-01

    The prototype fast breeder reactor MONJU has been shut down since the secondary sodium leak accident which occurred in December 1995. After the accident, the investigation of its cause and the comprehensive review were performed and the various counter measures against the sodium leak were also discussed. The main modification works of MONJU started in September 2005. The work should adopt suitable methods to treat sodium, since MONJU uses chemically active sodium as a coolant. Considering the chemical activity of sodium, MONJU learned the modification methods from the experimental fast reactor JOYO and precedent plants of overseas and adopted plastic bags when the sodium boundary is opened, management of oxygen concentration in the plastic bags, a slightly positive control of the cover gas pressure, compress cut by the roll cutters to prevent the entry of the chips, etc.. Owing to introduction of these methods, the modification works have proceeded almost on schedule without troubles. (author)

  8. Experience of secondary cooling system modification at prototype fast breeder reactor MONJU (Translated document)

    International Nuclear Information System (INIS)

    Kisohara, Naoyuki; Sakamoto, Yoshihiko

    2010-09-01

    The prototype fast breeder reactor MONJU has been shut down since the secondary sodium leak accident that occurred in December 1995. After the accident, an investigation into the cause and a comprehensive safety review of the plant were conducted, and various countermeasures for sodium leak were examined. Modification work commenced in September 2005. Since sodium, a chemically active material, is used as coolant in MONJU, the modification work required work methods suitable for the handling of sodium. From this perspective, the use of a plastic bag when opening the sodium boundary, oxygen concentration control in a plastic bag, slightly-positive pressure control of cover gas in the systems, pressing and cutting with a roller cutter to prevent the incorporation of metal fillings, etc. were adopted, with careful consideration given to experience and findings from previous modification work at the experimental fast reactor JOYO and plants abroad. Owing to these work methods, the modification work proceeded close to schedule without incident. (author)

  9. Steam condenser optimization using Real-parameter Genetic Algorithm for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Jayalal, M.L.; Kumar, L. Satish; Jehadeesan, R.; Rajeswari, S.; Satya Murty, S.A.V.; Balasubramaniyan, V.; Chetal, S.C.

    2011-01-01

    Highlights: → We model design optimization of a vital reactor component using Genetic Algorithm. → Real-parameter Genetic Algorithm is used for steam condenser optimization study. → Comparison analysis done with various Genetic Algorithm related mechanisms. → The results obtained are validated with the reference study results. - Abstract: This work explores the use of Real-parameter Genetic Algorithm and analyses its performance in the steam condenser (or Circulating Water System) optimization study of a 500 MW fast breeder nuclear reactor. Choice of optimum design parameters for condenser for a power plant from among a large number of technically viable combination is a complex task. This is primarily due to the conflicting nature of the economic implications of the different system parameters for maximizing the capitalized profit. In order to find the optimum design parameters a Real-parameter Genetic Algorithm model is developed and applied. The results obtained are validated with the reference study results.

  10. Design and development of a zoom lens objective for the fast breeder test reactor periscope

    International Nuclear Information System (INIS)

    Das, N.C.; Udupa, D.V.; Shukla, R.P.

    2003-10-01

    A three lens optically compensated zoom lens useful for the 5 meter long periscope in the Fast Breeder Test Reactor (FBTR) has been designed, fabricated and tested. The zoom lens fabricated using radiation resistant glasses has a zoom ratio of 2.5 with a focal length range of l00 mm to 250 mm. The zoom lens objective has been designed for viewing the objects kept at a distance in the range of 1.5 m to 3 m from the objective lens. It is found that the zoom lens objective can be used for resolving objects with a linear resolution of 0.2 mm inside the reactor when viewed with an eye piece of focal length 50 mm. (author)

  11. Steam condenser optimization using Real-parameter Genetic Algorithm for Prototype Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Kumar, L. Satish, E-mail: satish@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Jehadeesan, R., E-mail: jeha@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Rajeswari, S., E-mail: raj@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Satya Murty, S.A.V., E-mail: satya@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Balasubramaniyan, V.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India)

    2011-10-15

    Highlights: > We model design optimization of a vital reactor component using Genetic Algorithm. > Real-parameter Genetic Algorithm is used for steam condenser optimization study. > Comparison analysis done with various Genetic Algorithm related mechanisms. > The results obtained are validated with the reference study results. - Abstract: This work explores the use of Real-parameter Genetic Algorithm and analyses its performance in the steam condenser (or Circulating Water System) optimization study of a 500 MW fast breeder nuclear reactor. Choice of optimum design parameters for condenser for a power plant from among a large number of technically viable combination is a complex task. This is primarily due to the conflicting nature of the economic implications of the different system parameters for maximizing the capitalized profit. In order to find the optimum design parameters a Real-parameter Genetic Algorithm model is developed and applied. The results obtained are validated with the reference study results.

  12. Application of a bistable convection loop to LMFBR [liquid metal fast breeder reactor] emergency core cooling

    International Nuclear Information System (INIS)

    Anand, G.; Christensen, R.N.

    1990-01-01

    The concept of passive safety features for nuclear reactors has been developed in recent years and has gained wide acceptance. A literature survey of current reactors with passive features indicates that these reactors have some passive features but still do not fully meet the design objectives. Consider a current liquid-metal reactor design like PRISM. During normal operation, liquid sodium enters the reactor at ∼395 degree C and exits at ∼550 degree C. In the event of loss of secondary cooling with or without scram, the primary coolant (liquid sodium) initially acts as a heat sink and its temperature increases. For events without scram, the negative reactivity induced by the increase in temperature shuts the reactor down. When the average temperature of the sodium reaches ∼600 to 650 degree C, it overflows from the reactor vessel, activating the auxiliary cooling system. The auxiliary cooling system uses natural circulation of air around the reactor guard vessel. An alternative to the current design incorporates a bistable convection loop (BCL). The incorporation of the BCL concept remarkably improves the safety of the nuclear reactors. Application of the BCL concept to liquid-metal fast breeder reactors is described in this paper

  13. Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.i [Indira Gandhi Centre for Atomic Research (IGCAR), Department of Atomic Energy (DAE), Kalpakkam, Tamilnadu 603102 (India); Raj, Baldev, E-mail: dir@igcar.gov.i [Indira Gandhi Centre for Atomic Research (IGCAR), Department of Atomic Energy (DAE), Kalpakkam, Tamilnadu 603102 (India)

    2011-03-15

    Research highlights: Production of thin fluoroelastomer profiles by cold feed extrusion and continuous cure involving microwave and hot air heating. Use of peroxide curing in air during production. Use of fluoroelastomers based on advanced polymer architecture (APA) for the production of profiles. Use of the profiles in inflatable seals for critical application of Prototype Fast Breeder Reactor. Tailoring of material formulation by synchronized optimization of material and production technologies to ensure that the produced seal ensures significant gains in terms of performance and safety in reactor under synergistic influences of temperature, radiation, air and sodium aerosol. - Abstract: The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of {approx}2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 {sup o}C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for other large

  14. Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Sinha, N.K.; Raj, Baldev

    2011-01-01

    Research highlights: → Production of thin fluoroelastomer profiles by cold feed extrusion and continuous cure involving microwave and hot air heating. → Use of peroxide curing in air during production. → Use of fluoroelastomers based on advanced polymer architecture (APA) for the production of profiles. → Use of the profiles in inflatable seals for critical application of Prototype Fast Breeder Reactor. → Tailoring of material formulation by synchronized optimization of material and production technologies to ensure that the produced seal ensures significant gains in terms of performance and safety in reactor under synergistic influences of temperature, radiation, air and sodium aerosol. - Abstract: The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of ∼2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 o C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for

  15. A model of gas cavity breakup behind a blockage in fast breeder reactor subassembly geometry

    International Nuclear Information System (INIS)

    Fukuzawa, Y.

    1980-05-01

    A semi-empirical model has been developed to describe the transient behaviour of a gas cavity due to breakup behind a blockage in Liquid Metal Fast Breeder Reactor subassembly geometry. The main mechanisms assumed for gas cavity breakup in the present model are as follows: The gas cavity is broken up by the pressure fluctuation at the interface due to turbulence in the liquid. The centrifugal force on the liquid opposes breakup. The model is able to describe experimental results on the transient behaviour of a gas cavity due to breakup after the termination of gas injection. On the basis of the present model the residence time of a gas cavity behind a blockage in sodium is predicted and the dependence of the residence time on blockage size is discussed. (orig.) [de

  16. Heat transfer from the roughened surface of gas cooled fast breeder reactor fuel element

    International Nuclear Information System (INIS)

    Tang, I.M.

    1979-01-01

    The temperature distributions and the augmentation of heat transfer performance by artificial roughening of a gas cooled fast breeder reactor (GCFR) fuel rod cladding are studied. Numerical solutions are based on the axisymmetric assumption for a two-dimensional model for one rib pitch of axial distance. The local and axial clad temperature distributions are obtained for both the rectangular and ramp rib roughened surface geometries. The transformation of experimentally measured convective heat transfer coefficients, in terms of Stanton number, into GCFR values is studied. In addition, the heat transfer performance of a GCFR fuel rod cladding roughened surface design is evaluated. Approximate analytical solution for correlating an average Stanton number is also obtained and satisfactorily compared with the corresponding numerical result for a GCFR design. The analytical correlation is useful in assessing roughened surface heat transfer performance in scoping studies and conceptual design

  17. Fission and corrosion product behaviour in liquid metal fast breeder reactors (LMFBRs)

    International Nuclear Information System (INIS)

    1993-02-01

    It is intended that this review will be useful not only to scientists but also to those concerned with design, day-to-day operation of plant, with liquid metal fast breeder reactors (LMFBRs), safety and decommissioning. Because of this, the review has been widened to include not only the mass transfer behaviour of the various radionuclides in experimental and operating systems, but also the monitoring of the various species, the methods of measurement and the development of methods to control the build-up of the more important long half-life species in operating plants. The information used in the review has been taken from open literature sources to provide an up-to-date presentation of the behaviour of the various isotopes in LMFBRs. 172 refs, 14 figs, 22 tabs

  18. Simplified inelastic analysis methods applied to fast breeder reactor core design

    International Nuclear Information System (INIS)

    Abo-El-Ata, M.M.

    1978-01-01

    The paper starts with a review of some currently available simplified inelastic analysis methods used in elevated temperature design for evaluating plastic and thermal creep strains. The primary purpose of the paper is to investigate how these simplified methods may be applied to fast breeder reactor core design where neutron irradiation effects are significant. One of the problems discussed is irradiation-induced creep and its effect on shakedown, ratcheting, and plastic cycling. Another problem is the development of swelling-induced stress which is an additional loading mechanism and must be taken into account. In this respect an expression for swelling-induced stress in the presence of irradiation creep is derived and a model for simplifying the stress analysis under these conditions is proposed. As an example, the effects of irradiation creep and swelling induced stress on the analysis of a thin walled tube under constant internal pressure and intermittent heat fluxes, simulating a fuel pin, is presented

  19. Online monitoring of steam/water chemistry of a fast breeder test reactor

    International Nuclear Information System (INIS)

    Subramanian, K.G.; Suriyanarayanan, A.; Thirunavukarasu, N.; Naganathan, V.R.; Panigrahi, B.S.; Jambunathan, D.

    2005-01-01

    Operating experience with the once-through steam generator of a fast breeder test reactor (FBTR) has shown that an efficient water chemistry control played a major role in minimizing corrosion related failures of steam generator tubes and ensuring steam generator tube integrity. In order to meet the stringent feedwater and steam quality specifications, use of fast and sensitive online monitors to detect impurity levels is highly desirable. Online monitoring techniques have helped in achieving feedwater of an exceptional degree of purity. Experience in operating the online monitors in the steam/water system of a FBTR is discussed in detail in this paper. In addition, the effect of excess hydrazine in the feedwater on the steam generator leak detection system and the need for a hydrazine online meter are also discussed. (orig.)

  20. International review on safety requirements for the prototype fast breeder reactor “Monju”

    International Nuclear Information System (INIS)

    2016-01-01

    In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

  1. Evaluation of neutron streaming in fast breeder reactor fuel assembly by double heterogeneous modelling

    International Nuclear Information System (INIS)

    Unesaki, Hironobu; Takeda, Toshikazu

    1988-01-01

    Neutron streaming in a fast breeder reactor fuel assembly caused by the double heterogeneity structure is estimated by double heterogeneous modelling. The conventional pin cell model, a two-region subassembly model and the exact pin cluster model are used to take into account the streaming effect caused by the pin cell structure and the surrounding wrapper tube structure. The heterogeneity of wrapper tube and its surrounding sodium is explicitly considered. The streaming effect is evaluated based on Benoist's diffusion coefficient. The total streaming effect caused by the double heterogeneity structure of a fuel subassembly is found to be -0.2 % dk/kk' for k eff , which is almost twice that obtained from the conventional pin cell model of -0.1 % dk/kk'. (author)

  2. Application of acoustic agglomerators for emergency use in liquid-metal fast breeder reactor plants

    International Nuclear Information System (INIS)

    Shaw, D.T.; Rajendran, N.

    1979-01-01

    The use of acoustic agglomerators for the suppression of sodium-fire aerosols in the case of a hypothetical core disruptive accident of a liquid-metal fast breeder reactor is discussed. The basic principle for the enhancement of agglomeration of airborne particles under the influence of an acoustic field is first discussed, followed by theoretical predictions of the optimum operating conditions for such application. It is found that with an acoustic intensity of 160 dB (approx. 1 W/cm 2 ), acoustic agglomeration is expected to be several hundred times more effective than gravitational agglomeration. For particles with a radius larger than approx. 2 μm, hydrodynamic interaction becomes more important than the inertial capture. For radii between 0.5 and 2 μm, both mechanisms have to included in the theoretical predictions of the acoustic agglomeration rate

  3. International review on safety requirements for the prototype fast breeder reactor “Monju” (Translated document)

    International Nuclear Information System (INIS)

    2016-02-01

    In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

  4. Identification and adoptive multivariable control method. Application to fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Dang Van Mien, H.; Irving, E.; Rousseau, I.

    1982-01-01

    Motivated by the limitations inherent in the standard approach, a new model reference multivariable adaptive control method is described. This latter control method utilizes as a design tool a simple vector difference equation of the controlled system. The adaptive control method is of the series-parallel direct reference model type and the adjustment algorithm is the standard least squares estimation technics with hyperstability conditions, controlled convergence speed forgetting, regularization and threshold operations. Numerical results are presented which illustrate the interest of the latter approach. The precise problem which has been tackled is the control of the steam generator of the second fast breeder French nuclear reactor called Super-Phenix. After a short description of the plant and its responses at different loads, the principles and the performances of the standard technique control scheme which has been adopted are discussed [fr

  5. Selection of hardfacing material for components of the Indian Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Bhaduri, A.K.; Indira, R.; Albert, S.K.; Rao, B.P.S.; Jain, S.C.; Asokkumar, S.

    2004-01-01

    Nickel-base hardfacing alloys have been chosen to replace cobalt-base alloys as hardfacing material for components of the Indian Prototype Fast Breeder Reactor, for minimising the dose rate to personnel during maintenance and decommissioning, and to reduce the shielding thickness required for component handling. Induced activity, dose rate and shielding computations showed that replacing cobalt-base alloys with nickel-base alloys for hardfacing of components would result in a marked reduction in both the dose rate from the components and the thickness of lead handling flasks. Long-term ageing studies on the nickel-base hardface deposits on austenitic stainless steel showed that the hardface deposit would retain adequate hardness at the end of the components' design service-life of 40 years of exposure at 823 K

  6. Reliability analysis of diverse safety logic systems of fast breeder reactor

    International Nuclear Information System (INIS)

    Ravi Kumar, Bh.; Apte, P.R.; Srivani, L.; Ilango Sambasivan, S.; Swaminathan, P.

    2006-01-01

    Safety Logic for Fast Breeder Reactor (FBR) is designed to initiate safety action against Design Basis Events. Based on the outputs of various processing circuits, Safety logic system drives the control rods of the shutdown system. So, Safety Logic system is classified as safety critical system. Therefore, reliability analysis has to be performed. This paper discusses the Reliability analysis of Diverse Safety logic systems of FBRs. For this literature survey on safety critical systems, system reliability approach and standards to be followed like IEC-61508 are discussed in detail. For Programmable Logic device based systems, Hardware Description Languages (HDL) are used. So this paper also discusses the Verification and Validation for HDLs. Finally a case study for the Reliability analysis of Safety logic is discussed. (author)

  7. Development of ferritic steels for steam generators of fast breeder reactors

    International Nuclear Information System (INIS)

    Nguyen-Thanh; Vigneron, G.; Vanderschaeghe, A.

    1988-01-01

    STEIN INDUSTRIE, a manufacturer of equipment for the conventional and nuclear power industry, has built up expertise in the use of Cr-Mo steels used at high temperatures. The main ferritic steels developed were 10 CD 9-10 (AFNOR), Z10 CDNb V 9-2 (AFNOR), X 20 Cr Mo V 12-1 (DIN) and ASTM Grade 9.1. For the fast breeder reactor system, STEIN INDUSTRIE proposes the use of these steels in the construction of steam generators. The wide programme of development undertaken by STEIN INDUSTRIE is aimed at the following main subjects: - characterization of materials - welding and bending tests - studies of special junctions. This article reports the results obtained

  8. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    International Nuclear Information System (INIS)

    Geraldo, Issa Cherif; Bose, Tanmoy; Pekpe, Komi Midzodzi; Cassar, Jean-Philippe; Mohanty, A.R.; Paumel, Kévin

    2014-01-01

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected

  9. Annual technical report of the prototype fast breeder reactor Monju. 2011

    International Nuclear Information System (INIS)

    2012-08-01

    The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the primary achievements and the data related to the plant management in Monju during fiscal 2011. From the aspect of the design evaluation, the following items are summarized: 1) the evaluation of the decay heat removal of Monju core by natural convection, and the safety measures against earthquake and tsunami, which were carried out from the lessons learned at the Fukushima-daiichi accident due to the Great East Japan Earthquake on March 11, 2011, 2) the control rod worth confirmation and the evaluation of nuclear data library based on the data of Core Confirmation Test, which is the first step of Monju system startup test restarted in 2010, 3) the evaluation of the hydrogen concentration behavior, which detects the leak of water from the heat transfer tube of steam generator. Then, from the aspect of the maintenance technology, the following items are summarized: 1) the results of the function confirmation test on the water/steam system, after the long-term suspension, 2) confirmation of the integrity of cracked cylinder liners of emergency diesel generator, 3) replacement of the annulus ventilation duct, 4) evaluation of reduction of the periodic inspection schedule after full power operation. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

  10. Development of a nuclear information system for the MONJU Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Metz, Peter

    2000-01-01

    At the MONJU Fast Breeder Reactor information is collected on a continuous basis. This information consists of measured data, design data, simulations data, maintenance data etc. which may be in any kind of electronic form, i.e. text documents, code input files, reports or even scanned documents. The amount and nature of these data has introduced the need for a software system, which will provide an efficient infrastructure for the maintenance of and operations on the data. Thus a Nuclear Information System for the MONJU Fast Breeder Reactor is under development. The system consists of remote databases hosting the information along with clients for handling them, remote clients providing the users with an interface and a local server for handling the client requests and the communication between the database and user clients. The system is composed of independent server, database and user modules, which communicates using the RMI-IIOP (Remote Method Invocation - Internet InterORB Protocol) technology. The RMI-IIOP is a CORBA (Common Object Request Broker Architecture) compliant subset of the RMI thereby facilitating the possibility of implementing the database and user modules in any kind of programming language and on any kind of operating system by providing a standard, platform independent communications interface. The user interface consists of dynamic HTML web pages which instantiates servlets in the user module when the user submits queries. The database module consists of controllers for handling the communication with the user module and database drivers for handling the connections with the databases. In this paper the overall system design and schemes for data flow and remote method invocations are presented and the requirements imposed on the system are discussed. (author)

  11. 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)

  12. BN-350 decommissioning problems of radioactive waste management

    International Nuclear Information System (INIS)

    Galkin, A.; Tkachenko, V.

    2002-01-01

    Pursuant of modern concept on radioactive waste management applied in IAEA Member States all radioactive wastes produced during the BN-350 operation and decommissioning are subject to processing in order to be transformed to a form suitable for long-term storage and final disposal. The first two priority objectives for BN-350 reactor are as follows: cesium cleaning from sodium followed by sodium drain, and processing; processing of liquid and solid radioactive waste accumulated during BN-350 operation. Cesium cleaning from sodium and sodium processing to NaOH will be implemented under USA engineering and financial support. However the outputted product might be only subject to temporary storage under special conditions. Currently the problem is being solved on selection of technology for sodium hydroxide conversion to final product incorporated into cement-like matrix ready for disposal pursuant to existing regulatory requirements. Industrial installation is being designed for liquid radioactive waste processing followed by incorporation to cement matrix subject to further disposal. The next general objective is management of radioactive waste expected from BN-350 decommissioning procedure. Complex of engineering-radiation investigation that is being conducted at BN-350 site will provide estimation of solid and liquid radioactive waste that will be produced during the course of the BN-350 decommission. Radioactive wastes that will be produced may be shared for primary (metal structures of both reactor and reactor plant main and auxiliary systems equipment as well as construction wastes of dismantled biological protection, buildings and structures) and secondary (deactivation solutions, tools, materials, cloth, special accessory, etc.). Processing of produced radioactive wastes (including high activity waste) requires the use of special industrial facilities and construction of special buildings and structures for arrangement of facilities mentioned as well as for

  13. The simulation of the process of sodium freezing in the tubes for the optimization of fast breeder reactor units maintenance

    International Nuclear Information System (INIS)

    Tashlykov, O.L.; Shcheklein, S.E.; Annikov, S.V.

    2013-01-01

    The peculiarities of the repair works of the fast breeder reactor sodium systems are considered. The requirements for the sodium melting exclusion inside the equipment and piping during their opening and repair are given. The results of the sodium cooling process simulation with SolidWorks software are also described [ru

  14. Development of safety evaluation methods applied to the safety regulations for the operation stage of fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purposes of this study is to establish the safety evaluation methods needed in the operation stage of fast breeder reactor (FBR). Related the above purpose of this study, some investigation studies for the technical standard applied to Monju were achieved in JFY 2012. (author)

  15. Fast breeder project (PSB)

    International Nuclear Information System (INIS)

    1976-07-01

    Activities performed during the 1st quarter of 1976 at or on behalf of the Gesellschaft fuer Kernforschung mbH, Karlsruhe, within the framework of the Fast Breeder Project are given a survey. The following project subdivisions are dealt with: Fuel rod development; materials testing and developments; corrosion studies and coolant analyses; physical experiments; reactor theory; safety of fast breeders; instrumentation and signal processing for core monitoring; effects on the environment; sodium technology tests; thermodynamic and fluid flow tests in gas. (HR) [de

  16. Development of components for the gas-cooled fast breeder reactor program

    International Nuclear Information System (INIS)

    Dee, J.B.; Macken, T.

    1977-01-01

    The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

  17. Status and development programme of the fast breeder reactor system in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Daeunert, U.; Kessler, G.

    1977-01-01

    The Federal Republic of Germany depends on imports of raw materials for energy purposes, and special importance is therefore given to nuclear energy for reasons of energy policy. Against the background of a nuclear power station capacity amounting to 24,000MW(e) in 1976 (power stations being either on order, under construction or actually in operation), the fast breeder reactor (FBR) is outstandingly significant in the Federal Republic for the economic use both of the uranium and the plutonium accumulating in light-water reactors (LWRs). Work in this sector started in the FRG in 1960. Important stages in development were: the initial activities on the plutonium critical SNEAK at Karlsruhe; successful operation of the KNK 20-MW(e) experimental reactor (sodium-cooled) with the KNK II (plutonium core) follow-up stage; costruction of the 300-MW(e) FBR prototype nuclear power station at Kalkar; and initial planning carried out for a 1300-MW(e) FBR demonstration nuclear power station. A viable infrastructure for safety assessment must go hand in hand with the development of new reactor systems. Prototype reactors in the FRG are therefore already subjected to the same strict licensing procedures under nuclear legislation as are commercial nuclear power stations. Several years' practical experience have shown that this was the right approach. It was recognized in the FRG early on that major long-term technological projects could no longer be implemented by unaided single nations. The SNR-300 FBR prototype project combines the efforts of the governments, power supply companies, reactor manufacturers and nuclear research centres of the FRG, Belgium and the Netherlands. The agreement signed in 1971 by the Rheinisch-Westfaelische Elektrizitaetswerke AG (FRG), Ente Nazionale per l'Energia Elettrica (Italy) and Electricite de France (France) power supply companies on the joint construction of the Super-Phenix and SNR-2 FBR demonstration nuclear power stations, and the German

  18. 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

  19. A strategy analysis of the fast breeder reactor introduction and nuclear fuel cycle systems deployment

    International Nuclear Information System (INIS)

    Wajima, Tsunetaka; Kawashima, Katsuyuki; Yamashita, Takashi

    1996-01-01

    A study is made on a strategy analysis of the long term nuclear fuel cycle systems deployment in accordance with the nuclear power growth projection and fast breeder reactor (FBR) introduction. In the analysis, the reprocessed plutonium (Pu) is charged into the reactor in such a way that the reprocessed Pu is not stored outside the reactor, i.e., there is no excess Pu outside the reactor. The analysis characterized the fuel cycle systems, and showed the usefulness of the present method to determine future directions for the FBR introduction and nuclear fuel cycle systems deployment. Concerning an intermediate-term strategy, the time of introduction and required capacities of a second commercial LWR reprocessing plant, Pu-thermal, and the first FBR reprocessing plant deployment are evaluated. A long term strategy analysis shows that the two or three large plants are run in parallel for each fuel cycle facility and that FBR related facilities deal with a markedly large amount of Pu. It is concluded that the early stage introduction of FBRs of significant capacities seems necessary to materialize a consistent total FBR/fuel cycle system where Pu balance becomes feasible through its flexible operation of, for instance, adjusting breeding ratio, in order to keep the transparency of the Pu utilization. (author)

  20. Neutronics studies on the feasibility of developing fast breeder reactor with flexible breeding ratio

    International Nuclear Information System (INIS)

    Xiao Yunlong; Wu Hongchun; Zheng Youqi; Wang Kunpeng

    2016-01-01

    This paper investigates the feasibility of designing a flexible fast breeder reactor from the view of neutronics. It requires that the variable breeding ratio can be achieved in operating a fast reactor without significant changes of the core, including the minimum change of fuel assembly design, the minimum change of the core configuration and the same control system arrangement in the core. The sodium cooled fast reactor is investigated. Two difficulties are overcome: (1) the different excess reactivity is well controlled for different cores, especially for the one with small breeding ratio; (2) the maximum linear power density is well controlled while the breeding ratio changes. The optimizations are done to meet the requirements. The U–Pu–Zr alloy is applied to enhance the breeding. The enrichment-zoning technique with unfixed blanket assembly loading position is searched to get acceptable power distributions when the breeding ratio changes. And the control system is designed redundantly to fulfill the control needs. Then, the achieved breeding ratio can be adjusted from 1.1 to 1.4. The reactivity coefficients, temperature distributions and preliminary safety performances are evaluated to investigate the feasibility of the new concept. All the results show that it is feasible to develop the fast reactor with flexible breeding ratios, although it still highly relies on the advancement of the coolant flow control technology. (author)

  1. The DSNP simulation language and its application to liquid-metal fast breeder reactor transient analyses

    International Nuclear Information System (INIS)

    Saphier, D.; Madell, J.T.

    1982-01-01

    A new, special purpose block-oriented simulation language, the Dynamic Simulator for Nuclear Power Plants (DSNP), was used to perform a dynamic analysis of several conceptual design studies of liquid metal fast breeder reactors. The DSNP being a high level language enables the user to transform a power plant flow chart directly into a simulation program using a small number of DSNP statements. In addition to the language statements, the DSNP system has its own precompiler and an extensive library containing models of power plant components, algorithms of physical processes, material property functions, and various auxiliary functions. The comparative analysis covered oxide-fueled versus metal-fueled core designs and loop- versus pool-type reactors. The question of interest was the rate of change of the temperatures in the components in the upper plenum and the primary loop, in particular the reactor outlet nozzle and the intermediate heat exchanger inlet nozzle during different types of transients. From the simulations performed it can be concluded that metal-fueled cores will have much faster temperature transients than oxide-fueled cores due mainly to the much higher thermal diffusivity of the metal fuel. The transients in the pool-type design (either with oxide fuel or metal fuel) will be much slower than in the loop-type design due to the large heat capacity of the sodium pool. The DSNP language was demonstrated to be well suited to perform many types of transient analysis in nuclear power plants

  2. Interim report of the subcommittee on putting fast breeder reactors in practical use

    International Nuclear Information System (INIS)

    1983-01-01

    Light water reactors are expected to be the main power reactors in Japan for considerably long period hereafter, however, the basic policy of long term energy utilization is to use the plutonium and depleted uranium recovered by the reprocessing of spent fuel for fast breeder reactors, in view of the effective use of energy resources and the energy security. At present, the experimental FBR ''Joyo'' is in operation, and the preparation work was started for the prototype FBR ''Monju'', aiming at the criticality at the end of fiscal 1990. In the discussion, it was confirmed that for the practical use of FBRs, the improvement of their economy is the most important. But the construction cost for the Monju and the type of the demonstration FBR are not yet decided, and there are many uncertain elements. The demand and supply of uranium and the prospect of FBR development, the appropriate way of FBR development in Japan, the trend of FBR development in Japan and foreign countries, the evaluation of the economy of FBRs and the time of putting in practical use, the schedule of FBR development, the cost reduction in FBRs, the fuel cycle and others are reported. (Kako, I.)

  3. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  4. Real Time Computer for Plugging Indicator Control of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Manimaran, M.; Manoj, P.; Shanmugam, A.; Murali, N.; Satya Murty, S.A.V.

    2013-06-01

    Prototype Fast Breeder Reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Liquid sodium is used as coolant to transfer the heat produced in the reactor core to steam water circuit. Impurities present in the sodium are removed using purification circuit. Plugging indicator is a device used to measure the purity of the sodium. Versa Module Europa bus based Real Time Computer (RTC) system is used for plugging indicator control. Hot standby architecture consisting of dual redundant RTC system with switch over logic system is the configuration adopted to achieve fault tolerance. Plugging indicator can be controlled in two modes namely continuous and discontinuous mode. Software based Proportional-Integral-Derivative (PID) algorithms are developed for plugging indicator control wherein the set point changes dynamically for every scan interval of the RTC system. Set points and PID constants are kept as configurable in runtime in order to control the process in very efficient manner, which calls for reliable communication between RTC system and control station, hence TCP/IP protocol is adopted. Performance of the RTC system for plugging indicator control was thoroughly studied in the laboratory by simulating the inputs and monitored the control outputs. The control outputs were also monitored for different PID constants. Continuous and discontinuous mode plots were generated. (authors)

  5. Investigation of safety measures to severe accident of Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    So as to plan the accident management to severe accident of Fast Breeder Reactor (FBR), it is primary important to understand the progression of severe accident (SA) precisely. In this study, it has been aimed to reveal two items that work as keys in the evaluation of SA in sodium cooled FBR. One is the cool-ability of degraded core on the core support plate by sodium natural circulation in the post accident heat removal (PAHR) phase. An obstacle that hinders the smooth heat transfer from fuel debris to coolant is the formation of sodium-uranate by chemical reaction between sodium and fuel. Following the measurement of physical values of sodium-uranate in FY 2011, experiments has been performed to reveal the conditions for sodium-uranate formation on fuel debris in sodium pool simulating the actual situation of the degraded core. The cool-ability of the debris bed was analyzed using the Lipinski 1-D model. Another research performed in this study is the measurement of fission product (cesium and antimony) evaporation rates from FBR fuel as a function of temperature, because presently the fission product evaporation rates data for LWR is also temporarily used for FBR SA analysis. The measurement was performed using the irradiated fuels in the Test Reactor JOYO. (author)

  6. Evolution of actinides in ThO2 blanket of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Bachchan, Abhitab; Riyas, A.; Devan, K.; Puthiyavinayagam, P.

    2015-01-01

    The third stage of India's nuclear program focuses on fissile fuel production through Th- 233 U cycle in view of the better abundance and relative merits of thorium. For early introduction of Thorium into the nuclear energy system, several R and D program has started to find the best possible route of thorium utilization. Towards this, efforts were made to assess the feasibility of Th-U cycle in a fast spectrum reactor like Prototype Fast Breeder Reactor (PFBR). The effect on core neutronic parameters and actinide evolution with the replacement of depleted UO 2 in the PFBR blanket SA with thorium oxide has been studied using 3-D diffusion code FARCOB. Study shows that by the introduction of thorium blanket, core excess reactivity is coming down by ∼ 535 pcm and core breeding ratio is slightly lower than conventional oxide blanket. The distribution of region wise power production is slightly changed. Power from radial blanket is reduced from 3% to 2% while the core-1 power is increased from 49 % to 50 %. The estimated 233 U production is 7.6, 11.5 and 14.1 kg/t with 180, 360 and 540 days of irradiation respectively. (author)

  7. Behaviour and microstructure of stainless steels irradiated in the french fast breeder reactors

    International Nuclear Information System (INIS)

    Dubuisson, P.; Gilbon, D.

    1991-01-01

    The burn-up of Fast Breeder Reactors is limited by the irradiation induced dimensional changes and mechanical properties of structural materials used for replaceable in-core components. This paper describes the behaviour improvements and also the radiation-induced microstructures of the different steels used for fuel pin cladding and wrapper tubes in French reactors. Materials of fuel pin cladding are austenitic steels whose main problem is swelling. Improvements in swelling resistance by cold-working, titanium additions and modifications of matrix (Fe-Cr-Ni) from SA 316 to CW 15-15 Ti are shown. These improvements are correlated with a best stability of microstructure under irradiation. Beneficial effects of phosphorus addition and multistabilisation (NbVTi) on radiation induced microstructure and swelling resistance are also shown. Austenitic steels used for wrapper tubes are limited both by swelling and by void embrittlement. The ferritic F17 (17Cr), ferritic-martensitic EM12 (9Cr-2MoNbV) and martensitic EM10 (9Cr-1Mo) steels present high swelling resistance. Nevertheless radiation-induced embrittlement is observed in EM12 and especially in F17. This embrittlement results from a fine and uniform radiation enhanced precipitation in ferrite grains. By contrast, the microstructure of fully martensitic EM10 steel is mush more stable and its ductile-brizzle transition temperature stays below 0 0 C. 12 figs

  8. Methods of preventing fast breeder reactor shield plug from adhesion of sodium

    International Nuclear Information System (INIS)

    Hashiguchi, Koh; Hara, Johji; Nei, Hiromichi; Daiku, Motoichi; Wagatsuma, Kenji

    1980-01-01

    The shield plug, which is located at the upper part of a reactor vessel of a sodium-cooled fast breeder reactor, is composed of a rotating and a stationary plug. Fuel exchange is performed easily by the rotation of the rotating plug. The vapor or mist of sodium evaporated from liquid sodium deposits on the gap surfaces of the rotating and stationary plugs and is solidified or changed into a solid reactant. If such condition continues for a long period, harmful effects are exerted on the fuel exchange operation. In order to develop methods of preventing the sodium deposition, investigation was made on the phenomenon of sodium deposition. By the use of the testing equipment simulating the shield plug, deposition tests and specimen measurements were made for different gap width test section size and condition. On the basis of the effects of these parameters clarified by experiments, the effectiveness of three kinds of mechanism for preventing sodium deposition were investigated experimentally. In addition, by using a thermo-siphon analogical model, analysis was performed to deduce experimental equations for sodium deposition. (author)

  9. A new physics design of control safety rods for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Devan, K.; Riyas, A.; Alagan, M.; Mohanakrishnan, P.

    2008-01-01

    The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B 4 C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B 4 C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B 4 C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B 4 C sections in top and bottom. There is significant savings in the initial inventory of enriched B 4 C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles

  10. Status of Fast Breeder Reactor Development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    Hüper, R.

    1988-01-01

    In 1967 and 1968, the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (''DeBeNe'') agreed to develop breeder reactors in a joint program. The following research organizations have taken part in this effort: - Kernforschungszentrum Karlsruhe (KfK) - Interatom, Bergisch Gladbach Alkem, Hanau - SCK/CEN, Mol - Belgonucleaire, Brussels - ECN, Petten - TNO, Apeldoorn - Neratoom, The Hague. The first three institutions mentioned above have been associated in the Entwicklungsgemeinschaft Schneller Brüter since 1977. KfK, INTERATOM, and the French Commissariat à l'Energie Atomique entered into contracts in 1977 about close cooperation in the fast breeder field, to which the Belgian and Dutch partners acceded. The results of activities carried out by the DeBeNe partners in 1987 have been compiled in this report. The report begins with a survey of the fast reactor plants, which is followed by an R&D summary. In an additional chapter, a survey is given of international cooperation in 1987

  11. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Veerasamy, R.; Patri, Sudheer; Ignatius Sundar Raj, S.; Kumar Krovvidi, S.C.S.P.; Dash, S.K.; Meikandamurthy, C.; Rajan, K.K.; Puthiyavinayagam, P.; Chellapandi, P.; Vaidyanathan, G.; Chetal, S.C.

    2010-01-01

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO 2 process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

  12. The detection of sodium vapor bubble collapse in a liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Carey, W.M.; Gavin, A.P.; Bobis, J.P.; Sheen, S.H.; Anderson, T.T.; Doolittle, R.D.; Albrecht, R.W.

    1977-01-01

    Sodium boiling detection utilizing the sound pressure emanated during the collapse of a sodium vapour bubble in a subcooled media is discussed in terms of the sound characteristic, the reactor ambient noise background, transmission loss considerations and performance criteria. Data obtained in several loss of flow experiments on Fast Test Reactor Fuel Elements indicate that the collapse of the sodium vapour bubble depends on the presence of a subcooled structure or sodium. The collapse pressure pulse was observed in all cases to be on the order of a kPa, indicating a soft type of cavitational collapse. Spectral examination of the pulses indicates the response function of the test structure and geometry is important. The sodium boiling observed in these experiments was observed to occur at a low ( 0 C) liquid superheat with the rate of occurrence of sodium vapor bubble collapse in the 3 to 30 Hz range. Reactor ambient noise data were found to be due to machinery induced vibrations flow induced vibrations, and flow noise. These data were further found to be weakly stationary enhancing the possibility of acoustic surveillance of an operating Liquid Metal Fast Breeder Reactor. Based on these noise characteristics and extrapolating the noise measurements from the Fast Flux Test Facility Pump (FFTP), one would expect a signal to noise ratio of up to 20 dB in the absence of transmission loss. The requirement of a low false alarm probability is shown to necessitate post detection analysis of the collapse event sequence and the cross correlation with the second derivative of the neutronic boiling detection signal. Sodium boiling detection using the sounds emitted during sodium vapor bubble collapse are shown to be feasible but a need for in-reactor demonstration is necessary. (author)

  13. Investigation on in-vessel thermal transients in a fast breeder reactor

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Kasahara, Naoto

    1999-01-01

    Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of the phenomena in the design of the internal structures in an LMFBR plenum. To evaluate thermal stress characteristics for the inner barrel in a typical LMFBR upper plenum, numerical analysis was carried out with a multi-dimensional thermohydraulics code AQUA for a scram condition from full power operation conditions. Thereafter, thermal stress conditions for the inner barrel were evaluated by the use of a structural analysis code FINAS with the thermohydraulic results calculated by the AQUA code as boundary conditions. From the thermohydraulic analysis and the thermal stress analysis, the following results have been obtained. (1) A large axial temperature gradient was calculated at the region between the upper and lower flow holes located on the inner barrel. The axial position of the thermal stratification interface was fixed in the various circumferential directions. As for the comparison with a 40% operation condition, maximum temperature gradients at the lower flow hole region indicated a 2 times value of that in the 40% operation condition. (2) Transient thermal stratification phenomena were observed after 120 sec from the reactor scram in the numerical results. These tendencies on thermal stratification phenomena were sameness with the transient results from the 40% operation condition. (3) During the reactor trip from full power operation, large temperature gradient in both vertical and sectional direction are enforced around the lower flow hole, since there exists flow pass of low temperature sodium through this hole. As a result, the maximum thermal stress within 32.6 kg/mm 2 was predicted at the lower flow hole when considering stress concentration at the hole edge. (J.P.N.)

  14. Safety evaluation for the prototype Fast Breeder Reactor MONJU as a Japanese TSO

    International Nuclear Information System (INIS)

    Endo, Hiroshi

    2010-01-01

    In the safety field of fast breeder reactors (FBRs), JNES is conducting an evaluation work of the safety regulation by Nuclear and Industry Safety Agency (NISA) for the re-start of a prototype FBR MONJU. MONJU has been stopped over 14 years since 1995 due to a sodium leakage accident at a secondary heat transport system, and is now reached to the criticality on 8th of May, 2010. JNES is supporting the safety regulation work conducted by NISA based on the following activities: i) Support of the technical evaluation of the application for the establishment license prepared by Japan Atomic Energy Agency (JAEA), ii) Support of the description of the safety review report by NISA based on independent safety analyses for the major accident events such as unprotected loss-of-flow (ULOF) by employing the latest findings on the study of core disruptive accidents (CDAs) independently conducted by JNES, iii) Support of the risk-informed-regulation (RIR) such as an accident management (AM) review, iv), and Consideration on the safety regulation policy from the points of severe accidents and source-term behaviors including the cesium (Cs). The objective of this paper is to introduce the major activities of JNES in the safety domain of MONJU regulations. (author)

  15. Leakage limits for inflatable seals of sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.in; Raj, Baldev

    2014-01-15

    Highlights: • All possible types/modes of gas escape covered. • Limits include simultaneous contributions from bypass and permeation leakage modes. • Leakage of radioactive cover gas with fission products assumed. • Possibility of sodium frost deposition in sealed gap considered. • Cover gas activity decay during fuel handling and relative importance of types/modes of leakage considered for realistic results and simpler seal design. -- Abstract: Estimation and stipulation of allowable leakage for inflatable seals of 500 MWe Prototype Fast Breeder Reactor is depicted. Leakage limits are specified using a conservative approach, which assumes escape of radioactive cover gas with fission products across the seals in bypass and permeation modes and possibility of sodium frost deposition in sealed gaps because of permeation leakage of inflation gas. Procedures to arrive at the allowable leakages of argon cover gas (normal-operation/fuel-handling: 10{sup −3}/10{sup −2} scc/s/m length of seal) and argon inflation gas (10{sup −3} scc/s/m length of seal) is described.

  16. Optimal measurement uncertainties for materials accounting in a fast breeder reactor spent-fuel reprocessing plant

    International Nuclear Information System (INIS)

    Dayem, H.A.; Kern, E.A.; Markin, J.T.

    1982-01-01

    Optimization techniques are used to calculate measurement uncertainties for materials accountability instruments in a fast breeder reactor spent-fuel reprocessing plant. Optimal measurement uncertainties are calculated so that performance goals for detecting materials loss are achieved while minimizing the total instrument development cost. Improved materials accounting in the chemical separations process (111 kg Pu/day) to meet 8-kg plutonium abrupt (1 day) and 40-kg plutonium protracted (6 months) loss-detection goals requires: process tank volume and concentration measurements having precisions less than or equal to 1%; accountability and plutonium sample tank volume measurements having precisions less than or equal to 0.3%, short-term correlated errors less than or equal to 0.04%, and long-term correlated errors less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having precisions less than or equal to 0.4%, short-term correlated errors less than or equal to 0.1%, and long-term correlated errors less than or equal to 0.05%

  17. Development of an ISI robot for the fast breeder reactor MONJU primary heat transfer system piping

    International Nuclear Information System (INIS)

    Tagawa, Akihiro; Ueda, Masashi; Yamashita, Takuya; Narisawa, Masataka; Haga, Kouichi

    2011-01-01

    This paper describes the development of a new inspection robot for the In-Service Inspection of the heat transfer system of the Fast Breeder Reactor MONJU. The inspection was carried out using a tire-type ultrasonic sensor for volumetric tests at high temperature (atmosphere, 55degC; piping surface, 80degC) and radiation exposure condition (dose rate, 10 mGy/h; piping surface dose rate, 15 mGy/h). An inspection robot using a new tire type for the ultrasonic testing sensor and a new control method was developed. A signal-to-noise ratio S/N over 2 was obtained during the functional test for a calibration defect with a depth of 50%t (from the tube wall thickness). In the automatic inspection test, an EDM slit with a depth of 9% from the pipe thickness was detectable and with an S/N ratio = 4.0 (12.0 dB). (author)

  18. Development of austenitic stainless steel plate (316MN) for fast breeder reactors

    International Nuclear Information System (INIS)

    Nakazawa, Takanori; Abo, Hideo; Tanino, Mitsuru; Komatsu, Hazime.

    1989-01-01

    High creep-fatigue resistance is required for the structural materials for fast breeder reactors. As creep-fatigue life is closely related to creep-rupture ductility, the effects of C, N and Mo on creep-rupture properties were investigated with a view to improving the creep-fatigue resistance of stainless steel. Strengthening by the addition of C has a great adverse effect on rupture ductility, but N can strengthen the steel without decreasing rupture ductility. Strengthening by Mo decreases rupture ductility but this effect is small. The low-C-medium-N (0.01%C - 0.07%N) stainless steel 316 MN developed based on the findings described above exhibits only a small decrease in creep-rupture strength in long-time periods compared with the conventional 316 steel. This steel offers excellent rupture ductility and the 10,000-hour rupture strength which is about 1.2 times that of conventional steel. Moreover, this steel exhibits excellent properties in creep fatigue test. (author)

  19. Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Padmakumar, G.; Pandey, G.K.; Vaidyanathan, G.

    2009-01-01

    Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained. Studies were conducted simulating both low and full power operating conditions. This paper discusses the model simulation, similarity criteria, the various thermal hydraulic studies that were carried out, the results obtained and the comparison with the prototype measurements.

  20. Promotion of research and development for fast breeder reactors in Japan

    International Nuclear Information System (INIS)

    1989-01-01

    The ''Long-term Program for Development and Utilization of Nuclear Energy'' by Japan Atomic Energy Commission (JAEC) plans to develop fast breeder reactors (FBRs) as the mainstay of Japan's future nuclear power generation. For the development of the FBR, the main principle is to pursue the establishment of FBR plutonium utilization systems, which are superior to the LWR uranium utilization systems. The above program also states as follows: ''Incessant efforts, with proper co-operation of the public and private sectors, will be required for a considerably long time for FBR development, and the concrete measures to be implemented in this connection will be discussed, hereafter, in continuation at the Advisory Committee on FBR Development Program of the Atomic Energy Commission''. It was clearly stated in the ''Long-term Program'' that utilities would play a leading role in the design, construction, and operation of DFBRs to follow ''Monju''. They have entrusted Japan Atomic Power Company (JAPC) with the above tasks. Technical verification and demonstration tests for seismic design and feasibility studies of main components are being conducted with government support. Regarding the nuclear fuel cycle, PNC has been developing FBR fuel reprocessing technologies. Currently, the Recycle Equipment Test Facility (RETF) is being designed to conduct engineering-scale tests. A pilot plant is planned to start operation after 2000 assuming positive results of the tests. (author)

  1. Protected plutonium breeding by transmutation of minor actinides in fast breeder reactor

    International Nuclear Information System (INIS)

    Meiliza, Yoshitalia; Saito, Masaki; Sagara, Hiroshi

    2008-01-01

    The improvement of proliferation resistance properties of Pu and the burnup characteristics of fast breeder reactor (FBR) had been studied by utilizing minor actinides (MAs) to produce more 238 Pu from 237 Np and 241 Am through neutron capture reaction. The higher the 238 Pu content in the fuel, the higher the proliferation resistance of the fuel would be owing to the natural characteristics of 238 Pu with high decay heat and high neutron production. The present paper deals with the assessment of passive measure against nuclear material proliferation by focusing on improving the inherent proliferation barrier of discharged Pu from an FBR. Results showed that 5% MA doping to the blanket of an FBR gives as high as 17-19% 238 Pu, which could be seen as a significant improvement of the proliferation properties of Pu. Moreover, additional 5% ZrH 2 , together with 5% MA doping to the blanket, could enhance the 238 Pu fraction much more (22-24%). With an assumption of protected Pu whose 238 Pu isotopic fraction is more than 12%, the present paper revealed that protected Pu could be produced more than the Pu consumed (protected Pu breeding) through incineration in an FBR with doping of a minimum 3% MAs or (2% MAs+5% ZrH 2 ) to the blanket. (author)

  2. Study on safety of crystallization method applied to dissolver solution in fast breeder reactor reprocessing

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Fujine, Yukio; Asakura, Toshihide; Murazaki, Minoru; Koyama, Tomozo; Sakakibara, Tetsuro; Shibata, Atsuhiro

    1999-03-01

    The crystallization method is proposed to apply for recovery of uranium from dissolution liquid, enabling to reduce handling materials in later stages of reprocessing used fast breeder reactor (FBR) fuels. This report studies possible safety problems accompanied by the proposed method. Crystallization process was first defined in the whole reprocessing process, and the quantity and the kind of treated fuel were specified. Possible problems, such as criticality, shielding, fire/explosion, and confinement, were then investigated; and the events that might induce accidental incidents were discussed. Criticality, above all the incidents, was further studied by considering exampled criticality control of the crystallization process. For crystallization equipment, in particular, evaluation models were set up in normal and accidental operation conditions. Related data were selected out from the nuclear criticality safety handbooks. The theoretical densities of plutonium nitrates, which give basic and important information, were estimated in this report based on the crystal structure data. The criticality limit of crystallization equipment was calculated based on the above information. (author)

  3. Pulsed Nd-YAG laser welding of Prototype Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    Suresh Varma, P.V.; Gupta, Amit; Amit, K.; Bhatt, R.B.; Afzal, Mohd.; Panakkal, J.P.; Kamath, H.S.

    2009-02-01

    End plug welding of Prototype Fast Breeder Reactor (PFBR) fuel elements involves welding of fully Austenitic Stainless Steel (ASS) of grade D9 clad tube with 316M end plug. Pulsed Gas Tungsten Arc Welding (GTAW) is being used for the production of PFBR fuel elements at Advanced Fuel Fabrication Facility (AFFF). GTAW is an established process for end plug welding and hence adopted by many countries. GTAW has got certain limitations like heat input, arc gap sensitivity and certain sporadic defects like tungsten inclusion. Experiments have been carried out at AFFF to use Laser Beam Welding (LBW) technique as LBW offers a number of advantages over the former process. This report mainly deals with the optimization of laser parameters for welding of PFBR fuel elements. To facilitate pulsed Nd-YAG laser spot welding, parameters like peak power, pulse duration, pulse energy, frequency and defocusing of laser beam on to the work piece have been optimized. On the basis of penetration requirement laser welding parameters have been optimized. (author)

  4. Outline of nuclear power plant project of fast breeder reactor (FBR) 'Monju'

    International Nuclear Information System (INIS)

    Suzuki, Fumio; Higashide, Noboru

    1982-01-01

    The first step of the construction of the fast breeder prototype reactor ''Monju'' in Japan was taken by the declaration of agreement of Fukui Prefecture governor in May, 1982, and the succeeding approval by the cabinet council. In this paper, the outline of the project is described. The paper first introduces the development status of FBRs in the world, and next, describes the progress of planning the ''Monju'' project. Initially, ''Monju'' was invited by Shiraki region of Tsuruga City, but the move of village was not agreed. Thus, the center of the site was moved to ''Haseda'' about 1 km north-east. In addition, there were some limitations, and the layout of the facility and the design of public work structures including harbour construction were done under many limiting conditions. However, the site is rather in good conditions, and the design rationalization was attempted as far as possible within a range which does not affect the safety of the nuclear power plant. The outline of the examination related to public works and the siting conditions are reported, dividing them into topographic features, geology, meteorology, sea conditions, and natural environment. As the public works plan, site preparation, roads, harbour installations and water intake and discharge facility are also described. (Wakatsuki, Y.)

  5. Multiple lead seal assembly for a liquid-metal-cooled fast-breeder nuclear reactor

    International Nuclear Information System (INIS)

    Hutter, E.; Pardini, J.A.

    1977-01-01

    A reusable multiple lead seal assembly provides leak-free passage of stainless-steel-clad instrument leads through the cover on the primary tank of a liquid-metal-cooled fast-breeder nuclear reactor. The seal isolates radioactive argon cover gas and sodium vapor within the primary tank from the exterior atmosphere and permits reuse of the assembly and the stainless-steel-clad instrument leads. Leads are placed in flutes in a seal body, and a seal shell is then placed around the seal body. Circumferential channels in the body and inner surface of the shell are contiguous and together form a conduit which intersects each of the flutes, placing them in communication with a port through the wall of the seal shell. Liquid silicone rubber sealant is injected into the flutes through the port and conduit; the sealant fills the space in the flutes not occupied by the leads themselves and dries to a rubbery hardness. A nut, threaded onto a portion of the seal body not covered by the seal shell, jacks the body out of the shell and shears the sealant without damage to the body, shell, or leads. The leads may then be removed from the body. The sheared sealant is cleaned from the body, leads, and shell and the assembly may then be reused with the same or different leads. 3 claims, 6 figures

  6. Liquid-metal fast breeder reactor fuel rod performance and modeling at high burnup

    International Nuclear Information System (INIS)

    Verbeek, P.; Toebbe, H.; Hoppe, N.; Steinmetz, B.

    1978-01-01

    The fuel rod modeling codes IAMBUS and COMETHE were used in the analysis and interpretation of postirradiation examination results of mixed-oxide fuel pins. These codes were developed in the framework of the SNR-300 research and development (R and D) program at Interatom and Belgonucleaire, respectively. SNR-300 is a liquid-metal fast breeder reactor demonstration plant designed and presently constructed in consortial cooperation by Germany, Belgium, and the Netherlands. RAPSODIE I, the two-bundle irradiation experiment, was irradiated in the French test FBR RAPSODIE FORTISSIMO and is one of the key irradiation experiments within the SNR-300 R and D program. The comparison of code predictions with postirradiation examination results concentrates on clad diameter expansions, clad total axial elongations, fuel differential and total axial elongations, fuel restructuring, and fission gas release. Fuel rod modeling was considered in the light of benchmarking of the codes, and there was consideration of fuel rod design for operation at low and high burnup

  7. Power supply for control and instrumentation in Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Raghavan, K.; Shanmugam, T.K.

    1977-01-01

    The design and operation of the four 'no-break' power supplies for control and instrumentation in the Fast Breeder Test Reactor (FBTR), Kalpakkam, are described. Interruptions in the power supplies are eliminated by redundancy and battery back-up source while voltage dips and transients are taken care by automatic regulation system. The four power supplies are : (1) 24 V D.C. exclusively for neutronic and safety circuits, (2) 48 V D.C. for control logic indication lamps and solenoid valves, (3) 220 V D.C. for switchgear control, control room emergency lighting and D.C. flushing oil pump for the turbine and (4) 220 V A.C. single-phase 50 H/Z for computers and electronics of control and instrumentation. Stationary lead-acid batteries (lead antimony type) in floating mode operation with rectifier/charger are used for emergency back-up. All these power supplies are fed by 415 V, 3-phase, 50 HZ emergency supply buses which are provided with diesel generator back-up. Static energy conversion system (in preference to mechanical rotation system) is used for A.C. to D.C. and also for A.C. to A.C. conversion. (M.G.B.)

  8. Tridimensional ultrasonic images analysis for the in service inspection of fast breeder reactors

    International Nuclear Information System (INIS)

    Dancre, M.

    1999-11-01

    Tridimensional image analysis provides a set of methods for the intelligent extraction of information in order to visualize, recognize or inspect objects in volumetric images. In this field of research, we are interested in algorithmic and methodological aspects to extract surface visual information embedded in volume ultrasonic images. The aim is to help a non-acoustician operator, possibly the system itself, to inspect surfaces of vessel and internals in Fast Breeder Reactors (FBR). Those surfaces are immersed in liquid metal, what justifies the ultrasonic technology choice. We expose firstly a state of the art on the visualization of volume ultrasonic images, the methods of noise analysis, the geometrical modelling for surface analysis and finally curves and surfaces matching. These four points are then inserted in a global analysis strategy that relies on an acoustical analysis (echoes recognition), an object analysis (object recognition and reconstruction) and a surface analysis (surface defects detection). Few literature can be found on ultrasonic echoes recognition through image analysis. We suggest an original method that can be generalized to all images with structured and non-structured noise. From a technical point of view, this methodology applied to echoes recognition turns out to be a cooperative approach between morphological mathematics and snakes (active contours). An entropy maximization technique is required for volumetric data binarization. (author)

  9. Performance characterization of geopolymer composites for hot sodium exposed sacrificial layer in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Haneefa, K. Mohammed, E-mail: mhkolakkadan@gmail.com [Department of Civil Engineering, IIT Madras, Chennai (India); Santhanam, Manu [Department of Civil Engineering, IIT Madras, Chennai (India); Parida, F. C. [Radiological Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2013-12-15

    Highlights: • Performance evaluation of geopolymers subjected to hot liquid sodium is performed. • Apart from mechanical properties, micro-analytical techniques are used for material characterization. • The geopolymer composite showed comparatively lesser damage than conventional cement composites. • Geopolymer technology can emerge as a new choice for sacrificial layer in SCFBRs. - Abstract: A sacrificial layer of concrete is used in sodium cooled fast breeder reactors (SCFBRs) to mitigate thermo-chemical effect of accidentally spilled sodium at and above 550 °C on structural concrete. Performance of this layer is governed by thermo-chemical stability of the ingredients of sacrificial layer concrete. Concrete with limestone aggregate is generally used as a sacrificial layer. Conventional cement based systems exhibit instability in hot liquid sodium environment. Geo-polymer composites are well known to perform excellently at elevated temperatures compared to conventional cement systems. This paper discusses performance of such composites subjected to exposure of hot liquid sodium in air. The investigation includes comprehensive evaluation of various geo-polymer composites before any exposure, after heating to 550 °C in air, and after immersing in hot liquid sodium initially heated to 550 °C in air. Results from the current study indicate that hot liquid sodium produces less damage to geopolymer composites than to the existing conventional cement based system. Hence, the geopolymer technology has potential application in mitigating the degrading effects of sodium fires and can emerge as a new choice for sodium exposed sacrificial layer in SCFBRs.

  10. Development of three-dimensional nuclear design program for large fast breeder reactor

    International Nuclear Information System (INIS)

    Inoue, Kohtaro

    1987-01-01

    The report describes a calculation program for core design, called HICOM, and its calculation accuracy. HICOM is designed for three-dimensional neutron diffusion calculation and combustion calculation for large fast breeder reactors to be conducted according to a control rod plan and fuel replacement plan. The improved coarse mesh technique is applied to neutron diffusion calculation. It is demostrated that HICOM permits rapid and accurate operation. For the evaluation of the applicability of HICOM, three-dimensional six-group neutron diffusion calculation is conducted for a 1,000 MWe axial heterogeneous FBR core. Results demonstrate that the program can perform numerical calculation in a time period shorter than 1-40 that for calculation by CITATION (triangle mesh method). This is achieved by using the improved coarse mesh method and carrying out the operation by a vectorial procedure. For the evaluation of the nuclear calculation accuracy of HICOM, analysis is made of reactivity, output distribution and B 4 C control rod worth emasured in an FCA criticality experiment carried out by the Japan Atomic Energy Research Institute. Calculations are found to agree with measurements within a permissible error. The same level of calculation accuracy is obtained for homogneous core, axial heterogeneous core and cores with internal blankets with different forms. (Nogami, K.)

  11. Status of fast breeder reactors and associated fuel cycle in India

    International Nuclear Information System (INIS)

    Chellapandi, P.

    2009-01-01

    Full text: India is the largest democracy with the current population of about 1.05 billion and is on a road to rapid growth in economy. An impressive average domestic product (GDP) growth rate of about 8 % per year has been achieved in 2006-07 and it is targeted to touch 10 % per year for the next 10 years. Towards realizing this targeted growth, development activities are planned based on well-conceived road map and clear vision. Like elsewhere, the energy and electricity growth in India are also closely linked to growth in economy. Indices of socio-economic development like literacy, longevity, GDP and human development are directly dependent upon the per capita energy consumption of a country. India is aiming to reach at least per capita energy consumption equal to the present world average (2200 kWh/a) by 2030 from the current value of (660 kWh/a). The current installed capacity of ∼138 GW(e) needs to be increased to about 600 GWe by 2030 assuming the population of about 1.4 billion. Energy strategists in the country have realized the importance of judicious mix of energy resources to meet this challenge. A large share of nuclear energy is an inevitable choice in this judicious energy mix from resources, sustainability and environment considerations. The nuclear is expected to contribute about 63 GWe by 2030, which will be steadily increased to 275 GWe by 2052, against the total projected capacity of 1344 GWe. The three stage visionary programme of India envisages Water Reactors (first stage), Fast Breeders with high breeding (second stage) and Thorium based Reactors as third stage. Closed fuel cycle in all stages is an essential ingredient. The success of each stage depends upon expeditious maturity of the earlier stage as India has limited indigenous resources of uranium, but vast resources of thorium. India ranks high in nuclear technology scale with strong R and D, high quality human resources, sound infrastructure, unwavering Government support and

  12. Web-enabled work permit system for fast breeder test reactor

    International Nuclear Information System (INIS)

    Madurai Meenachi, N.; Vinolia, K.; Ramanathan, V.

    2003-01-01

    The objective of this project is to computerize and web-enable the Work Permit System for the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam. The existing Work Permit System at FBTR was studied in detail. Since all the formalities were paper-based, the risk of human error in scrutinizing all permits before reactor start-up was high. Compilation of reports (daily, monthly, yearly etc.) was tedious. The work permit system was therefore automated in order to enable the operation group manage the maintenance work carried out in the plant systematically with entries. The entire project was classified into five permit modules -maintenance, transfer, return, cancellation and reissue. Each module takes care of the entry and maintenance of data in their respective fields in their respective tables. The user is also provided with an option to take a hard copy of the report of his/her choice. A client/server based system was designed to web-enable the entire project. The server program was designed using VB 6.0 as the front-end and MS Access database as the back end to store the data. The client software was developed using Active Server Pages and published using personal web server in the Intranet. A number of administrative tools have been incorporated in the software to ensure access security and integrity of the database. An online help feature with search facilities was added to the software. The work permit system software is now already being used at FBTR and has been deemed to be an invaluable aid in empowering the availability of the reactor and determining the performance history of the equipment. (author)

  13. Evaluation of cobalt and nickel base materials for sliding and static contact applications in a liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Hoffman, N.J.; Droher, J.J.; Chang, J.Y.; Galioto, T.A.; Miller, R.L.; Schrock, S.L.; Whitlow, G.A.; Wilson, W.L.; Johnson, R.N.

    1976-01-01

    The paper covers pertinent metallurgical and tribological aspects of three alloys that are being considered for surfaces that must rub while immersed in liquid sodium coolant within a fast breeder reactor system. The alloys are cobalt-base hardfacing alloy type 6, Tribaloy 700, and Inconel 718. Topics discussed include chemistry and microstructure, hardness, and behavior in high-temperature sodium with respect to dynamic friction, diffusion bonding, and corrosion

  14. A perspective on research and development in austenitic stainless steels for fast breeder reactor technology at Kalpakkam

    International Nuclear Information System (INIS)

    Baldev Raj; Jayakumar, T.; Shankar, P.

    2010-01-01

    A fast breeder reactor with closed fuel cycle is an inevitable technology option to provide energy security for India. Innovations in materials technology have enabled the realization of unique and advanced features in the Indian fast breeder reactors and their associated fuel cycles. Materials development and materials technologies, particularly the widely used austenitic stainless steels discussed in this paper, have a deterministic influence on the advancement, safety, reliability, cost effectiveness and thus success of the fast breeder programme. Rigorous research and development for alloy development complemented with detailed structure-property evaluation of relevant mechanical and corrosion behaviour data have been possible with the state of art facilities housed at IGCAR. These data provide useful inputs for design engineers to ensure reliable and safe operation of the components. Advanced concepts in alloy design and grain boundary engineering are utilized to enhance the corrosion resistance and mechanical properties of various structural materials. Advanced NDE techniques for the assessment of manufactured components and in-service inspection have been developed, enhancing the confidence in the performance of the plant components and systems. The technology demonstration of critical stainless steel components using advanced forming and welding technologies with support from modelling for optimization of the fabrication processes enhanced the confidence in the development of the complex fast breeder reactor and associated fuel cycle technologies, with active support from national academic and research institutes and industry. This chapter presents a comprehensive overview on the advances in stainless steel technology as well as the challenges ahead for aspiring young minds in the field of fast reactor technology. (author)

  15. Storage and management of fuel from fast breeder test reactor and KAlpakkam MINI

    International Nuclear Information System (INIS)

    Sodhi, B.S.; Rao, M.S.; Natarajan, R.

    1999-01-01

    Two Research Reactors, FBTR (Fast Breeder Test Reactor) and KAMINI (KAlpakkam MINI) are in operation at Kalpakkam, India. FBTR is a 40 MWt reactor. It is the first reactor to use mixed carbide (70% PuC-30% UC) as driver fuel. Special precautions are needed to fabricate pellets in glove boxes under inert atmosphere to take into account the possibility of criticality, radiation, pyrophoricity and toxicity of PuC. FBTR has been operating with small core up to 12 MWt power. The initial limit was 250 W/cm, linear heat rating and 25,000 MWd/t peak burnup. This limit was increased to 320 W/cm and 50,000 MWd/t respectively after rigorous analysis. At present the core has reached 40,000 MWd/t without any pin failure. After 25,000 MWd/t burnup one fuel subassembly (SA) was removed and PEE was carried out. The results were as expected by the analysis. In FBTR, fuel is stored in a container filled with argon and the container is cooled by forced circulation of air (during storage). Closing the fuel cycle is important for the breeder programme. Therefore, efforts have been made to set up a reprocessing plant. It uses the well proven purex process. The irradiated fuel is sheared in a single pin chopper and dissolved in an electrochemical dissolver. The resulting solution after adjusting the valency of Pu to IVth state is processed in the solvent extraction plant using 30% Tri-n-Butyl phosphate/n-dodecane as solvent. KAMINI is 30 kWt neutron source reactor which uses light water as moderator and coolant and has as a fuel U-233 aluminium alloy. Uranium-233 has been indigenously recovered from thorium irradiated in CIRUS reactor at Trombay. KAMINI was made critical on October 1996. It is housed in a vault below one of the hot cells in the Radiometallurgy laboratories of IGCAR. This reactor is planned to be used for neutron radiography of fuel elements and neutron activation analysis. It is available for use by research institutions and universities also. This paper describes the

  16. Hydraulic experiments on the failed fuel location module of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Rajesh, K.; Kumar, S.; Padmakumar, G.; Prakash, V.; Vijayashree, R.; Rajan Babu, V.; Govinda Rajan, S.; Vaidyanathan, G.; Prabhaker, R.

    2003-01-01

    The design of Prototype Fast Breeder Reactor (PFBR) is based on sound design concepts with emphasis on intrinsic safety. The uncertainties involved in the design of various components, which are difficult to assess theoretically, are experimentally verified before design is validated. In PFBR core, the coolant (liquid sodium) enters the bottom of the fuel subassembly, passes over the fuel pins picking up the fission heat and issues in to a hot pool. If there is any breach in the fuel pins, the fission products come in direct contact with the coolant. This is undesirable and it is necessary to locate the subassembly with the failed fuel pin and to isolate it. A component called Failed Fuel Location Module (FFLM) is employed for locating the failed SA by monitoring the coolant samples coming out of each Subassembly. The coolant sample from each Subassembly is drawn by FFLM using an EM pump through sampling tube and selector valve and is monitored for the presence of delayed neutrons which is an indication of failure of the Subassembly. The pressure drop across the selector valve determines the rating of the EM Pump. The dilution of the coolant sample across the selector valve determines the effectiveness of monitoring for contamination. It is not possible to predict pressure drop across the selector valve and dilution of the coolant sample theoretically. These two parameters are determined using a hydraulic experiment on the FFLM. The experiment was carried out in conditions that simulate the reactor conditions following appropriate similarity laws. The paper discusses the details of the model, techniques of experiments and the results from the studies

  17. Study of the accidental risk of the German fast breeder prototype reactor SNR-300

    International Nuclear Information System (INIS)

    Koeberlein, K.

    1983-01-01

    A fact-finding committee of the German Federal Parliament in July 1980 recommended to perform a 'risk-oriented study' of the SNR-300, the German 300 MWe fast breeder prototype reactor being under construction at Kalkar. The main aim of this study was to allow a comparative safety evaluation between the SNR-300 and a modern PWR, thus to prepare a basis for a political decision on the SNR-300. Methods and main results of the study are presented in this paper. In the first step of the study six groups of accidents have been identified which may initiate core destruction. By reliability analyses, expected frequency of each group has been estimated. Conditional probabilities for conceivable reactor tank failure modes have been analysed. Tank failure after core destruction leads to release of energy and radioactive material into the containment system. Such accident sequences have been pursued further. Based on a number of core destruction initiators and tank failure modes and various combinations of success and failure states of the containment systems, detailed calculations of different containment scenarious were carried out. From the results of the plant systems analysis, five release categories have been defined. Possible effects of external events and releases of radioactivity from the spent fuel storage pool have also been analysed. In order to quantify the degree of uncertainty of the calculated frequencies, subjective probability distributions of fixed, but inaccurately known quantities have been propagated through the calculations. Using the release categories as input, accident consequences were calculated for the site Kalkar. Though the uncertainty bandwidths for the accident frequencies estimated in the SNR-300 analysis are much wider than for the PWR, the analysis indicates that frequencies of severe accidents, and consequences, are smaller for the SNR-300 than for the PWR as analysed in the 'German Risk Study'. (orig.)

  18. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-10-15

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  19. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Manimaran, M.; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-01-01

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored

  20. Studies on natural circulation cooling enhancement in a spent fuel in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Isamu; Akamatsu, Mikio; Toda, Shinichi; Sato, Manabu [Kawasaki Heavy Industries Ltd., Kobe (Japan); Mayumi, Masami

    2001-01-01

    Fast breeder reactor (FBR) has some advantages such as effective application of plutonium, excellent capacity to fire minor-actinides (longer half-life nuclides such as Np, Am, Cm, and so on) contained in radioactive wastes in the reactor to convert their shorter half-life nuclides. However, fuels containing the minor-actinides have a characteristic with higher exotherm and radioactive intensity than those of conventional ones, it is essential at their actual stages to prepare some rational fuel handling systems on their transportation, storage and so forth. In addition, there are few examples on natural circulation heat transfer test of a liquid metal using long sized container. Then, in order to establish an evaluating method on decay-heat removing property of a spent fuel assembly in sodium canister and pot, some natural circulation tests on a long sized container including a quasi pin-bundle structure for a working fluid of lead-bismuth (Pb-Bi) mixture with easier handling than that of sodium was carried out. A specimen could be mounted at optional angles from horizontal to vertical positions so as to evaluate effects of inclined angles. In addition, in order to estimate temperature and flow rate distribution in a long sized container and understand thermal flowing phenomenon in specimen system, numerical analysis using multi-dimensional analysis code was carried out. As a result, it was found that in vertical arrangement system, natural circulation phenomenon is limited at upper portion of the exothermal portion, and its maximum temperature was tested at central portion of top pin-bundle of the exothermal portion. And, it was also found that at horizontal arrangement maximum temperature was 40 centigrade less than that of vertical arrangement, and so forth. (G.K.)

  1. Deterioration of limestone aggregate mortars by liquid sodium in fast breeder reactor environment

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed Haneefa, K., E-mail: mhkolakkadan@gmail.com [Department of Civil Engineering, IIT Madras, Chennai (India); Santhanam, Manu [Department of Civil Engineering, IIT Madras, Chennai (India); Parida, F.C. [Radiological Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2014-08-15

    Highlights: • Limestone mortars were exposed to liquid sodium exposure at 550 °C. • Micro-analytical techniques were used to characterize the exposed specimens. • The performance of limestone mortar was greatly influenced by w/c. • The fundamental degradation mechanisms of limestone mortars were identified. - Abstract: Hot liquid sodium at 550 °C can interact with concrete in the scenario of an accidental spillage of sodium in liquid metal cooled fast breeder reactors. To protect the structural concrete from thermo-chemical degradation, a sacrificial layer of limestone aggregate concrete is provided over it. This study investigates the fundamental mechanisms of thermo-chemical interaction between the hot liquid sodium and limestone mortars at 550 °C for a duration of 30 min in open air. The investigation involves four different types of cement with variation of water-to-cement ratios (w/c) from 0.4 to 0.6. Comprehensive analysis of experimental results reveals that the degree of damage experienced by limestone mortars displayed an upward trend with increase in w/c ratios for a given type of cement. Performance of fly ash based Portland pozzolana cement was superior to other types of cements for a w/c of 0.55. The fundamental degradation mechanisms of limestone mortars during hot liquid sodium interactions include alterations in cement paste phase, formation of sodium compounds from the interaction between solid phases of cement paste and aggregate, modifications of interfacial transition zone (ITZ), decomposition of CaCO{sub 3}, widening and etching of rhombohedral cleavages, and subsequent breaking through the weakest rhombohedral cleavage planes of calcite, staining, ferric oxidation in grain boundaries and disintegration of impurity minerals in limestone.

  2. Atmospheric dispersion of sodium aerosol due to a sodium leak in a fast breeder reactor complex

    International Nuclear Information System (INIS)

    Punitha, G.; Sudha, A. Jasmin; Kasinathan, N.; Rajan, M.

    2008-01-01

    Liquid sodium at high temperatures (470 K to 825 K) is used as the primary and secondary coolant in Liquid Metal cooled Fast Breeder Reactors (LMFBR). In the event of a postulated sodium leak in the Steam Generator Building (SGB) of a LMFBR, sodium readily combusts in the ambient air, especially at temperatures above 523 K. Intense sodium fire results and sodium oxide fumes are released as sodium aerosols. Sodium oxides are readily converted to sodium hydroxide in air due to the presence of moisture in it. Hence, sodium aerosols are invariably in the form of particulate sodium hydroxide. These aerosols damage not only the equipment and instruments due to their corrosive nature but also pose health hazard to humans. Hence, it is essential to estimate the concentration of sodium aerosols within the plant boundary for a sodium leak event. The Gaussian Plume Dispersion Model can obtain the atmospheric dispersion of sodium aerosols in an open terrain. However, this model dose not give accurate results for dispersion in spaces close to the point of release and with buildings in between. The velocity field due to the wind is altered to a large extent by the intervening buildings and structures. Therefore, a detailed 3-D estimation of the velocity field and concentration has to be obtained through rigorous computational fluid dynamics (CFD) approach. PHOENICS code has been employed to determine concentration of sodium aerosols at various distances from the point of release. The dispersion studies have been carried out for the release of sodium aerosols at different elevations from the ground and for different wind directions. (author)

  3. Deterioration of limestone aggregate mortars by liquid sodium in fast breeder reactor environment

    International Nuclear Information System (INIS)

    Mohammed Haneefa, K.; Santhanam, Manu; Parida, F.C.

    2014-01-01

    Highlights: • Limestone mortars were exposed to liquid sodium exposure at 550 °C. • Micro-analytical techniques were used to characterize the exposed specimens. • The performance of limestone mortar was greatly influenced by w/c. • The fundamental degradation mechanisms of limestone mortars were identified. - Abstract: Hot liquid sodium at 550 °C can interact with concrete in the scenario of an accidental spillage of sodium in liquid metal cooled fast breeder reactors. To protect the structural concrete from thermo-chemical degradation, a sacrificial layer of limestone aggregate concrete is provided over it. This study investigates the fundamental mechanisms of thermo-chemical interaction between the hot liquid sodium and limestone mortars at 550 °C for a duration of 30 min in open air. The investigation involves four different types of cement with variation of water-to-cement ratios (w/c) from 0.4 to 0.6. Comprehensive analysis of experimental results reveals that the degree of damage experienced by limestone mortars displayed an upward trend with increase in w/c ratios for a given type of cement. Performance of fly ash based Portland pozzolana cement was superior to other types of cements for a w/c of 0.55. The fundamental degradation mechanisms of limestone mortars during hot liquid sodium interactions include alterations in cement paste phase, formation of sodium compounds from the interaction between solid phases of cement paste and aggregate, modifications of interfacial transition zone (ITZ), decomposition of CaCO 3 , widening and etching of rhombohedral cleavages, and subsequent breaking through the weakest rhombohedral cleavage planes of calcite, staining, ferric oxidation in grain boundaries and disintegration of impurity minerals in limestone

  4. Uncertainty evaluation of reliability of shutdown system of a medium size fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zeliang, Chireuding; Singh, Om Pal, E-mail: singhop@iitk.ac.in; Munshi, Prabhat

    2016-11-15

    Highlights: • Uncertainty analysis of reliability of Shutdown System is carried out. • Monte Carlo method of sampling is used. • The effect of various reliability improvement measures of SDS are accounted. - Abstract: In this paper, results are presented on the uncertainty evaluation of the reliability of Shutdown System (SDS) of a Medium Size Fast Breeder Reactor (MSFBR). The reliability analysis results are of Kumar et al. (2005). The failure rate of the components of SDS are taken from International literature and it is assumed that these follow log-normal distribution. Fault tree method is employed to propagate the uncertainty in failure rate from components level to shutdown system level. The beta factor model is used to account different extent of diversity. The Monte Carlo sampling technique is used for the analysis. The results of uncertainty analysis are presented in terms of the probability density function, cumulative distribution function, mean, variance, percentile values, confidence intervals, etc. It is observed that the spread in the probability distribution of SDS failure rate is less than SDS components failure rate and ninety percent values of the failure rate of SDS falls below the target value. As generic values of failure rates are used, sensitivity analysis is performed with respect to failure rate of control and safety rods and beta factor. It is discovered that a large increase in failure rate of SDS rods is not carried to SDS system failure proportionately. The failure rate of SDS is very sensitive to the beta factor of common cause failure between the two systems of SDS. The results of the study provide insight in the propagation of uncertainty in the failure rate of SDS components to failure rate of shutdown system.

  5. Numerical prediction of fire extinguishment characteristics of sodium leak collection tray in a fast breeder reactor

    International Nuclear Information System (INIS)

    Diwakar, S.V.; Mangarjuna Rao, P.; Kasinathan, N.; Das, Sarit K.; Sundararajan, T.

    2011-01-01

    Highlights: ► Sodium fire extinguishment in a leak collection tray is modeled by lumped approach. ► Hydrodynamics of liquid sodium on tray is emulated through a draining/sloshing model. ► Pool burning rates in the tray and holdup vessel are numerically estimated. ► The model directly yields the mass of sodium recovered after extinction of fire. ► Model predictions are in reasonable agreement with the available experimental data. - Abstract: Sodium leak collection tray (LCT) is an efficient passive device used for the extinguishment of liquid sodium fire in case of an accidental leakage from the secondary circuit of a fast breeder reactor. The LCT essentially isolates the leaking sodium into closed containers where the resulting fire is extinguished due to limited availability of oxygen. The current work aims to highlight the combustion extinguishment characteristics of LCT through a lumped formulation by conserving the mass and energy of liquid sodium and constituent gases in various parts of the LCT. Here, the complex hydrodynamics of liquid sodium is emulated through a semi-analytical draining/sloshing model and its burning rates are predicted through a three-dimensional open pool combustion model for the tray region and a closed pool combustion model for the holdup vessel. These simulations evaluate the burning rates at discrete levels of liquid sodium which are subsequently interpolated to establish correlations involving instantaneous liquid levels and oxygen concentration. Using the correlations obtained from the draining and combustion models, the overall lumped formulation directly predicts the un-burnt sodium recoverable after the extinguishment of fire in the LCT. The predicted results of this model compare well with the available experimental data.

  6. Theoretical and experimental investigation of the nonlinear structural dynamics of Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    Liebe, R.

    1978-04-01

    This study describes theoretical and experimental investigations of the dynamic deformation behavior of single and clustered fuel elements under local fault conditions in a Fast Breeder Reactor core. In particular an energetic molten-fuel-coolant-interaction (FCI) is assumed in one subassembly with corresponding pressure pulses, which may rupture the wrapper and load the adjacent fuel elements impulsively. Associated coherent structural deformation may exceed tolerable and damage the control rods. To attack the outlined coupled fluid-structure-interaction problem it is assumed, that the loading at the structures is known in space and time, and that there is no feedback from the deformation response. Then current FCI-knowledge and experience from underwater core model explosion tests is utilized to estimate upper limits of relevant pulse characteristics. As a first step the static carrying capacity of the rigid-plastic hexagonal wrapper tube is calculated using the methods of limit analysis. Then for a general dynamic simulation of the complete elastoplastic subassembly response the concept of a discrete nonlinear hinge is introduced. A corresponding physical lumped parameter hinge model is presented, and general equations of motion are derived using D'Alembert's principle. Application to the static and dynamic analysis of a single complete fuel element includes the semiempirical modelling of the fuel-pin bundle by a homogeneous compressible medium. Most important conclusions are concerning the capability of the theoretical models, the failure modes and threshold load levels of single as well as clustered SNR-300 fuel elements and the safety relevant finding, that only limited deformations are found in the first row around the incident element. This shows in agreement with explosion test results that the structured and closely spaced fuel elements constitute an effective, inherent barrier against extreme dynamic loadings. (orig.) [de

  7. Fault tolerant distributed real time computer systems for I and C of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2014-03-15

    Highlights: • Architecture of distributed real time computer system (DRTCS) used in I and C of PFBR is explained. • Fault tolerant (hot standby) architecture, fault detection and switch over are detailed. • Scaled down model was used to study functional and performance requirements of DRTCS. • Quality of service parameters for scaled down model was critically studied. - Abstract: Prototype fast breeder reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Three-tier architecture is adopted for instrumentation and control (I and C) of PFBR wherein bottom tier consists of real time computer (RTC) systems, middle tier consists of process computers and top tier constitutes of display stations. These RTC systems are geographically distributed and networked together with process computers and display stations. Hot standby architecture comprising of dual redundant RTC systems with switch over logic system is deployed in order to achieve fault tolerance. Fault tolerant dual redundant network connectivity is provided in each RTC system and TCP/IP protocol is selected for network communication. In order to assess the performance of distributed RTC systems, scaled down model was developed with 9 representative systems and nearly 15% of I and C signals of PFBR were connected and monitored. Functional and performance testing were carried out for each RTC system and the fault tolerant characteristics were studied by creating various faults into the system and observed the performance. Various quality of service parameters like connection establishment delay, priority parameter, transit delay, throughput, residual error ratio, etc., are critically studied for the network.

  8. Development of variable width ribbon heating elements for liquid metal and gas-cooled fast breeder reactor fuel rod simulators

    International Nuclear Information System (INIS)

    McCulloch, R.W.; Lovell, R.T.; Post, D.W.; Snyder, S.D.

    1980-01-01

    Variable width ribbon heating elements have been fabricated which provide a chopped cosine, variable heat flux profile for fuel rod simulators used in test loops by the Breeder Reactor Program Thermal Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor Core Flow Test Loop. Thermal, mechanical, and electrical design considerations result in the derivation of an analytical expression for the ribbon contours. From this, the ribbons are machined and wound on numerically controlled equipment. Postprocessing and inspection results in a wound, variable width ribbon with the precise dimensional, electrical, and mechanical properties needed for use in fuel pin simulators

  9. Some study on radiation resistance and reliability of piston ring of waste gas compressor for fast breeder experimental reactor

    International Nuclear Information System (INIS)

    Muramatsu, Takio; Hidaka, Tsukasa

    1976-01-01

    In the fast breeder experimental reactor ''Joyo'', the gaseous wastes such as reactor cover argon, reactor seal nitrogen gas, fuel handling waste gas etc. shall be collected, compressed and storaged for decaying their activity. Compressors applied in the above process have new type oilless piston rings of Teflon filled with graphite, which might be affected by radioactivity of the waste gases. This report deals with some study on the gamma iradiation effects on the plastic piston rings such as tensile strength, elongation, shock and hardness effects under several irradiation doses and on durability test of the irradiated piston rings under the same compression ratio. (auth.)

  10. 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

  11. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands - January 1986

    International Nuclear Information System (INIS)

    Hueper, R.

    1986-01-01

    In 1967 and 1968, the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (DeBeNe) agreed to develop breeder reactors in a joint program. The French Commissariat a l'Energie Atomique entered into contracts in 1977 about dose cooperation in the fast breeder field. The results of activities carried out by DeBeNe partners in 1985 have been compiled in this report. The report begins with a survey of the fast reactor plants followed by the R and D summary. In an additional chapter, a survey is given of international cooperation in 1985

  12. Three-dimensional calculation of the flow in the cold plenum of the Fast Breeder Reactor 1500 MW

    International Nuclear Information System (INIS)

    Chabard, J.P.; Daubert, O.; Gregoire, J.P.; Hemmerich, P.

    1987-01-01

    To solve thermalhydraulics problems which are rising for example on the various parts of nuclear reactors, several departments of the Direction des Etudes et Recherches are developing the N3S code, three-dimensional code using the finite element method. First, this paper presents the basic equations (Navies-Stokes with turbulence modelling and coupled with the thermal equation) and well suited algorithms to solve them. The industrial adequacy of the code is clearly demonstrated through the application to the computation of the flow in the cold plenum of the Fast Breeder Reactor 1500 MW on a mesh of about 20000 velocity nodes [fr

  13. Social and ethical aspects of the liquid-metal fast breeder reactor

    International Nuclear Information System (INIS)

    Lovins, A.B.

    1975-01-01

    Development of liquid fast breeder reactors not only indirectly entails (through commitments of time and resources that foreclose other options), but also directly entails large-scale centralized electrification. The massive economic commitments of such a policy, wether or not it is a nuclear policy, demand and cause major social changes, bypass traditional market mechanisms, concentrate political and economic power, persistently distort political structures and social priorities, compromise professional ethics, are probably inimical to greater distributional equity within and among nations, enhance vulnerability and the paramilitarization of civilian life, introduce major economic and social risks, and reinforce current trends toward centrifugal politics. Deployment of fission technology produces further social and ethical problems, since attempts to reduce potential hazards from operating accidents, from escape of nuclear wastes, or from nuclear violence and coercion will have socio-political side-effects even if they succeed, not to mention the side-effects if they fail. These side-effects, many of which would be worse with fast than with thermal reactors, include repressiveness, abrogation of civil liberties, social rigidity and homogeneity, elitist technocracy, dirigiste autarchy, and suppression of ethical objections. The inability of modern political institutions to cope with the persistent hazards of toxic and explosive nuclear materials strains the competence and perceived legitimacy of those institutions as they try to compromise between individual liberties and public safety and to subject to democratic decision technically tinged policy questions that turn largely on unknown or unknowable information. There is no scientific basis for calculating the likelihood on the maximum long-term of nuclear mishaps, nor for guaranteeing that the effects will not exceed a particular level; it is only known that all precautions are, for fundamental reasons

  14. Choice of rotatable plug seals for prototype fast breeder reactor: Review of historical perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.in; Raj, Baldev, E-mail: baldev.dr@gmail.com

    2015-09-15

    Highlights: • Choice and arrangement of elastomeric inflatable and backup seals as primary and secondary barriers. • With survey (mid-1930s onwards) of reactor, sealing, R&D and rubber technology. • Load, reliability, safety, life and economy of seals and reactors are key factors. • PFBR blends concepts and experience of MOX fuelled FBRs with original solutions. • R&D indicates inflatable seal advanced fluoroelastomer pivotal in unifying nuclear sealing. - Abstract: Choice and arrangement of elastomeric primary inflatable and secondary backup seals for the rotatable plugs (RPs) of 500 MW (e), sodium cooled, pool type, 2-loop, mixed oxide (MOX) fuelled Prototype Fast Breeder Reactor (PFBR) is depicted with review of various historical perspectives. Static and dynamic operation, largest diameters (PFBR: ∼6.4 m, ∼4.2 m), widest gaps and variations (5 ± 2 mm) and demanding operating requirements make RP openings on top shield (TS) the most difficult to seal which necessitated extensive development from 1950s to early 1990s. Liquid metal freeze seals with life equivalent to reactor prevailed as primary barrier (France, Japan, U.S.S.R.) during pre-1980s in spite of bulk, cost and complexity due to the abilities to meet zero leakage and resist core disruptive accident (CDA). Redefinition of CDA as beyond design basis accident, tolerable leakage and enhanced economisation drive during post-1980s established elastomeric inflatable seal as primary barrier excepting in U.S.S.R. (MOX fuel, freeze seal) and U.S.A. (metallic fuel). Choice of inflatable seal for PFBR RPs considers these perspectives, inherent advantages of elastomers and those of inflatable seals which maximise seal life. Choice of elastomeric backup seal as secondary barrier was governed by reliability and minimisation as well as distribution of load (temperature, radiation, mist) to maximise seal life. The compact sealing combination brings the hanging RPs at about the same elevation to reduce

  15. Optimization of intermediate heat exchangers for sodium cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Barratt, R.O.; Cox, J.; Beith, R.M.V.

    1978-01-01

    Design evolution of the Intermediate Heat Exchanger for the U.S. Fast Breeder Programme is traced from the initial FFTF through to the Clinch River Designs and anticipated onward Commercial Plant Concepts. Supporting development work is outlined. Technical merits of certain features, such as sine wave tube banks and replaceable tube banks (in-situ), are weighed up against quantified cost advantages with simpler arrangements. Future design trends are outlined which will lead eventually to fully optimised designs. (author)

  16. High-definition radiography of tube-to-tubesheet welds of steam generator of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Venkatraman, B.; Sethi, V.K.; Jayakumar, T.; Raj, B.

    1995-01-01

    In the steam generator of the Prototype Fast Breeder Reactor (PFBR), steam is generated by the transfer of heat from secondary sodium to water. Due to the inherent dangers of sodium-water reaction, the integrity of weld joints separating sodium and water/steam is of paramount importance. This is particularly true and very important for the tube-to-tubesheet joints. This paper discusses the use of projective magnification technique by microfocal radiography for the quality evaluation and optimisation of the welding parameters of such small tube-to-tubesheet welds of the steam generator of PFBR. (author)

  17. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Malathi, N.; Sahoo, P., E-mail: sahoop@igcar.gov.in; Ananthanarayanan, R.; Murali, N. [Real Time Systems Division, Electronics, Instrumentation and Radiological Safety Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2015-02-15

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.

  18. Assessment of the thorium and uranium fuel cycle in the fast breeder and the high temperature reactor

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    1977-01-01

    This report assesses the fissile fuel economy of the uranium and thorium cycle in the advanced reactors currently under development, the fast breeder reactor (FBR) and the high temperature reactor (HTR). It is shown by means of detailed burnup calculations that replacing UO 2 with ThO 2 or Th-metal as the radial blanket breeding material will not have any significant imapct on the breeding and burnup properties of the FBR. A global, analytical investigation is performed to study the fissile fuel economy of the many fissile fuel cycles possible in the HTR. Here it is demonstrated that the optimum conversion ratio of CR 3 O 8 ) demands are evaluated for a country such as the FRG under the assumptions of different future reactor strategy scenarios. Here it is demonstrated that the employement of both HTRs and FBRs can lead to a practically resource independent energy supply system within the next 40 to 60 years. However only through the large scale employement of the fast breeder can the future nuclear resource requirements be assured. (orig.) [de

  19. 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

  20. Summary of estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities

    International Nuclear Information System (INIS)

    Miller, C.W.; Meyer, H.R.

    1985-01-01

    A project is underway at Oak Ridge National Laboratory to assess the human health and environment effects associated with operation of Liquid Metal Fast Breeder Reactor fuel cycle. In this first phase of the work, emphasis was focused on routine radionuclide releases from reactor and reprocessing facilities. For this study, sites for fifty 1-GW(e) capacity reactors and three reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also, the results of these analyses are being used in the identification of research priorities. 13 refs., 2 figs., 3 tabs

  1. Improvement in the long term creep rupture strength of SUS 316 steel for fast breeder reactors by nitrogen addition

    International Nuclear Information System (INIS)

    Nakazawa, Takanori; Abo, Hideo; Tanino, Mitsuru; Komatsu, Hazime; Tashimo, Masanori; Nishida, Takashi.

    1989-01-01

    Improvement of creep fatigue property of structural materials for fast breeder reactors. In order to improve the resistance to creep fatigue of SUS 316 steels, the effects of nitrogen, carbon, and molybdenum on creep properties have been investigated, under the concept that creep fatigue endurance is correspond to creep rupture ductility. Creep rupture tests and slow strain rate tensile tests were conducted at 550degC and extensive microstructural works were performed. The strengthening by nitrogen is much greater than carbon. Moreover, while carbon reduces rupture ductility, nitrogen does not change it. The addition of carbon results in coarse carbide formation on grain boundaries during creep, but with nitrogen very fine Fe 2 Mo particles precipitate on grain boundaries. The difference between the effects of nitrogen and carbon on creep properties is arise from the different morphology of precipitation. Strengthening by molybdenum brings about a slight decrease in rupture ductility. On the basis of these results, 0.01%C-0.07%N-11%Ni-16.5%Cr-2%Mo steel is selected as a promising material for fast breeder reactors. This steel has higher rupture ductility and strength than SUS 316 steel. It is also confirmed that this steel has a higher resistance to creep fatigue. (author)

  2. Status of the fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    Hueper, R.

    1982-01-01

    In 1967 and 1968 the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (''DeBeNe'') agreed to develop, in a joint program, breeder reactors to the point of commercial maturity. The following research organizations take part in this effort: Kernforschungszentrum Karlsruhe (KfK); INTERATOM, Bergisch Gladbach; ALKEM, Wolfgang near Hanau; SCK/CEN, Mol; Belgonucleaire, Brussels; ECN, Petten; TNO, Apeldoorn; NERATOOM, The Hague. The three German institutions mentioned above have been interrelated since 1977 by the Entwicklungsgemeinschaft (EG) Schneller Brueter. Between KfK, INTERATOM, and the French Commissariat a l'Energie Atomique contracts were concluded in 1977 about close cooperation in the Fast Breeder field, with association of the Belgian and Dutch partners. The results of research and development activities carried out by the DeBeNe partners in 1981 have been compiled in this report. The report begins with a short survey of the fast reactor plants, followed by an R and D summary. The bulk of the report gives more detailed information about those plants and about results reported by the Working Groups of the R and D Program Working Committee of the Fast Breeder Project. In an additional chapter a survey is given of international cooperation. (author)

  3. Sodium flow measurement in large pipelines of sodium cooled fast breeder reactors with bypass type flow meters

    International Nuclear Information System (INIS)

    Rajan, K.K.; Jayakumar, T.; Aggarwal, P.K.; Vinod, V.

    2016-01-01

    Highlights: • Bypass type permanent magnet flow meters are more suitable for sodium flow measurement. • A higher sodium velocity through the PMFM sensor will increase its sensitivity and resolution. • By modifying the geometry of bypass line, higher sodium velocity through sensor is achieved. • With optimized geometry the sensitivity of bypass flow meter system was increased by 70%. - Abstract: Liquid sodium flow through the pipelines of sodium cooled fast breeder reactor circuits are measured using electromagnetic flow meters. Bypass type flow meter with a permanent magnet flow meter as sensor in the bypass line is selected for the flow measurement in the 800 NB main secondary pipe line of 500 MWe Prototype Fast Breeder Reactor (PFBR), which is at the advanced stage of construction at Kalpakkam. For increasing the sensitivity of bypass flow meters in future SFRs, alternative bypass geometry was considered. The performance enhancement of the proposed geometry was evaluated by experimental and numerical methods using scaled down models. From the studies it is observed that the new configuration increases the sensitivity of bypass flow meter system by around 70%. Using experimentally validated numerical tools the volumetric flow ratio for the bypass configurations is established for the operating range of Reynolds numbers.

  4. Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

    1991-01-01

    The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

  5. Development of standards and investigation of safety examination items for advancement of safety regulation of fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purposes of this study are to prepare the fuel technical standard and the structure and materials standard of fast breeder reactors (FBRs), and to develop the requirements in a reactor establishment permission. The objects of this study are mainly the Monju high performance core and a demonstration FBR. In JFY 2012, the following results were obtained. As for the fuel technical standard, the fuel technical standard adapting the examination of integrity of the FBR fuels was prepared based on the information and data obtained in this study. As for the structure and material standard, the investigation of the revised parts of the standard was carried out. And as for the examination of the safety requirements, safety evaluation items for the future FBR plant and the fission products to be considered in a reactor establishment permission were investigated and examined. (author)

  6. Test facility for auxiliary cooling system (ACS) of fast breeder reactor for Power Reactor and Nuclear Fuel Development Corporation (PNC)

    International Nuclear Information System (INIS)

    1983-01-01

    In preparation of constructing ''Monju'', a prototype fast breeder reactor, PNC has been pushing forward its research and development projects and the ACS was constructed under these projects. The auxiliary cooling system is an important engineered safety feature, and is used for safe removal of heat from the reactor at the shutdown. The ACS serves as a means of testing and assessing the auxiliary cooling system for the ''Monju'' and is designed and manufactured to have one fifth capacity of the Monju. The air heat exchanger and the ACS system was designed to withstand higher temperature range of the conventional design code (MITI-501), and finned tubes were applied for effective heat removal. Preheating system was designed to heat up the whole system over 200 0 C within 20 hours to prevent sodium from freezing. Basic performance of ACS was verified satisfactorily by a series of performance tests, such as start up test, flow rate measurement and preheating test before delivery. The experience from designing and construction of ACS and data obtained by these tests will be very instructive for designing and construction of the ''Monju''. (author)

  7. 3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

    International Nuclear Information System (INIS)

    Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

    2013-01-01

    Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

  8. Development of a new measurement method for fast breeder reactor fuel burnup using a shielded ion microprobe analyzer

    International Nuclear Information System (INIS)

    Mizuno, M.; Enokido, Y.; Itaki, T.; Kono, K.; Unno, I.; Yamanouchi, S.

    1985-01-01

    A new method of burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. The method is based on the isotope analysis of uranium, plutonium, and fission products in irradiated mixed oxide fuel by means of secondary ion mass spectrometry (SIMS). Fourteen samples irradiated in the Japanese experimental fast reactor JOYO were examined. The maximum local burnup of JOYO MK-I core fuels was about5.1 at. %. The axial burnup distribution of the fuel pin was in good agreement with that of the sibling pin in the same subassembly, measured by surface ionization mass spectrometry, which requires the chemical separation of fission products and heavy metals. The new method facilitates the rapid and accurate measurement of fast breeder reactor fuel burnup without human radiation exposure during sample preparation and analysis

  9. Use of EBR-II as a principal fast breeder reactor irradiation test facility in the U.S

    International Nuclear Information System (INIS)

    Staker, R.G.; Seim, O.S.; Beck, W.N.; Golden, G.H.; Walters, L.C.

    1975-01-01

    The EBR-II as originally designed and operated by the Argonne National Laboratory was successful in demonstrating the operation of a sodium-cooled fast breeder power plant with a closed fuel reprocessing cycle. Subsequent operation has been as an experimental facility where thousands of irradiation tests have been performed. Conversion to this application entailed the design and fabrication of special irradiation subassemblies for in-core irradiations, additions to existing facilities for out-of-core irradiations, and additions to existing facilities for out-of-core experiments. Experimental subassemblies now constitute about one third of the core, and changes in the core configuration occur about monthly, requiring neutronic and thermal-hydraulics analyses and monitoring of the reactor dynamic behavior. The surveillance programs provided a wealth of information on irradiation induced swelling and creep, in-reactor fracture behavior, and the compatibility of materials with liquid sodium. (U.S.)

  10. Role of fast breeder reactors in the generation of electricity in the coming decades in India

    International Nuclear Information System (INIS)

    Paranjpe, S.R.; Char, N.L.; Sundaram, C.V.

    1982-01-01

    For a country like India which must industrialise rapidly to provide a modest living standard above the poverty line for its huge population, there is no alternate to developing an alternate energy source by the turn of the century so as to permit a smooth transition from fossil fuels to this new source of energy by the year 2050. If such a thing can be achieved, the fossil fuel reserves could be preserved for more important uses in the following centuries. Fast breeders fuelled with mixed carbides capable of achieving a short doubling time of the order of 10 years or less appears to be the solution which is attainable if the requisite priority is assigned and concerted efforts are made to achieve the objective

  11. Observations of the behaviour of gas in the wake behind a corner blockage in fast breeder reactor subassembly geometry

    International Nuclear Information System (INIS)

    Fukuzawa, Y.

    1979-07-01

    Observations were made of gas behaviour in the wake behind a 21% corner blockage in the subassembly geometry of a liquid metal fast breeder reactor. The test section used represented one half of the reactor fuel subassembly, divided along the vertical plane of symmetry through the blockage. A glass wall occupied the position of this plane. Water was allowed to flow between glass rods simulating fuel pins, the velocity being changed from 1.2 to 4.5 m/s. Argon was injected into the wake or into the flow upstream of the blockage, the injection rate being changed from 1 to 230 Ncm 3 /s (standard temperature and pressure). From the present experiment, the following is evident: The gas is accumulated in the wake behind the blockage, forming a gas cavity. The flow patterns of the two-phase mixture in the wake are classified into three types, depending on the liquid velocity. In the lower velocity range, a gas cavity cannot be present at rest, rising up through the wake as a single bubble due to buoyancy. In the higher velocity range, the gas cavity is broken up by the liquid flow forces, only small gas bubbles circulating in the wake. In the velocity range in between, the gas cavity is present in the wake. The cavity size depends on the gas injection rate and on the liquid velocity. From the results, the possibility of fuel failure caused by fission gas release at a blockage in the fast breeder reactor can be considered to depend on the operating conditions of the reactor, specially on the coolant velocity. (orig.) [de

  12. Economic incentives of short out-of-reactor time for fast breeder reactor fuel

    International Nuclear Information System (INIS)

    Bentley, B.W.; Haffner, D.R.

    1975-01-01

    Economic benefits (primarily reduced uranium ore and enrichment expenditures) can be realized by reducing the LMFBR out-of-reactor fuel cycle time only if process cost penalties and R and D costs can be minimized. The results of the evaluation presented show the potential gross benefits of reducing the out-of-reactor time and the effects of various associated cost penalties on these benefits. The gross benefit results estimate the potential savings in electrical power generation in the next 50 years using constant 1975 dollars and discounting the costs at 7 1 / 2 percent per year

  13. The design and fabrication of an optical periscope for core viewing of fast breeder test reactor (FBTR)

    International Nuclear Information System (INIS)

    Das, N.C.; Sanjiva Kumar; Udupa, D.V.; Shukla, R.P.; Kadu, A.M.; Modi, R.K.

    2004-08-01

    A FBTR (Fast Breeder Test Reactor) periscope has been designed and fabricated indigenously for viewing and photography/ video recording the objects in the reactor core. The periscope consists of a scanning prism mechanism, zoom lens objective, a system of relay lenses and an eyepiece sub-assembly for viewing the objects. The objective of the periscope is a zoom lens system for obtaining a continuously varying magnification from 2X to 5X. Zoom lens objective system has a variable focal length from 100 mm to 250 mm with an aperture varying from 10 mm to 25 mm respectively. This covers a semi- field angle of 3 deg for the objective lens of focal length of 250 mrn and 4 deg for the objective of focal length of l00 mm. Two prisms of 45 deg -90 deg -45 deg types are used for scanning the object space in vertical direction. One prism is fixed, whereas the prism facing the object can be rotated about the horizontal axis through an angle of 110 deg. The rotation of the entire periscope assembly along the vertical axis scans the object space on the horizontal plane. The combination of these two rotations is used to scan the field of interest. It may be noted here that it is absolutely essential to introduce a Pechan prism before each eyepiece. Pechan prism is used for the rotation of the image, which is produced due to the rotation of the scanning prisms. The measured value of the linear resolution of the instrument is 0.7 mm at an object distance of 2.5 meter from the zoom lens objective system. The periscope has two arm labeled I and II. The arm I is used for visual inspection, while the arm II is used for video recording/photography. The periscope will be used as an in-service instrument for Fast Breeder Test Reactor, IGCAR, Kalpakkam. (author)

  14. Effect of turbulent natural convection on sodium pool combustion in the steam generator building of a fast breeder reactor

    International Nuclear Information System (INIS)

    Karthikeyan, S.; Sundararajan, T.; Shet, U.S.P.; Selvaraj, P.

    2009-01-01

    A computational model is proposed to simulate sodium pool combustion considering the effect of turbulent natural convection in a vented enclosure of the steam generator building (SGB) of a fast breeder reactor. The model is validated by comparing the simulated results with the experimental results available in literature for sodium pool combustion in a CSTF vessel. After validation, the effects of vents and the location of the pool on the burning rate of sodium and the associated heat transfer to the walls are studied in an enclosure comparable in size to one floor of the steam generator building. In the presence of ventilation, the burning rate of sodium increases, but the total heat transferred to the walls of the enclosure is reduced. It is also found that the burning rate of sodium pool and the heat transfer to the walls of the enclosures vary significantly with the location of sodium pool.

  15. Analysis of an out-of-pile experiment for materials redistribution under core disruptive accident condition of fast breeder reactors

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Ninokata, Hisashi; Shimizu, Akinao

    1995-01-01

    Calculation of one of the SIMBATH experiments was performed using the SIMMER-II code. The experiments were intended to simulate the fuel pin disintegration, the molten materials relocation and following materials redistribution that could occur during core disruptive accidents assumed in fast breeder reactors. The calculation by SIMMER-II showed that the incorporated step-wise fuel pin disintegration model and the modified particle jamming model were capable of reproducing the course of materials relocation within the identified ranges of the parameters which governed the blockages formation, i.e. the characteristic radius of solid particles jamming and/or sieving out in the flow and the effective particle viscosity. In particular the final materials redistribution calculated by SIMMER-II very well reproduced the experiment. This fact made it possible to interpret theoretically the mechanisms of flow blockages formation and related materials redistribution. (author)

  16. Fabrication and quality assurance of some important components and sub-assemblies for Prototype Fast Breeder Reactor (PFBR) project

    International Nuclear Information System (INIS)

    Dutta, N.G.; More, S.S.

    2010-01-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500 MWe prototype fast breeder reactor (PFBR) at Kalpakkam, Chennai. In this very important and prestigious national programmed M/s Kay Bouvet Engg. Pvt. Ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies. M/s KBEPL is engaged in manufacturing, quality assurance and supply of many subassemblies of PFBR like under water trolley (UWT), shielding door, container and container storage rack (CSR), vessel in fuel transfer cell (FTC), personnel air lock (PAL), emergency air lock (EAL) and material air lock (MAL), absorber rod drive mechanism (ARDM) flask assembly and carriage in MAL etc. Two partition doors and four nos. of embedded parts (SS 304L) have already been supplied to Bhavini. The paper deals with manufacturing and Q.A. activities being carried out for supply of these important assemblies to PFBR projects. (author)

  17. A knowledge based on-line diagnostic system for the fast breeder reactor KNKII

    International Nuclear Information System (INIS)

    Eggert, H.; Scherer, K.P.; Stiller, P.

    1989-01-01

    In the nuclear research center at Karlsruhe, a diagnostic expert system is developed to supervise a fast breeder process (KNKII). The problem is to detect critical phases in the beginning state before fault propagation. The expert system itself is integrated in a computer network (realized by a local area network), where different computers are involved as special detection systems (for example acoustic noise, temperature noise, covergas monitoring and so on), which produce partial diagnoses, based on intelligent signal processing techniques like pattern recognition. Additional to the detection systems a process computer is integrated as well as a test computer, which simulates hypothetical and real fault data. On the logical top level the expert system manages the partial diagnoses of the detection systems with the operating data of the process computer and to produce a final diagnosis including the explanation part for operator support. The knowledge base is developed by typical Artificial Intelligence tools. Both fact based and rule based knowledge representations are stored in form of flavors and predications. The inference engine operates on a rule based approach. Specific detail knowledge, based on experience about any years, is available to influence the decision process by increasing or decreasing of the generated hypotheses. In a meta knowledge base, a rule master triggers the special domain experts and contributes the tasks to the specific rule complexes. Such a system management guarantees a problem solving strategy, which operates event triggered and situation specific in a local inference domain. (author). 3 refs, 6 figs, 2 tabs

  18. Few-group constants for the calculation of ksub(eff) and Δ(1/ksub(eff)) of fast breeder reactors

    International Nuclear Information System (INIS)

    Svarny, J.

    1978-01-01

    A theoretical and numerical analysis is presented of the linear and bilinear weighting of group constants. Special attention is paid to error accumulation in the few-group calculations of reactivity (ksub(eff)) and its first order perturbations caused by inaccuracies in weighting functions. Some theoretical conclusions are supported by calculations of the BN-600 fast breeder reactor. (author)

  19. Status and programme of development of the fast breeder reactor systems in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Daeunert, U.; Kessler, G.

    1977-01-01

    Owing to the dependence of the Federal Republic of Germany on imports of raw materials for energy purposes, special importance is given to nuclear energy for reasons of energy policy. Against the background of a nuclear power station capacity amounting to 24,000 MWe in 1976 (power stations being either on order, under construction or actually in operation), the Fast Breeder Reactor (FBR) has quite outstanding significance in this country where the economical utilization both of uranium and of the plutonium accumulating in light water reactors is concerned. Work in this sector started in the Federal Republic of Germany in 1960. Milestones on the path of development were the initial activities on the plutonium critical SNEAK at Karlsruhe, the successful operation of the KNK 20 MWe experimental reactor (sodium-cooling) with the KNK II (plutonium core) follow-up stage, the construction of the 300 MWe FBR prototype nuclear power station at Kalkar, as well as the initial planning work carried out in connection with a 1,300 MWe FBR demonstration nuclear power station. The provision of a viable infrastructure for safety assessment must go hand in hand with the development of new reactor systems. For this reason, prototype reactors in the Federal Republic of Germany are already subjected to the same strict licensing procedures under nuclear legislation as are commercial nuclear power stations. After several years' practical experience, it is becoming evident that this approach was the right one to take. Already early on, it was recognized in the Federal Republic of Germany that the execution of long-term technological major projects were increasingly exceeding the capacities of unaided development by single nations. The SNR 300 FBR prototype project combines the efforts of the governments, power supply companies, reactor manufacturers and nuclear research centres of the Federal Republic of Germany and of the Kingdoms of both Belgium and the Netherlands. The Agreement signed

  20. Proposed method of the modeling and simulation of corrosion product behavior in the primary cooling system of fast breeder reactors

    International Nuclear Information System (INIS)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    2011-01-01

    Radioactive corrosion products (CP) are main cause of personal radiation exposure during maintenance without fuel failure in FBR plants. In order to establish the techniques of radiation dose estimation for worker in radiation-controlled area, Program SYstem for Corrosion Hazard Evaluation code 'PSYCHE' has been developed. The PSYCHE is based on the Solution-Precipitation model. The CP transfer calculation using the Solution-Precipitation model needs a fitting factor for the calculation of the precipitation of CP. This fitting factor must be determined based on the measured values in reactors that have operating experience. For this reason, the inability to make accurate predictions for reactor without measured values is a major issue. In this study, in addition to existing Solution-Precipitation model in PSYCHE, a transfer-model of CP species in particle form was applied to calculations of CP behavior in the primary cooling system of fast breeder reactor MONJU. Based on the calculated results, we estimated the contribution of CP deposition in the particle-form. It was suggested that the improved model including transfer-model of CP species in particle-form could be used for evaluation of CP transfer and radiation-source distribution in place of conventional Solution-Precipitation model with fitting factor in the PSYCHE. Moreover, it was predicted that CP particles would tend to be deposited in region with high-flow rate of coolant. (author)

  1. Tests of the heat transfer characteristic of air cooler during cooling by natural convection of the Fast Breeder Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purpose of this study is to confirm the heat transfer characteristics of the air cooler (AC) of the Fast Breeder Reactor(FBR) which has a function to remove the residual heat of the reactor by heat exchange between sodium and air in natural convection region if electric power would be lost. In order to confirm the characteristics of the AC installed in the FBR plant, the heat transfer test by using the AC which is installed in the sodium test loop owned by Toshiba Corporation has been planned. In this study, the heat transfer characteristic tests were performed by using the AC in sodium test loop, and the CFD analyses were conducted to evaluate the test results and the heat transfer characteristics of the plant scale AC at the condition of natural convection. In addition, the elemental tests to confirm the influence of the heat transfer tube placement by using the heat transfer tube of the same specification as the AC of Monju were performed. (author)

  2. Theoretical investigations of the meltoff and resolidification process of fuel claddings during accidents in liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Angerer, G.

    1978-08-01

    During loss-of-coolant-flow accidents in liquid metal cooled fast breeder reactors with failure to scram the fuel claddings will melt after boiling and evaporation of the coolant. The CMOT model presented here describes the subsequent process of relocation and resolidification of the molten claddings. The basic thermohydrodynamics equations of the two-phase flow of cladding material and sodium vapor are solved numerically by differential approximations in a Eulerian reference net. The results calculated by the model improved the insight into the dynamics of the cladding relocation process. Here are the main results: - Shortly after the onset of cladding relocation large waves of molten cladding material are generated. The motion of these waves contributes considerably to the material transport. - The dynamics of cladding relocation exhibits strong local incoherences. - The formation of cladding blockages observed at the ends of the fuel region is confirmed by the calculations. - In case of incoherent cladding meltoff less cladding material is transported upwards. - Cladding relocation strongly depends on the axial pressure drop and the underlying friction factor correlations. Recalculation of the R5 loss-of-coolant-flow experiment performed in the U.S. TREAT test reactor is in good agreement with the experimental data. (orig./HP) 891 HP [de

  3. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and Netherlands - February 1983

    International Nuclear Information System (INIS)

    Hueper, R.

    1983-01-01

    In 1967 and 1968 the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (''DeBeNe'') agreed to develop breeder reactors in a joint program. The following research organizations have taken part in this effort: Kernforschungszentrum Karlsruhe (KfK); INTERATOM, Bergisch Gladbach; ALKEM, Wolfgang near Hanau; SCK/CEN, Mol; Belgonucleaire, Brussels; ECN, Petten; TNO, Apeldoorn; NERATOOM, The Hague. The three German institutions mentioned above have been connected since 1977 in the Entwicklungsgemeinschaft (EG) Schneller Brueter. Kfk, INTERATOM, and the French Commissariat a l'Energie Atomique entered into contracts in 1977 about close cooperation in the fast breeder field, to which the Belgian and Dutch partners acceded. The results of activities carried out by the DeBeNe partners in 1982 have been compiled in this report. The report begins with a survey of the fast reactor plants and the political background, followed by an R and D summary. In an additional chapter, a survey is given of international cooperation in 1982

  4. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands - February 1985

    International Nuclear Information System (INIS)

    Hueper, R.

    1986-01-01

    In 1967 and 1968, the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (''DeBeNe'') agreed to develop breeder reactors in a joint program. The following research organizations have taken part in this effort: Kernforschungszentrum Karlsruhe (KfK); INTERATOM, Bergisch Gladbach; ALKEM, Wolfgang near Hanau; SCK/CEN, Mol; Belgonucleaire, Brussels; ECN, Petten; TNO, Apeldoorn; NERATOOM, The Hague. The three Germany institutions mentioned above have been associated since 1977 in the Entwicklungsgemeinschaft (EG) Schneller Brueter. KfK, INTERATOM, and the French Commissariat a l'Energie Atomique entered into contracts in 1977 about close cooperation in the fast breeder field, to which the Belgian and Dutch partners acceded. The results of activities carried out by the DeBeNe partners in 1984 have been compiled in this report. The report begins with a survey of the fast reactor plants followed by a R and D summary. In an additional chapter, a survey is given of international cooperation in 1984

  5. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and The Netherlands - February 1984

    International Nuclear Information System (INIS)

    Hueper, R.

    1984-01-01

    In 1967 and 1968 the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (''DeBeNe'') agreed to develop breeder reactors in a joint program. The following research organizations have taken part in this effort: Kernforschungszentrum Karlsruhe (KfK); INTERATOM, Bergisch Gladbach; ALKEM, Wolgang near Hanau; SCK/CEN, Mol; Belgonucleaire, Brussels; ECN, Petten; TNO, Apeldoorn; NERATOOM, The Hague. The three German institutions mentioned above have been connected since 1977 in the Entwicklungsgemeinschaft (EG) Schneller Brueter. KfK, INTERATOM, and the French Commissariat a l'Energie Atomique entered into contracts in 1977 about close cooperation in the fast breeder field, to which the Belgian and Dutch partners acceded. The results of activities carried out by the DeBeBe partners in 1983 have been compiled in this report. The report begins with a survey of the fast reactor plants followed by an R and D summary. In an additional chapter, a survey is given of international cooperation in 1983

  6. Report to the Congress: liquid metal fast breeder reactor program--past, present, and future, Energy Research and Development Administration

    International Nuclear Information System (INIS)

    1975-01-01

    The past, present, and future of the liquid metal fast breeder reactor (LMFBR) program, the Nation's highest priority energy program, are studied. ERDA anticipates that the operation of the first large commercial breeder will start in 1987, and that 186 commercial-size breeders will be in operation by the year 2000. The breeder program is made up of six major areas, each dealing with an important element of technology: reactor physics; fuels and materials; fuel recycle; safety; component development; plant experience; and facilities used in the LMFBR program. ERDA is implementing a new system for administering, managing, and controlling the breeder program that will provide increased program visibility and control. Federal funding for breeder development was $168 million in FY 1971, accounting for 40% of the total Federal R and D energy budget; in FY 1976 Federal funding for this program will be $474 million, only 26% of total Federal funding for energy research. Besides Federal funds, over half a billion dollars have been or will be invested by industry over the next 5 to 10 years to develop the breeder and to build a demonstration plant. Five other nations--the United Kingdom, France, Japan, West Germany, and the Soviet Union--have a high priority national energy program for developing the LMFBR. These foreign breeder programs could contribute important data and information to the U.S. program

  7. 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

  8. Experiences with fast breeder reactor education in laboratory and short course settings

    International Nuclear Information System (INIS)

    Waltar, A.E.

    1983-01-01

    The breeder reactor industry throughout the world has grown impressively over the last two decades. Despite the uncertainties in some national programs, breeder reactor technology is well established on a global scale. Given the magnitude of this technological undertaking, there has been surprisingly little emphasis on general breeder reactor education - either at the university or laboratory level. Many universities assume the topic too specialized for including appropriate courses in their curriculum - thus leaving students entering the breeder reactor industry to learn almost exclusively from on-the-job experience. The evaluation of four course presentations utilizing visual aids is presented

  9. The radiological consequences of notional accidental releases of radioactivity from fast breeder reactors: the influence of the meteorological conditions

    International Nuclear Information System (INIS)

    Hemming, C.R.; Hallam, J.; Kelly, G.N.

    1980-04-01

    The radiological consequences of a wide range of notional accidental releases from a 1300 MW(e) LMFBR (liquid Metal-cooled Fast Breeder Reactor) were assessed in a previous study by the National Radiological Protection Board. In the present study the sensitivity of the predicted consequences to the meteorological conditions was investigated. The influence of the wind direction, and hence the distribution of the exposed population, and of atmospheric stability at the time of the release are considered separately. Conclusions are reached on the precision required in specifying atmospheric stability conditions to estimate reliably the distribution of possible consequences following a given release of activity. The probability distributions of consequences for selected releases from two locations are evaluated taking account of the frequency distribution of wind direction and atmospheric stability at each location. The variation is large and demonstrates the importance of taking account of the whole range of meteorological conditions when assessing the risk presented by an accidental release of activity. (author)

  10. Shielding design study of the demonstration fast breeder reactor. 2. Shielding design on the basis of the JASPER analysis

    International Nuclear Information System (INIS)

    Suzuoki, Zenro; Tabayashi, Masao; Handa, Hiroyuki; Iida, Masaaki; Takemura, Morio

    2000-01-01

    Conceptual shielding design has been performed for the Demonstration Fast Breeder Reactor (DFBR) to achieve further optimization and reduction of the plant construction cost. The design took into account its implications in overall plant configuration such as reduction of shields in the core, adoption of fission gas plenum in the lower portion of fuel assemblies, and adoption of gas expansion modules. Shielding criteria applied for the design are to secure fast neutron fluence on in-vessel structures as well as responses of the nuclear instrumentation system and to restrict secondary sodium activation. The design utilized the cross sections and the one- and two-dimensional discrete ordinates transport codes, whose verification had been performed by the JASPER experiment analysis. Correction factors yielded by the JASPER analysis were applied to the design calculations to obtain design values with improved accuracy. Design margins, which are defined by the ratios of the design criteria to the design values, were more than two for all shielding issues of interest, showing the adequacy of the shielding design of the DFBR. (author)

  11. Techniques for processing remote field eddy current signals from bend regions of steam generator tubes of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thirunavukkarasu, S. [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India); Rao, B.P.C., E-mail: bpcrao@igcar.gov.in [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India); Jayakumar, T.; Raj, Baldev [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)

    2011-04-15

    Steam generator (SG) is one of the most critical components of sodium cooled fast breeder reactor. Remote field eddy current (RFEC) technique has been chosen for in-service inspection (ISI) of these ferromagnetic SG tubes made of modified 9Cr-1Mo steel (Grade 91). Expansion bends are provided in the SGs to accommodate differential thermal expansion. During ISI using RFEC technique, in expansion bend regions, exciter-receiver coil misalignment, bending stresses, probe wobble and magnetic permeability variations produce disturbing noise hindering detection of defects. Fourier filtering, cross-correlation and wavelet transform techniques have been studied for noise reduction as well as enhancement of RFEC signals of defects in bend regions, having machined grooves and localized defects. Performance of these three techniques has been compared using signal-to-noise ratio (SNR). Fourier filtering technique has shown better performance for noise reduction while cross-correlation technique has resulted in significant enhancement of signals. Wavelet transform technique has shown the combined capability of noise reduction and signal enhancement and resulted in unambiguous detection of 10% of wall loss grooves and localized defects in the bend regions with SNR better than 7 dB.

  12. Effect of geometric factors on performance of a sodium to air heat exchanger in a fast breeder reactor

    International Nuclear Information System (INIS)

    Kannan, K.; Vinod, V.; Padmakumar, G.; Rudramoorthy, R.; Rajan, K.K.

    2015-01-01

    Highlights: • A heat exchanger analysis (HE) before scale up reduces excess heat transfer area. • Representative Elementary Volume analysis of a HE speeds up the solution. • The error in air temperature rise prediction by numerical across HE is within 5%. • When both pitches are reduced, the maximum increase in heat flux is experienced. • The experience has resulted in better design of next level heat exchangers. - Abstract: Prototype fast breeder reactor (PFBR) has a safety grade decay heat removal system whose performance depends on the effective functioning of natural convection heat exchangers called sodium to air heat exchangers. The development of Representative Elementary Volume (REV) model for the sodium to air heat exchanger is necessary to envisage its design and to study the effect of various factors for continuous improvement in design. With a Representative Elementary Volume, the hydrodynamic and heat transfer characteristics of the heat exchanger was studied and the results agree well with experimental data. The effect of longitudinal pitch and transverse pitch on the heat exchanger performance has been studied and an improvement of 22% in heat transfer is predicted

  13. Simulation of radioactive corrosion product in primary cooling system of Japanese sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    2012-01-01

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54 Mn and 60 Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54 Mn was estimated to constitute approximately 20% and 60 Co approximately 40% in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO. (author)

  14. Status of the fast breeder reactor development in the Federal Republic of Germany, Belgium and The Netherlands, February 1981

    Energy Technology Data Exchange (ETDEWEB)

    Hueper, R; Friese, R [Kernforschungszentrum Karlsruhe, Karlsruhe (Germany)

    1981-05-01

    In 1967 and 1968 the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (DeBeNe) agreed to develop, in a joint program, breeder reactors to the point of commercial maturity. The following research organizations take part in this effort: Kernforschungszentrum Karlsruhe (KfK); INTERATOM, Bergisch Gladbach; ALKEM, Wolfgang near Hanau; SCK/CEW, Mol; Belgonucleaire, Brussels; ECN, Petten; TWO, Apeldoorn; NERATOOM, The Hague. The results of research and development activities carried out by the DeBeNe partners in 1980 have been compiled in this report. The report begins with a review of the energy policy background, followed by an R and D summary. The bulk of the report following next is organized by the Working Groups of the R and D Program Working Committee of the Fast Breeder Project; additional chapters provide information about the operation of KNK II and the construction of SNR 300. In the annexes a survey is given of international cooperation.

  15. Status of the fast breeder reactor development in the Federal Republic of Germany, Belgium and The Netherlands, February 1981

    International Nuclear Information System (INIS)

    Hueper, R.; Friese, R.

    1981-01-01

    In 1967 and 1968 the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (DeBeNe) agreed to develop, in a joint program, breeder reactors to the point of commercial maturity. The following research organizations take part in this effort: Kernforschungszentrum Karlsruhe (KfK); INTERATOM, Bergisch Gladbach; ALKEM, Wolfgang near Hanau; SCK/CEW, Mol; Belgonucleaire, Brussels; ECN, Petten; TWO, Apeldoorn; NERATOOM, The Hague. The results of research and development activities carried out by the DeBeNe partners in 1980 have been compiled in this report. The report begins with a review of the energy policy background, followed by an R and D summary. The bulk of the report following next is organized by the Working Groups of the R and D Program Working Committee of the Fast Breeder Project; additional chapters provide information about the operation of KNK II and the construction of SNR 300. In the annexes a survey is given of international cooperation

  16. 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.

  17. A vision of inexhaustible energy: The fast breeder reactor in Swedish nuclear power history 1945-80

    International Nuclear Information System (INIS)

    Fjaestad, Maja

    2010-01-01

    The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and 1960s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy, and thereby connected it to utopian ideas about an eternal supply of energy, Furthermore. the ideas of breeder reactors were a vital part of the post-war visions about the nuclear future. This dissertation investigates the plans for breeder reactors in Sweden, connecting them to the contemporary development of nuclear power with heavy or light water and the discussions of nuclear weapons, as well as to the general visions of a prosperous technological future. The history of the Swedish breeder reactor is traced from high hopes in the beginning, via the fiasco of the Swedish heavy water program, partly focusing on the activities at the company AB Atomenergi and investigating how it planned and argued for its breeder program and how this was received by the politicians. The story continues into the intensive environmental movement in the 1970s, ending with the Swedish referendum on nuclear energy in 1980, which can be seen as the final point for the Swedish breeder. The thesis discusses how the nuclear breeder reactor was transformed from an argument for nuclear power to an argument against it. The breeder began as a part of the vision of a society with abundant energy, but was later seen as a threat against the new sustainable world. The nuclear breeder reactor is an example of a technological vision that did not meet its industrial expectations. But that does not prevent the fact that breeder was an influential technology in an age where important decisions about nuclear energy were made. The thesis argues that important decisions about the contemporary reactors were taken with the idea that they in a foreseeable future would be replaced with the efficient breeder. And the last word on the breeder reactor is not said - today, reactor engineers around the world are

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  19. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and The Netherlands

    International Nuclear Information System (INIS)

    Hueper, R.

    1990-01-01

    The results of activities carried out by the DeBeNe partners in 1989 have been compiled in this report. The report begins with a survey of fast reactor plants, which is followed by an R and D summary. In an additional chapter, a survey is presented of international cooperation in 1989. Effective January 1990, KfK activities in the area of fast reactors have been incorporated in the Nuclear Safety Research Project. (author)

  20. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    International Nuclear Information System (INIS)

    1980-01-01

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases

  1. Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

    International Nuclear Information System (INIS)

    DeVault, G.P.; Bell, C.R.

    1985-01-01

    Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

  2. The 1500 MW fast breeder reactor the double envelope-vessel anchored in concrete

    International Nuclear Information System (INIS)

    Bolvin, M.

    1981-01-01

    This paper givers an account of EDF investigations to reduce the investment costs of the 1500 MW Fast Reactor (RNR 1500) without prejudice to the safety requirements. It deals with the double envelope-vessel, designed to minimize radiation consequences in the event of accidental leakage in the main vessel. In the Fast Reactors in operation (PHOENIX), under construction (CRYS-MALVILLE), and under project (NR 1500), the double envelope-steel vessel hangs down from the upper part of the reactor block, its weight being approximately 300 t. In the new design, the vessel is fixed into the concrete which supports the main vessel, by means of steel anchors. A thermal insulation isolates it from the main vessel. The installation of coils in the concrete, next to the lining, allows for water circulation to cool the concrete. (orig./GL)

  3. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    1978-07-01

    This contribution is prepared for the answer to the questionnaire of working group 5, subgroup B. B.1. is the short review of the fast breeder fuel cycles based on the reference large commercial Japanese LMFBR. The LMFBRs are devided into two types. FBR-A is the reactor to be used before 2000, and its burnup and breeding ratio are relatively low. The reference fuel cycle requirement is calculated based on the FBR-A. FBR-B is the one to be used after 2000, and its burnup and breeding ratio are relatively high. B.2. is basic FBR fuel reprocessing scheme emphasizing the differences with LWR reprocessing. This scheme is based on the conceptual design and research and development work on the small scale LMFBR reprocessing facility of Japan. The facility adopts a conventional PUREX process except head end portions. The report also describes the effects of technical modifications of conventional reprocessing flow sheets, and the problems to be solved before the adoption of these alternatives

  4. Selection of engineering materials and fabrication of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Patriarca, P.

    1975-01-01

    Information is presented graphically and pictorially concerning the need for nuclear power; basic nuclear concepts including BWR, PWR, HTGR, and LMFBR; the fissioning process; nuclear reactor fuel; fabrication of reactor vessels for LMFBR's; fabrication of intermediate heat exchangers for LMFBR's; piping fabrication for LMFBR's; transition welds; steam generators for LMFBR demonstration plants worldwide; stress corrosion cracking of steam generator materials and weldments; post--test examination of the Alco/BLH sodium-heated steam generator; alternate steam generator designs; and alternate structural materials. (DCC)

  5. Problems specific to the piping of sodium-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Vrillon, B.; Befre, J.; Schaller, K.

    1975-01-01

    A certain number of specific problems arising in connection with the sodium pipes in fast neutron reactors, especially those of large diameter, are presented. The supporting system must be designed to achieve the best compromise among stresses due to weight and various stresses of thermal origin. Large-scale experimental studies carried out on actual elements of the intermediate circuit of the Phenix reactor showed that the circuits can withstand considerable deformation collapse of the walls without danger of leakage. Protection studies against earthquakes are mentionned [fr

  6. Progress report on fast breeder reactor development in Japan, June-March, 1979

    International Nuclear Information System (INIS)

    1979-11-01

    The experimental fast reactor ''Joyo'' started the second cycle operation of 50 MW on January 12, and finished it on February 26. The operation was very stable throughout this period. The preparation to raise the output to 75 MW has been in progress since then in parallel with the periodic inspection. Experimental fuel subassemblies and a control rod used for the 50 MW operation were removed. Installations of a fuel storage and handling facility, a water cooling and purifying plant, and an in-water cutting equipment were completed. The works related to the analysis of the core characteristics are reported. The construction preliminary design (2) of the prototype fast reactor ''Monju'' was finished. The wind tunnel experiment and the calculation of earthquake response to artificial seismic waves are being carried on. The works of developing codes and the surveys on the construction site are reported. The fourth preliminary design of the demonstration fast reactor was completed. The research and development of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structures and structural materials, safety and steam generators are reported. The technological investigation into foreign LMFBRs was finished. (Kako, I.)

  7. Modelling of temperature distribution and temperature pulsations in elements of fast breeder reactor

    International Nuclear Information System (INIS)

    Sorokin, A.P.; Bogoslovskaia, G.P.; Ushakov, P.A.; Zhukov, A.V.; Ivanov, Eu.F.; Matjukhin, N.M.

    2004-01-01

    From thermophysical point of view, integrated configuration of liquid metal cooled reactor has some limitations. Large volume of mixing chamber causes a complex behavior of thermal hydraulic characteristics in such facilities. Also, this volume is responsible for large-scale eddies in the coolant, existence of stagnant areas and flow stratification, occurrence of temperature non-uniformity and pulsation of coolant and structure temperatures. Temperature non-uniformities and temperature pulsations depend heavily even on small variations in reactor core design. The paper presents some results on modeling of thermal hydraulic processes occurring in liquid metal cooled reactor. The behavior of following parameters are discussed: temperature non-uniformities at the core output and related temperature pulsations; temperature pulsations due to mixing of sodium jets at different temperatures; temperature pulsations arising if a part of loop (circuit) is shut off; temperature non-uniformities and pulsation at the core output and related temperature pulsation; temperature pulsations due to mixing of sodium jets at different temperatures; temperature pulsations arising if a part of loop (circuit) is shut off; temperature non-uniformities and pulsation of temperature during transients and during transition to natural convection cooling. Also, the issue of modeling of temperature behavior in compact arrangement of fast reactor fuel pins using water as modeling liquid is considered in the paper. One more discussion is concerned with experimental method of modeling of liquid metal mixing with the use of air. The method is based on freon tracer technique. The results of simulation of the thermal hydraulic processes mentioned above have been analyzed, that will allow the main lines of the study to be determined and conclusion to be drawn regarding the temperature behavior in fast reactor units. (author)

  8. Heterogeneity of structure and properties of 12Cr18Ni10Ti and 08Cr16Ni11Mo3 stainless steels irradiated up to high damaging doses in reactor Bn-350

    International Nuclear Information System (INIS)

    Maksimkin, O.P.; Tivanova, O.V.; Turubarova, L.G.; Silnyagina, N.S.; Doronina, T.A.

    2006-01-01

    Full text: Earlier, during investigation of post-operating properties and structure of responsible units of fast neutron reactors there was shown /1, 2/ that depending on character of preliminary treatment of austenite stainless steel (austenization, cold deformation, mechanical and thermal treatment) radiation effects could be different. In /2/ one could observe heterogeneity at swelling of cold- worked hexahedral ducts along perimeter, in particular, the swelling of corners was less than plates'. At the same time after mechanical-thermal treatment the corners swell in 3-5 times of magnitude higher than plates. By the present there are several assumptions about nature of this phenomenon. One of them is a difference of deformation degree of material in corners and plates of the duct. It is known that /3/ external effects (including deformation) induce martensitic γ→α transformation in austenitic steels, due to which the structure and properties of steel are changed. In particular, paramagnetic FCC matrix reveals sites with ferromagnetic BCC structure. Steel heating, containing martensitic α-phase higher than ∼ 450-800 deg C, results in reverse γ→α transformation in material, which in its turn leads to formation of phase phase-hardened austenite. We can expect that only peculiarities of processes of direct and reverse martensitic transformations, which took place during preliminary austenitic steel treatment, will predetermine its behavior under irradiation. Taking into account the above mentioned there have been carried out complex material-scientific investigations of 12Cr18Ni10Ti and 08Cr16Ni11Mo3 steel samples cut off from different sites (both adjacent to corners and far from them) of hexagonal ducts of spent fuel assemblies of BN-350 reactor. There were used samples in the form of plates of different sizes: 5x10x2 mm - for metallographic investigations (microscope Neophot-2) and determination of microhardness (PMT-3); 2x20x0,3 mm - for mechanical

  9. Studies on plant dynamics of sodium-cooled fast breeder reactors - verification of a plant model

    International Nuclear Information System (INIS)

    Schubert, B.

    1988-01-01

    For the analysis of sodium-cooled FBR safety and dynamics theoretical models are used, which have to be verified. In this report the verification of the plant model SSC-L is conducted by the comparison of calculated data with measurements of the experimental reactors KNK II and RAPSODIE. For this the plant model is extended and adapted. In general only small differences between calculated and measured data are recognized. The results are used to improve and complete the plant model. The extensions of the plant model applicability are used for the calculation of a loss of heat sink transient with reactor scram, considering pipes as passive heat sinks. (orig./HP) With 69 figs., 10 tabs [de

  10. Impact of delays in plutonium use on the stationary growth rate of fast breeder reactor fuel

    International Nuclear Information System (INIS)

    Borg, R.C.; Ott, K.O.

    1977-07-01

    The hierarchy of the four growth rate expressions originally derived from an instantaneous reuse scheme is expanded to account for finite burnup in the core and blanket, β-decay of 241 Pu, core and blanket loading schemes, reuse delays due to reprocessing and fabricating fuel and external fuel cycle losses. The most general growth rate expression, obtained from the asymptotic slope of the accumulating fuel material in an expanding park of breeder reactors, is formally the same in both cases. Formulation of the growth rate based on the condensation of the detailed information of the equilibrium fuel cycle for a single reactor, is more complicated than without delays due to the composition difference between the average residing and excess discharge material. The third growth rate expression results from a slightly more complicated fuel-cycle eigenvalue problem than without delays. The last definition employs isotopic breeding worth factors obtained from the adjoint fuel cycle eigenvalue problem

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

    International Nuclear Information System (INIS)

    Adams, S.R.

    1985-10-01

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

  12. Seismic design principles for the German fast breeder reactor SNR 2

    International Nuclear Information System (INIS)

    Busch, K.A.; Peters, K.A.; Rosenhauer, W.

    1987-01-01

    The safety issue of an adequate and optimized external event protection is of course that unnecessary hardware precautions might promote internal disturbances or hamper their control. It has up to now not satisfactorily been realized that the only serious context for seismic impacts on a fast reactor is their attributed potential of overriding core disruptive accident prevention, see e.g. GRS 1982. General and exaggerated antiseismic design features not focussed upon this point may as well turn out to be non-negligible initators in the absence of seismic vibrations. Unexpected snubber difficulties requiring additional reactor scrams and decay heat removal phases may be named as a simple example. The presented seismic design principles reflect the progress made in the concerned fields of analysis and do serve on the other hand as guidelines for research and development efforts under work. (orig./GL)

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

    Energy Technology Data Exchange (ETDEWEB)

    Adams, S.R.

    1985-10-01

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

  14. Spatial kinetics studies in liquid-metal fast breeder reactor critical assemblies

    International Nuclear Information System (INIS)

    Brumback, S.B.; Goin, R.W.; Carpenter, S.G.

    1988-01-01

    Recent measurements in the zero-power physics reactor have been used to study the effect of spatial decoupling in fast reactor critical assemblies of various sizes and compositions. Flux distributions in these assemblies had varying degrees of sensitivity to perturbations. Decoupling was investigated using rod-drop, boron-oscillator, and noise-coherence techniques, which emphasized different times following perturbations. Equilibrium flux distributions were also measured for subcritical configurations with inserted control rods. For most assemblies, accurate reactivity measurements were obtained by analyzing the power history from a single detector using inverse kinetics methods, assuming an instantaneous efficiency change for the detector. The instantaneous efficiency change assumption broke down, however, in assemblies with zones in which normal plutonium fuel was replaced by /sup 235/U fuel or fuel with a high /sup 240/Pu content. Flux redistributions caused by perturbations in these cores took several minutes to evolve

  15. Characterization of the insoluble sludge from the dissolution of irradiated fast breeder reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Aihara, Haruka; Arai, Yoichi; Shibata, Atsuhiro; Nomura, K.; Takeuchi, M. [Japan Atomic Energy Agency - JAEA, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan)

    2016-07-01

    Insoluble sludge is generated in the reprocessing of spent fuel. The sludge obtained from the dissolution of irradiated fuel from the Joyo experimental fast reactor was analyzed to evaluate its chemical form. The sludge was collected by the filtration of the dissolved fuel solution, and then washed in nitric acid. The yields of the sludge weight were less than 1% of the total fuel weight. The chemical composition of the sludge was analyzed after decomposition by alkaline fusion. Molybdenum, technetium, ruthenium, rhodium, and palladium were found to be the main constituent elements of the sludge. X-ray diffraction patterns of the sludge were attributable to Mo{sub 4}Ru{sub 4}RhPd, regardless of the experimental conditions. The concentrations of molybdenum and zirconium in the dissolved fast reactor fuel solutions were low, indicating that zirconium molybdate hydrate (ZMH) is produced in negligible amounts in the process. (authors)

  16. Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

    International Nuclear Information System (INIS)

    Tachibana, Toshimichi

    1985-01-01

    The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

  17. On the chemical constitution of a molten oxide core of a fast breeder reactor

    International Nuclear Information System (INIS)

    Hodkin, D.J.; Potter, P.E.

    1980-01-01

    A knowledge of the chemical constitution of a molten oxide fast reactor core is of great importance in the assessment of heat transfer from a cooling molten pool of debris and in the selection of materials for the construction of sacrificial beds for core containment. In this paper we describe some thermodynamic assessments of the likely chemical constitution of a molten oxide core, and then support our assessments by experimental observations

  18. The role of fast breeder reactors in the future energy economy

    International Nuclear Information System (INIS)

    1978-09-01

    In this contribution, the reactor design and performance characteristics of a wide range of design concepts are documented. Since the technical feasibility of various design options and associated breeding performance depend strongly on the status of fuels development, primary focus is given to review of oxide, carbide and metal fuels in terms of current technology status, irradiation performance, associated key issues and main R and D requirements to resolve the issues

  19. The Bottom supported fast breeder reactor vessel - an alternative approach to seismic accommodation and reduced cost

    International Nuclear Information System (INIS)

    Nakagawa, H.; Golan, S.; Petrozelli, J.; Kumaoka, Y.; Kawamura, Y.

    1988-01-01

    Most FBR vessels are supported by hanging from their top portions. A disadvantage of such a top supported reactor vessel (TSRV) structural configuration is that it may generate high reactor core accelerations. This is due to the long path the seismic vibrations must travel from the basemat up through the building and then down through the RV block to the core. To compensate for this disadvantage, TSRV blocks are often strengthened beyond what is required for other considerations, such as pressure, to satisfy seismic response criteria, thus increasing weights and costs. In addition to long load paths, TSRVs also have common load paths. For example, in a TSRV (with the core supported from the bottom of the RV) the sodium and core loads both travel along the RV pressure boundary. Therefore, one of these loads will likely control the RV thickness leaving excess margin for the other loads. It is the premise of this paper that the revision of a large pool FBR from a TSRV configuration to a specific bottom supported reactor vessel (BSRV) configuration can resolve the above TSRV disadvantages related to load path length and diversity, thereby improving seismic performance and simultaneously reducing RV block costs by reducing weights. This paper demonstrates this premise by comparing a reference TSRV block with a specific BSRV block design

  20. Ultra-large size austenitic stainless steel forgings for fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    Tsukada, Hisashi; Suzuki, Komei; Sato, Ikuo; Miura, Ritsu.

    1988-01-01

    The large SUS 304 austenitic stainless steel forgings for the reactor vessel of the prototype FBR 'Monju' of 280 MWe output were successfully manufactured. The reactor vessel contains the heart of the reactor and sodium coolant at 530 deg C, and its inside diameter is about 7 m, and height is about 18 m. It is composed of 12 large forgings, that is, very thick flanges and shalls made by ring forging and an end plate made by disk forging and hot forming, using a special press machine. The manufacture of these large forgings utilized the results of the basic test on the material properties in high temperature environment and the effect that the manufacturing factors exert on the material properties and the results of the development of manufacturing techniques for superlarge forgings. The problems were the manufacturing techniques for the large ingots of 250 t class of high purity, the hot working techniques for stainless steel of fine grain size, the forging techniques for superlarge rings and disks, and the machining techniques of high precision for particularly large diameter, thin wall rings. The manufacture of these large stainless steel forgings is reported. (Kako, I.)

  1. Influence of design features on decommissioning of a large fast breeder reactor

    International Nuclear Information System (INIS)

    Fournie, J.-L.; Alary, C.; Maire, D.; Seroux, N. de; Peyrard, G.

    1990-01-01

    The evolution of FBR design in Europe shows that pool-type design will become the reference design for future FBR and the projected European Fast Reactor (EFR) is based on this concept. The identification of design features shows that the main contributors of the sodium and structures activity are the Co60 for gamma radiation source and low decay, Ni63, Nb94 and Ni59 for long time decay. So, the technical benefits of a Co content reduction are interesting for the high activated structures and for diagrid thimbles coating and we made proposals to lower Co content in steels or alloys and to substitute coatings. We identify measures which must facilitate both the sodium draining and the reactor block and internal cleaning: all which improve the gravity draining and the downing of the sodium flow make easier the penetration of cleaning products. The features, connected with the dismantling of the very activated internal structures, of the roof and of the lay-out, are mentioned. (author)

  2. Development of a transfer model for design of sodium purification systems for Fast Breeder Reactors

    International Nuclear Information System (INIS)

    Khatcheressian, N.

    2013-01-01

    Operating a Sodium Fast Reactor (SFR) in reliable and safe conditions requires to master the quality of the sodium fluid coolant, regarding oxygen and hydrogen impurities contents. A cold trap is a purification unit in SFR, designed for maintaining oxygen and hydrogen contents within acceptable limits. The purification of these impurities is based on crystallization of sodium hydride on cold walls and sodium oxide or hydride on wire mesh packing. Indeed, as oxygen and hydrogen solubilities are nearly nil at temperatures close to the sodium fusion point, i.e. 97.8 C, on line sodium purification can be performed by crystallization of sodium oxide and hydride from liquid sodium flows. However, the management of cold trap performances is necessary to prevent from unforeseen maintenance operations, which could induce shut-down of the reactor. It is thus essential to understand how a cold trap fills up with impurities crystallization in order to optimize the design of this system and to overcome any problems during nominal operation. The objective is to develop a design and simulation tool for cold traps able to predict the location and the amount of the impurities deposited. Crystallization model involve phenomena coupling in a porous medium with hydrodynamics, heat and mass transfer, distinguishing nucleation and growth phases for each impurity. It enables to understand how thermo hydraulic conditions and growing impurities interact on each other. This analysis will adapt operating and management conditions in order to optimize purification requirements. (author) [fr

  3. Significance of coast down time on safety and availability of a pool type fast breeder reactor

    International Nuclear Information System (INIS)

    Natesan, K.; Velusamy, K.; Selvaraj, P.; Chellapandi, P.

    2015-01-01

    Highlights: • Plant dynamics studies for quantifying the benefits of flow coast down time. • Establishment of minimum flow coast down time required for safety. • Assessment of influence of flow coast down on enhancing plant availability. • Synthesis of thermo mechanical benefits of flow coast down time on component design. - Abstract: Plant dynamic investigation towards establishing the influence of flow coast down time of primary and secondary sodium systems on safety and availability of plant has been carried out based on one dimensional analysis. From safety considerations, a minimum flow coast down time for primary sodium circuit is essential to be provided to limit the consequences of loss of flow event within allowable limits. Apart from safety benefits, large primary coast down time also improves plant availability by the elimination of reactor SCRAM during short term power failure events. Threshold values of SCRAM parameters also need optimization. By suitably selecting the threshold values for SCRAM parameters, significant reduction in the inertia of pumping systems can be derived to obtain desirable results on plant availability. With the optimization of threshold values and primary flow coast down behaviour equivalent to a halving time of 8 s, there is a possibility to eliminate reactor SCRAM during short term power failure events extending up to 0.75 s duration. Benefits of secondary flow halving on reducing transient thermal loading on components have also been investigated and mixed effects have been observed

  4. Utilization of OR method toward realization of better fast breeder reactor cycle

    International Nuclear Information System (INIS)

    Shiotani, Hiroki

    2008-01-01

    Fast Reactor Cycle Technology Development (FaCT) Project was now started aiming at commercialization of new nuclear power plants system. In parallel with development of component technology and technology demonstration by test, development of comprehensive evaluation method of the FBR cycle system is under way and scenario study, discounted cash flow (DCF) method, analytic hierarchy process (AHP), real option, supply chain management (SCM) and others are used. Since commercialized FBR cycle would request long-term and large-scale development contributed by so many participants, modeling of nuclear system and knowledge management are beneficial even for development of evaluation method and further utilization of OR technology is highly expected. Comprehensive evaluation methods now utilized or developing were overlooked from the standpoint of OR, 'Science of Better'. (T. Tanaka)

  5. Power excursion models applied to the study of secundary excursion in sodium cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Messainguiral, Christiane.

    1980-06-01

    An evaluation of the energy that a secondary power excursion could release has been sought throughout the present work. A parametric study was therefore made by means of a power excursion code in fast reactors. The work submitted is therefore made up of the three following parts: Part 1. - (a), the secondary excursion is situated in the generally envisaged programmes and (b) the role of the principal parameters is studied in the calculation effected by the nuclear excursion code that was available at the start of the study. Part 2. - the results obtained for the power excursion calculations made are presented, Part 3. - the insufficient modelling of the reactivity present during the secondary power excursion is deduced from the parametric study just made. A definition is made of the characteristics of a model adapted to the calculation of this hypothetical accident and a new model as worked out within the scope of this work is submitted [fr

  6. Creys-Melville: A case study on the acceptability of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lacroix, Andre; D' Heilly, Brigitte [NERSA-EdF (France)

    1993-07-01

    The Operator of CREYS-MALVILLE, being convinced that in a democratic country like France, Nuclear Energy in general and fast reactors in particular may only be developed in agreement with public opinion, will continue to develop its external Communication - without passion but with conviction. This paper has recalled the interests at stake associated with the resumption of power at CREYS-MALVILLE, the media and political contexts which have followed the construction and operation of this Plant, the administrative resources, and the communication equipment which have been developed over these years. All these means would be wasted and all communication efforts useless if they did not rely on the perseverance and competence of the 650 technicians and engineers guarantors of the Safety at CREYS-MALVILLE, for whom the coming Public Enquiry will be yet another occasion to explain, to France and its neighbours, the necessity to restart SUPERPHENIX.

  7. Creys-Melville: A case study on the acceptability of fast breeder reactors

    International Nuclear Information System (INIS)

    Lacroix, Andre; D'Heilly, Brigitte

    1993-01-01

    The Operator of CREYS-MALVILLE, being convinced that in a democratic country like France, Nuclear Energy in general and fast reactors in particular may only be developed in agreement with public opinion, will continue to develop its external Communication - without passion but with conviction. This paper has recalled the interests at stake associated with the resumption of power at CREYS-MALVILLE, the media and political contexts which have followed the construction and operation of this Plant, the administrative resources, and the communication equipment which have been developed over these years. All these means would be wasted and all communication efforts useless if they did not rely on the perseverance and competence of the 650 technicians and engineers guarantors of the Safety at CREYS-MALVILLE, for whom the coming Public Enquiry will be yet another occasion to explain, to France and its neighbours, the necessity to restart SUPERPHENIX

  8. Method of preventing sodium from flowing when pipes of a fast breeder reactor are injured

    International Nuclear Information System (INIS)

    Nakai, Yasushi; Yamagishi, Yoshiaki; Koga, Tomonari.

    1975-01-01

    Object: To inject high pressure sodium into an inlet nozzle portion when fluid pressure in the inlet nozzle portion of a core cooling pipe on the inlet side is in an abnormal condition, to thereby quickly and positively prevent the flow of sodium in a high pressure chamber in a reactor vessel, when pipes are injured. Structure: When the core cooling pipe on the inlet side is injured and as a consequence the pressure gage detects an abnormal condition of fluid pressure in the inlet nozzle, the valve is opened to allow high pressure sodium to inject into the inlet nozzle through a high pressure sodium supply pipe, thereby blocking a back-flow of sodium in the high pressure chamber into the core cooling pipe. (Kamimura, M.)

  9. Burnup effects on criticality, breeding and safety of 1,000 MWe gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Ohta, Fumio

    1977-12-01

    Burnup characteristics of 1,000 MWe, PuO 2 - UO 2 fuelled helium-cooled fast breeder reactor have been studied concerning criticality, breeding and safety. A 26-energy group cross-section set produced from ENDF/B-3 was used. Criticality and breeding were studied with two-dimensional burnup code APOLLO and 4-energy group cross-section set generated by collapsing the mentioned cross-section set. Safety aspects such as Doppler reactivity effect, coolant-depressurisation and steam-ingression reactivity effect were studied with multi-dimensional diffusion theory code CITATION and perturbation theory code PERKY, as well as the 26-energy group cross-section set. The following were revealed: (1) The reactivity swing over a year's irradiation is merely 1.5% ΔK/K. This small swing may permit relatively long fuel dwelling in GCFR and , thus, the frequency of outages for refuelling can be minimised. (2) The surplus fissile plutonium over a year's irradiation is about 360 Kg, and the system doubling time is about 9 years. The GCFR studied has excellent breeding, compared with those in PuO 2 -UO 2 fuelled LMFBR and other GCFRs. (3) The coolant-depressurisation reactivity effect becomes more positive with burnup. This is not so serious as the sodium-void reactivity effect of LMFBR. (4) In the start-up core, the steam-ingression reactivity effect due to steam ingression to the core and blanket from the secondary coolant system becomes positive at certain steam density (0.02gr/cc) and this positive effect increases with steam density. With advance of burnup, however, the effect becomes negative, this increasing with steam density. After all, the steam ingression is no hazard in operation of GCFR since the reactivity effect is negative in the equilibrium state. (auth.)

  10. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    Hueper, R.

    1989-01-01

    The results of activities carried out by the DeBeNe partners in 1988 have been compiled in this report. The 1989 KNK II experimental program will focus on the management of fuel element failures. This includes in particular post-irradiation examinations in the hot cells and the THIBO experiments (thermally induced fuel rod oscillation). For this program, nine permits were received in 1988 for the installation and operation of test systems, including a new facility for localizing failed fuel elements. Increasing the purity of sodium is the purpose of a cesium trap recently installed, and of modifications to an oxygen probe operated for test purposes. The SNR 300 project is being continued within the framework of the ''holding phase''. The objective of this phase of operation is to keep the reactor plant in the loading and operational states and execute planning within the licensing procedure necessary to obtain the next partial permit. R and D work was concentrated on fuel and materials development, safety, physics, and components development. Materials for fuel, blanket, and absorber elements were studied and further developed with a view to very high burnup. In the sector of physics, the engineering design and the nuclear design of large breeders call for a reduction of the margins of uncertainty in reliable predictions of the major reactor parameters. The development of the joint European cell code, ECCO (European Cell Code) has progressed far enough in the meantime to allow the criticality parameter, k eff , of a cell arrangement to be calculated with ECCO for the first time at UKAEA Winfrith, the central agency for carrying out the development work. One of the major aspects covered in safety studies is the complex of fuel rod failures, loss of coolant flow, and power transients as possible causes of accidents. Studies conducted into the management of credible accident consequences were concentrated, among other topics, on the behaviour of aerosols, sodium fires

  11. 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

  12. 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

  13. Advanced fuel for fast breeder reactors: Fabrication and properties and their optimization

    International Nuclear Information System (INIS)

    1988-06-01

    The present design for FBR fuel rods includes usually MOX fuel pellets cladded into stainless steel tubes, together with UO 2 axial blanket and stainless steel hexagonal wrappers. Mixed carbide, nitride and metallic fuels have been tested as alternative fuels in test reactors. Among others, the objectives to develop these alternative fuels are to gain a high breeding ratio, short doubling time and high linear ratings. Fuel rod and assembly designers are now concentrating on finding the combination of optimized fuel, cladding and wrapper materials which could result in improvement of fuel operational reliability under high burnups and load-follow mode of operation. The purpose of the meeting was to review the experience of advanced FBR fuel fabrication technology, its properties before, under and after irradiation, peculiarities of the back-end of the nuclear fuel cycle, and to outline future trends. As a result of the panel discussion, the recommendations on future Agency activities in the area of advanced FBR fuels were developed. A separate abstract was prepared for each of the 10 presentations of this meeting. Refs, figs and tabs

  14. Status of fast breeder reactor development in the United States of America - April 1984

    International Nuclear Information System (INIS)

    Horton, K.E.

    1984-01-01

    The Breeder Technology program continues to produce viable information on fuel performance, nuclear systems technology, and power conversion technology. The unique testing capabilities design into the FFTF have resulted in well-validated materials and fuels irradiation information that has confirmed and extended previous data bases. Current directions for the research and development program are to improve the technology for power conversion systems, components, instrumentation, and materials technology to the point where cost reduction and reliability potentials are realized. Operation of the breeder test facility complex at the Hanford Engineering Development Laboratory (HEDL), the Energy Technology Engineering Center (ETEC), and the Argonne National Laboratory (ANL) continues to provide the experience base and test capability for the breeder R and D effort. International cooperation will be even more important in the future than in the past for several reasons. Significant new investments still have to be made in breeder R and D to improve designs, achieve economic competitiveness and to develop practical breeder fuel cycle capabilities. Progress can be accelerated, redundancies avoided, and economics achieved if nations coordinate their programs, and where possible, divide up the work. In addition, there is clear mutual benefit in encouraging the countries involved in breeder development to harmonize standards and regulations related to safety. It is also important that the advanced nations work together closely in assuring that adequate international safeguards, export controls, and national physical security measures keep pace with breeder reactor and fuel cycle developments

  15. Sodium/water reactions in steam generators of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Hori, M.

    1980-01-01

    The status of the research and development on sodium/water reactions resulting from the leakage of water into sodium in LMFBR steam generators is reviewed. The importance of sodium/water reaction phenomena in the design and operation of steam generators is discussed. The effects of sodium/water reactions are evaluated and methods of protection against these phenomena are surveyed. The products of chemical reactions between sodium and water under steam generator conditions are H 2 , NaOH, Na 2 O and NaH. Together with the temperature rise due to the associated exothermic reaction, these reaction products cause effects such as self-wastage, single- and multi-target wastage, and rapid pressure increase, depending on the size of the leak hole or the magnitude of leak rate. As for the wastage phenomena of small leaks, the effects of various factors have been studied and experimental correlations, as well as some theoretical work, have been performed. To investigate the pressure phenomena of a large leak, large-scale tests have been conducted by various organizations, and the computer codes to analyse these phenomena have been developed and verified by experiments. In the design of steam generators, an initial failure up to a hypothetical double-ended guillotine rupture of a single heat transfer tube is widely used as the design basis leak. Protection systems for LMFBR plants consist of leak detection devices, leak termination devices, and reaction pressure relief devices. From analyses based on research and development activities, as well as from experience with leaks in steam generator test loops and reactor plants, it has been confirmed that protection systems can satisfactorily be designed to accommodate leak incidents in LMFBR plants. (author)

  16. Development of polymeric applications for sodium cooled Fast Breeder Reactors: Chronicles of inception, progress and achievements

    International Nuclear Information System (INIS)

    Sinha, N.K.; Raj, Baldev

    2010-01-01

    The collaborative programme on development of important polymeric applications of Indian FBRs is chronicled from the days of motivation to its present state. Failure of inflatable seals of FBTR RPs (1985) and adoption of all-elastomer sealing concept for PFBR RPs (early 1990s), coupled with the unique characteristics of elastomeric materials, led to inception of the programme at IGCAR (1998) which involved DMSRDE as the first partner (1999). The planned initiative, which eventually involved more than 15 other Indian agencies, resulted in complete development of FKM backup seals for PFBR RPs which has been installed in reactor recently. Coated FKM and EPDM inflatable seals for PFBR and FBTR RPs have been developed, produced and evaluated up to ∼2 m diameter. Development methodologies for other critical polymeric applications of PFBR, FBTR and FCF have been formulated. Accomplishments and novelties of the development include EPDM and FKM compounds and designs for inflatable and backup seals, a common FEA procedure for elastomeric ring seals, PECVD based Teflon-like coating technology up to 7 m seal diameter, seal production process by cold feed extrusion and continuous cure, a robust quality control framework and the new facilities developed to support the programme. Future developments are focused on delivery of validated inflatable seals, life assessment and development of new elastomeric compounds which include silicone rubber and perfluoroelastomer, PECVD based coating on stainless steel and development of adhesionless joining of FKM. The achievements and future research will standardize the design and development of the elastomeric seals of Indian FBRs, PHWRs and AHWR based on a few well-characterized compounds, a common FEA method and PECVD based coating technology which can result in a universal design code.

  17. Seismic sloshing experiments of large pool-type fast breeder reactors

    International Nuclear Information System (INIS)

    Sakurai, A.; Masuko, Y.; Kurihara, C.; Ishihama, K.; Yashiro, T.; Rodwell, E.

    1989-01-01

    This paper presents the results of seismic sloshing experiments performed on large pool-type LMFBR vessels. Two types of tests were performed. The first type of test was designed to understand the basis phenomena of sloshing (limited to linear sloshing only) and evaluate the effects of the deck-mounted components (i.e., IHXs, pumps, and UIS) on sloshing wave heights using a 1/10-scale model (diameter 2.23 m x H 1.03 m) of the LSPB 1340 MWe pool plant. The second type of test was designed to evaluate the structural integrity of the thermal baffles of the roof-deck to withstand sloshing impulsive pressures (focused on nonlinear sloshing), using a two-dimensional 1/3-scale model (L 8 m x W 3 m x H 2.6 m) of a typical 1000 MWe pool plant. The results of the linear sloshing tests have shown that: 1. the vessel wall stiffness has no effect on the sloshing natural frequency; 2. sloshing wave heights are lowered by 30% to 50% in the presence of the deck-mounted components; and 3. damping factors of sloshing are not influenced by the wall stiffness while they are increased by the presence of the deck-mounted components. The results of the nonlinear sloshing tests are that: 1. the maximum impulsive pressure occurs when the first effective wave strikes at the roof-deck, and thereafter the impulsive pressure decreases irrespective of the impact velocity of the fluid; 2. the first effective wave refers to the case in which the height of the fluid free surface becomes nearly twice the height of the cover gas space; and 3. the structural integrity of the thermal baffles for the roof-deck against the sloshing load was confirmed. In addition to these results, two sloshing-caused problems were identified. The first one is the spillover of hot sodium into the gas-dam type thermal insulator. The second one is cover-gas entrainment into sodium which might lead to a transient overpower (TOP) incident because of the presence of gas bubbles in the reactor core. (orig./HP)

  18. A study of sodium-cooled fast breeder reactor with thorium blanket for supply of U-233 to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Yoshida, H.; Nishimura, H.; Osugi, T.

    1978-08-01

    Symbiotic energy system between fast breeder reactor and thermal reactor would have a potential merit for nuclear proliferation problem. And when using HTGR as the thermal reactor in the system, the energy system appears to be promising as an energy system self-sufficient in fuels, which can generate both electricity and high temperature process heat. In the system the fast breeder reactor has to supply sufficient amount of fissile plutonium to keep the reactor going, and also produce U-233 necessary to the associated U-233 fuelled process heat production HTGR. Three types of LMFBR concepts with thorium blanket, conventional homogeneous core LMFBR, and axial and radial parfait heterogeneous core LMFBRs, have been investigated to find out suitable configurations of LMFBR for supply of U-233 to the HTGR with relatively high conversion ratio of 0.85, in the symbiotic energy system between LMFBR and HTGR. The investigation on LMFBR has been made on fuel sufficiency of the system, inherent safety such as sodium-void and Doppler coefficients, and fuel cycle cost. The followings were revealed; (1) Conventional homogeneous core LMFBR with thorium radial blanket well satisfies the condition of fuel sufficiency, if adequate radial blanket thickness is chosen. However, the sodium-void coefficient and fuel cycle cost are inferior to the other concepts. (2) Axial parfait heterogeneous core LMFBR can be regarded as one of the best LMFBR concepts installed in the symbiotic energy system, from the viewpoints of fuel sufficiency, inherent safety and fuel cycle cost. However, further investigations should be needed on reliability and operationability of the concept. (3) Radial parfait heterogeneous core LMFBR seems inadequate as the LMFBR in the system, because the configurations based on this concept does not satisfy plutonium and U-233 breedings, simultaneously. This LMFBR concept, however, has excellent breeding performance in the internal radial blanket. So further

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

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

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

  20. Development of safety evaluation methods and analysis codes applied to the safety regulations for the design and construction stage of fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purposes of this study are to develop the safety evaluation methods and analysis codes needed in the design and construction stage of fast breeder reactor (FBR). In JFY 2012, the following results are obtained. As for the development of safety evaluation methods needed in the safety examination conducted for the reactor establishment permission, development of the analysis codes, such as core damage analysis code, were carried out following the planned schedule. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied. (author)

  1. Project fast breeder (PSB)

    International Nuclear Information System (INIS)

    1978-01-01

    The annual report of the fast breeder project (PSB) contains contributions of the participating institutes on the four subjects: 1) Development of oxidic fuel rods and materials for the SNR line, 2) Physics and safety investigations for the SNR line, 3) Carbidic fuel elements, and 4) Back-up solution with gaseous coolant. (HK) [de

  2. Impact of partitioning and transmutation on high-level waste disposal for the fast breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

    2010-01-01

    The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large

  3. Experimental simulation of fragmentation and stratification of core debris on the core catcher of a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pillai, Dipin S.; Vignesh, R. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Sudha, A. Jasmin, E-mail: jasmin@igcar.gov.in [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Pushpavanam, S.; Sundararajan, T. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Nashine, B.K.; Selvaraj, P. [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2016-05-15

    Highlights: • Fragmentation of two simultaneous metals jets in a bulk coolant analysed. • Particle size from experiments compared with theoretical analysis. • Jet breakup modes explained using dimensionless numbers. • Settling aspects of aluminium and lead debris on collector plate studied. • Results analysed in light of core debris settling on core catcher in a FBR. - Abstract: The complex and coupled phenomena of two simultaneous molten metal jets fragmenting inside a quiescent liquid pool and settling on a collector plate are experimentally analysed in the context of safety analysis of a fast breeder reactor (FBR) in the post accident heat removal phase. Following a hypothetical core melt down accident in a FBR, a major portion of molten nuclear fuel and clad/structural material which are collectively termed as ‘corium’ undergoes fragmentation in the bulk coolant sodium in the lower plenum of the reactor main vessel and settles on the core catcher plate. The coolability of this decay heat generating debris bed is dependent on the particle size distribution and its layering i.e., stratification. Experiments have been conducted with two immiscible molten metals of different densities poured inside a coolant medium to understand their fragmentation behaviour and to assess the possibility of formation of a stratified debris bed. Molten aluminium and lead have been used as simulants in place of molten stainless steel and nuclear fuel to facilitate easy handling. This paper summarizes the major findings from these experiments. The fragmentation of the two molten metals are explained in the light of relevant dimensionless numbers such as Reynolds number and Weber Number. The mass median diameter of the fragmented debris is predicted from nonlinear stability analysis of slender jets for lead jet and using Rayleigh's classical theory of jet breakup for aluminium jet. The agreement of the predicted values with the experimental results is good. These

  4. Fast breeder reactors

    International Nuclear Information System (INIS)

    Judd, A.M.

    1981-01-01

    The subject is covered in chapters: introduction; physics (introduction; calculation methods; flux distribution; breeding; control rods; reactivity coefficients; shielding); fuel (introduction; fuel temperatures; fuel design and manufacture; behaviour during irradiation); engineering (introduction; core heat transfer; structural materials; core structure; dynamics and instrumentation); coolant circuits and steam plant (introduction; primary circuit; steam plant; control systems); safety (introduction; inherent safety features; protective systems; hypothetical accidents; risks). (U.K.)

  5. Superphenix: Is the fast breeder dream over -- or over yonder?

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    A detailed history of France's Superphenix commercial fast breeder reactor project is presented. Important project milestones are discussed from the project's conception in 1971 to its current status. Recommendations of the Castaing Commission on the project and future plans for use of the reactor are outlined. In addition, world wide fast breeder projects are listed and discussed

  6. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. Pt. II. R and D necessities and development across the world. A review

    International Nuclear Information System (INIS)

    Sinha, Nilay K.; Raj, Baldev

    2013-01-01

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative study of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton registered GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or readymade supply. R and D necessities for inflatable seals and their development across the world are given closer look in Part II of the review in continuation of Part I. (orig.)

  7. Fast breeder fuel element development

    International Nuclear Information System (INIS)

    Marth, W.; Muehling, G.

    1983-08-01

    This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

  8. Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

    International Nuclear Information System (INIS)

    McCulloch, R.W.; Post, D.W.; Lovell, R.T.; Snyder, S.D.

    1981-04-01

    Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators

  9. Effect of fluid-to-structure heat transfer on the structural damage potential to a liquid-metal fast breeder reactor

    International Nuclear Information System (INIS)

    Hakim, S.J.; Abramson, P.B.

    1979-01-01

    Deterministic calculations simulating a hypothetical accident in a liquid-metal fast breeder reactor that leads to a hydrodynamic disassembly of the core have been carried out to estimate the system's damage potential due to the vapor-pressure-driven expansion of molten core material and its dependency on the heat transfer to the remaining structure. These calculations ignored the effect on the work potential of sodium left in the core during the disassembly. Results indicate that steel cladding in the upper axial blankets and fission gas plenum acts as a thermodynamic energy sink that could reduce the total thermodynamic work energy by between one and two orders of magnitude, provided little or no sodium is present in the core at the time of interaction. These results have been found to be insensitive to the rate of heat transferred from the molten fuel to the molten steel that comprises the molten core material

  10. Recovery of a broken inspection lance from the reactor vessel of the German sodium cooled fast breeder reactor SNR 300

    International Nuclear Information System (INIS)

    Menck, J.W.; Hoeft, E.; Kirchner, G.

    1988-01-01

    An inspection lance for flow and vibration measurements was installed into the SNR-300 rotating plug. Centering and guiding of the lower end of the lance was effected in the central grid plate insert. The lance was torn off due to handling problems. The task consisted in recovering all defective parts from the reactor vessel and re-establishing the intact initial state. (author)

  11. Criticality safety issues in the disposition of BN-350 spent fuel

    International Nuclear Information System (INIS)

    Schaefer, R. W.; Klann, R. T.; Koltyshev, S. M.; Krechetov, S.

    2000-01-01

    A criticality safety analysis has been performed as part of the BN-350 spent fuel disposition project being conducted jointly by the DOE and Kazakhstan. The Kazakhstan regulations are reasonably consistent with those of the DOE. The high enrichment and severe undermoderation of this fast reactor fuel has significant criticality safety consequences. A detailed modeling approach was used that showed some configurations to be safe that otherwise would be rejected. Reasonable requirements for design and operations were needed, and with them, all operations were found to be safe

  12. Overview of pool hydraulic design of Indian prototype fast breeder ...

    Indian Academy of Sciences (India)

    Flow sheet of prototype fast breeder reactor. ... over, the main vessel that houses radioactive primary sodium is free of any ..... with superficial velocity components in porous media. ..... The attenuation within thermal boundary layer was found.

  13. Scoping calculations for design and analysis of large reactor vessels for liquid-metal fast breeder reactor (LMFBR) plants

    International Nuclear Information System (INIS)

    Fiala, C.; Kulak, R.F.; Ma, D.C.; Pan, Y.C.; Seidensticker, R.W.; Wang, C.Y.; Zeuch, W.R.

    1982-01-01

    Reactor vessels for commercial-sized LMFBR plants are quite large - ranging 40 to 70 ft in diameter and 50 to 70 ft in overall depth. These stainless steel vessels contain liquid sodium at relatively low pressures, but at high temperatures. The resulting thin-walled vessels present the structural designer and analyst with special problems, particularly in providing a balanced design to accommodate seismic loads, design basis accident loads, and thermal loadings. A comprehensive set of scoping calculations - though preliminary in detail and depth of design - provides substantial guidance to the vessel designer for subsequent design iterations. Emphasis is placed on the analysis of the large-diameter top closure of the vessel - the deck structure

  14. Risk-oriented analysis of the German prototype fast breeder reactor SNR-300: off-site accident consequence model and results of the study

    International Nuclear Information System (INIS)

    Bayer, A.; Ehrhardt, J.

    1984-01-01

    Accident off-site consequence calculations and risk assessments performed for the ''risk oriented analysis'' of the German prototype fast breeder reactor SNR-300 were performed with a modified version of the off-site accident consequence model UFOMOD. The modifications mainly relate to the deposition and resuspension processes, the ingestion model, and the dose factors. Consequence calculations at the site of Kalkar on the Rhine River were performed for 115 weather sequences in 36 wind directions. They were based on seven release categories evaluated for the SNR-300 with two different fueling strategies: plutonium from Magnox reactors only and plutonium from light water reactors and Magnox reactors. In parallel, the corresponding frequencies of occurrence are determined. The following results are generated: 1. complementary cumulative frequency distribution functions for collective fatalities and collective doses 2. expected values of the collective fatalities and collective doses as well as distance-dependent expected values of individual fatality 3. contributions of the different exposure pathways to fatalities with respect to the various organs. For comparison with the risk of a PWR-1300, calculations for the PWR-1300 of the ''German Risk Study'' were repeated with the same modified consequence model. Comparison shows that smaller risks result for the SNR-300. However, the confidence interval bandwidths obtained for the frequencies of the release categories for the SNR-300 are larger than those of the PWR-1300

  15. An investigation into the use of a fast breeder reactor to incinerate actinide waste from the U.K. nuclear power programme

    International Nuclear Information System (INIS)

    Harte, G.A.; Clarke, R.H.

    1976-12-01

    Preliminary investigations are described into the effect of actinide removal and incineration (i.e. re-irradiation in a high neutron flux) on the long term hazards of reprocessing wastes from a Magnox reactor and from a notional CFR operated on plutonium from a Magnox reactor and also from an AGR. The actinides neptunium, americium and curium extracted from these wastes would be amenable to incineration in a fast breeder reactor core with fluxes of the order of 10 16 ncm -2 s -1 . An incineration period of 35 years was found to achieve a reduction in hazard, for most cooling periods up to one million years. The reductions were 2 to 3 orders of magnitude for the Magnox wastes and 4 orders of magnitude for the waste from CFR fuelled with AGR-plutonium. The large reduction factors do not apply to the hazards of reprocessing wastes as a whole, however, since one actinide separation has been achieved the residual quantities of uranium and plutonium in the waste stream, assumed here to be 0.1% of the original fuel inventory of these elements, become dominant in determining the hazard. Shorter incineration periods may serve to reduce the hazards of the extracted transuranics below the levels determined by these residues. A discussion of the hazards associated with reprocessing wastes as compared to those associated with uranium mill tailings attempts to place the nuclear waste disposal problem in perspective. (author)

  16. Experimental study of the large-scale axially heterogeneous liquid-metal fast breeder reactor at the fast critical assembly: Power distribution measurements and their analyses

    International Nuclear Information System (INIS)

    Iijima, S.; Obu, M.; Hayase, T.; Ohno, A.; Nemoto, T.; Okajima, S.

    1988-01-01

    Power distributions of the large-scale axially heterogeneous liquid-metal fast breeder reactor were studied by using the experiment results of fast critical assemblies XI, XII, and XIII and the results of their analyses. The power distributions were examined by the gamma-scanning method and fission rate measurements using /sup 239/Pu and /sup 238/U fission counters and the foil irradiation method. In addition to the measurements in the reference core, the power distributions were measured in the core with a control rod inserted and in a modified core where the shape of the internal blanket was determined by the radial boundary. The calculation was made by using JENDL-2 and the Japan Atomic Energy Research Institute's standard calculation system for fast reactor neutronics. The power flattening trend, caused by the decrease of the fast neutron flux, was observed in the axial and radial power distributions. The effect of the radial boundary shape of the internal blanket on the power distribution was determined in the core. The thickness of the internal blanket was reduced at its radial boundary. The influence of the internal blanket was observed in the power distributions in the core with a control rod inserted. The calculation predicted the neutron spectrum harder in the internal blanket. In the radial distributions of /sup 239/Pu fission rates, the space dependency of the calculated-to-experiment values was found at the active core close to the internal blanket

  17. Fast breeder reactor program. Hearings before the Joint Economic Committee, Congress of the United States, Ninety-Fourth Congress, First Session

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The economics of the liquid metal fast breeder reactor (LMFBR) was the subject of hearings of the Joint Economic Committee, chaired by Sen. Hubert Humphrey. FY '76 funding for the breeder program is $450 million, the largest single item of the Federal energy program. Elmer B. Staats, U.S. Comptroller General, testified on the rising costs of demonstration facilities and pointed out that Federal agencies are required to make all estimates of costs and benefits in constant dollars rather than projecting for inflation. Staats recommended a joint ERDA-Congressional study of the possible use of foreign breeder technology. Sheldon Meyers of the Environmental Protection Agency, while not opposing the breeder program, recommended a delay to resolve three problem areas: (1) base energy demand projections; (2) timing of the commercial introduction of the LMFBR; and (3) uncertainties over possible benefits from the LMFBR program. Theodore B. Taylor, International Research and Technology Corp., discussed the costs and security safeguards of the LMFBR, which produces more spent fuel than the light water reactor. Other witnesses included Robert Seamans and officials from ERDA, Ralph Nader, and speakers from private study groups

  18. Control-rod, pressure and flow-induced accident and transient analysis of a direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Kitoh, Kazuaki; Koshizuka, Seiichi; Oka, Yoshiaki

    1996-01-01

    The features of the direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor (SCFBR) are high thermal efficiency and simple reactor system. The safety principle is basically the same as that of an LWR since it is a water-cooled reactor. Maintaining the core flow is the basic safety requirement of the reactor, since its coolant system is the one through type. The transient behaviors at control rod, pressure and flow-induced abnormalities are analyzed and presented in this paper. The results of flow-induced transients of SCFBR were reported at ICONE-3, though pressure change was neglected. The change of fuel temperature distribution is also considered for the analysis of the rapid reactivity-induced transients such as control rod withdrawal. Total loss of flow and pump seizure are analyzed as the accidents. Loss of load, control rod withdrawal from the normal operation, loss of feedwater heating, inadvertent start of an auxiliary feedwater pump, partial loss of coolant flow and loss of external power are analyzed as the transients. The behavior of the flow-induced transients is not so much different from the analyses assuming constant pressure. Fly wheels should be equipped with the feedwater pumps to prolong the coast-down time more than 10s and to cope with the total loss of flow accident. The coolant density coefficient of the SCFBR is less than one tenth of a BWR in which the recirculation flow is used for the power control. The over pressurization transients at the loss of load is not so severe as that of a BWR. The power reaches 120%. The minimum deterioration heat flux ratio (MDHFR) and the maximum pressure are sufficiently lower than the criteria; MDHFR above 1.0 and pressure ratio below 1.10 of 27.5 MPa, maximum pressure for operation. Among the reactivity abnormalities, the control rod withdrawal transient from the normal operation is analyzed

  19. CAMDYN: a new model to describe the axial motion of molten fuel inside the pin of a fast breeder reactor during accident conditions

    International Nuclear Information System (INIS)

    Peter, G.

    1991-01-01

    The new in-pin fuel motion model CAMDYN (Cavity Material Dynamics) describes the axial motion of both partially and fully molten fuel inside the pin of a fast breeder reactor during accident conditions. The motion of the two types of molten fuel and the imbedded fission gas bubbles is treated both before and after cladding failure. The basic modelling approach consists of the treatment of two one-dimensional flows which are coupled by interaction terms. Each of these flows is treated compressively and with axially variable flow cross sections. The mass and energy equations of both fields are solved explicitly using upwind differencing on a fixed Eulerian grid. The two momentum equations are solved simultaneously, using the convective momentum fluxes of the previous timestep. Both partially and fully molten fuel can move axially into a central hole extending to the plenum in the case of certain hollow pellet designs. The fuel temperature calculation includes the determination of a radial temperature profile. A simple conduction freezing model is included. After cladding failure, ejection into the coolant channel is modeled

  20. Comparative analysis of quality assurance systems which effectively control, review and verify the quality of components manufactured for liquid metal cooled fast breeder reactors within the EEC

    International Nuclear Information System (INIS)

    Benn, L.A.

    1985-01-01

    Comparative analyses are made of Quality Assurance Systems, by techniques and the methodology used, for the manufacture of component parts for the Liquid Metal Cooled Fast Breeder Reactor (LMFBR) within the EEC. Two differing alternative systems are presented in the analysis. First, a tabulated analytical treatment which analyses 14 codes and standards relating to Quality Assurance which can be applied to LMFBR's. The comparison equates equivalent clauses between codes and standards followed by an analysis of individual clauses in tabular form, the International Standard ISO 6215. A statistical summary and recommendations conclude this analysis. The second alternative system used in the comparison is a descriptive analytical method applied to 9 selected codes and standards relating to Quality Assurance based on the 13 criteria of the International IAEA Code of Practice no. 50 C.QA entitled ''Quality Assurance for Safety in Nuclear Power Plants''. An investigation is then made of the state of the art on the subject of classification of component parts bearing generally on Quality Assurance. The method of classification is segregated into General, Safety and Inspection categories. A summary of destructive and non destructive controls that may be applied during the manufacture of LMFBR components is given, together with tests that may be applied to selected components, namely Primary Tank, Secondary Sodium Pump and the Primary Cold Trap allocated to Safety Classes, 1, 2 and 3 respectively. The report concludes with a summary of typical records produced at the delivery of a component

  1. The radiological consequences of notional accidental releases of radioactivity from fast breeder reactors: sensitivity to the dose-effect relationships adopted for early biological effects

    International Nuclear Information System (INIS)

    Kelly, G.N.; Simmonds, J.R.; Smith, H.; Stather, J.W.

    1979-07-01

    This study considered the sensitivity to the dose-response relationships adopted for the estimation of early biological effects from notional accidental releases of radioactivity from fast breeder reactors. Two distinct aspects were considered: the sensitivity of the predicted consequences to variation in the dose-mortality relationships for irradiation of the bone marrow and the lung; and the influence of simple supportive medical treatment in reducing the incidence of early deaths in the exposed population. The numbers of early effects estimated in the initial study were relatively insensitive to variation in the dose-mortality relationships within the bounds proposed. The few exceptions concerned releases of particular nuclide composition, and the variation in the predicted consequences could be around an order of magnitude; the absolute numbers of effects however were in general small when the sensitivity was most pronounced. The reduction in the incidence of early deaths when using simple supportive treatment varied markedly with the nuclide composition of the release. Areas of uncertainty were identified where further research and investigation might most profitably be directed with a view to improving the reliability of the dose-effect relationships adopted and hence of the predicted consequences of the release considered. (author)

  2. The influence of the physico-chemical form of the aerosol on the radiological consequences of notional accidental releases of radioactivity from a fast breeder reactor

    International Nuclear Information System (INIS)

    Kelly, G.N.; Jones, J.A.; Simmonds, J.R.

    1979-01-01

    The radiological consequences of a wide range of notional accidental releases from a 1300 MW(e) LMFBR (Liquid Metal-cooled Fast Breeder Reactor) were assessed in a study published by the National Radiological Protection Board (NRPB) in 1977. In that study representative values were in general adopted for each of the important parameters while recognising that in reality they could vary considerably. The present study is concerned with the sensitivity of the predicted consequences to the physico-chemical form of the released aerosol. Of particular interest is the importance of a mixed sodium-transuranium element aerosol which may be formed in accidental releases of activity from sodium cooled FBRs. Two significant findings emerge from the study. First the predicted consequences in general are relatively insensitive to the range of physico-chemical forms analysed. For generic assessments therefore it is sufficient to assume the properties of the aerosol adopted in the initial study (1 μm AMAD and each element in the oxide form); the exception concerns the estimation of the incidence of early morbidity, and to a lesser extent early mortality, but only for a limited range of release composition. The second finding is that the radiological consequences are not, contrary to what might have been expected, significantly increased for the release of a mixed sodium-element aerosol

  3. Evolution in the design and development of the in-service inspection device for the Indian 500 MWe Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Singh, Ashutosh Pratap; Rajagopalan, C.; Rakesh, V.; Rajendran, S.; Venugopal, S.; Kasiviswanathan, K.V.; Jayakumar, T.

    2011-01-01

    Highlights: → Conceptual study on the configuration of an ISI device for FBR interspace environment has been carried out. → Prototyping of the concept has been experimentally validated in a mock up. → High temperature version of the ISI device has been made and tested in mock-up. Further experimentation is underway. → Simulation of different configurations of the device has been carried out with respect to reduced gap between main vessel and safety vessel for future FBRs. → Studies on wheel lining for the device have been carried out at 150 o C for better traction and payload capability. - Abstract: In-service inspection (ISI) plays a major role in monitoring the condition of nuclear power plant structures and components. Based on the information gathered during inspection and the studies carried out, it is possible to assess the extent of damage and take corrective measures to keep effects of ageing under control. In nuclear power plants comprehensive ISI is dictated by issues of increased safety to personnel and equipment, and efficiently enhances the plant life. A special emphasis has been laid on the development of robotic devices for the ISI of the indigenous Indian 500 MWe Prototype Fast Breeder Reactor (FBR) components. This paper traces the experiments and simulations in the key developments of a robotic device, for the ISI of main vessel and safety vessel of FBRs, carried out at Indira Gandhi Centre for Atomic Research, India.

  4. CALIPSO - a computer code for the calculation of fluiddynamics, thermohydraulics and changes of geometry in failing fuel elements of a fast breeder reactor

    International Nuclear Information System (INIS)

    Kedziur, F.

    1982-07-01

    The computer code CALIPSO was developed for the calculation of a hypothetical accident in an LMFBR (Liquid Metal Fast Breeder Reactor), where the failure of fuel pins is assumed. It calculates two-dimensionally the thermodynamics, fluiddynamics and changes in geometry of a single fuel pin and its coolant channel in a time period between failure of the pin and a state, at which the geometry is nearly destroyed. The determination of temperature profiles in the fuel pin cladding and the channel wall make it possible to take melting and freezing processes into account. Further features of CALIPSO are the variable channel cross section in order to model disturbances of the channel geometry as well as the calculation of two velocity fields including the consideration of virtual mass effects. The documented version of CALIPSO is especially suited for the calculation of the SIMBATH experiments carried out at the Kernforschungszentrum Karlsruhe, which simulate the above-mentioned accident. The report contains the complete documentation of the CALIPSO code: the modeling of the geometry, the equations used, the structure of the code and the solution procedure as well as the instructions for use with an application example. (orig.) [de

  5. Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium in fast breeder reactors (FBRs) fuel reprocessing

    International Nuclear Information System (INIS)

    Govindan, P.; Palamalai, A.; Vijayan, K.S.; Subba Rao, R.V.; Venkataraman, M.; Natarajan, R.

    2013-01-01

    Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium is developed for the recovery of uranium and plutonium present in spent fuel of fast breeder reactors (FBRs). Effect of pH on the solubility of carbonates of uranium and plutonium in ammonium carbonate medium is studied. Effect of mole ratios of uranium and plutonium as a function of uranium and plutonium concentration at pH 8.0-8.5 for effective separation of uranium and plutonium to each other is studied. Feasibility of reconversion of plutonium in carbonate medium is also studied. The studies indicate that uranium is selectively precipitated as AUC at pH 8.0-8.5 by adding ammonium carbonate solution leaving plutonium in the filtrate. Plutonium in the filtrate after acidified with concentrated nitric acid could also be precipitated as carbonate at pH 6.5-7.0 by adding ammonium carbonate solution. A flow sheet is proposed and evaluated for partitioning and reconversion of uranium and plutonium simultaneously in the FBR fuel reprocessing. (author)

  6. A middle evaluation report on R and D subjects in fiscal year 2000. Evaluation subject: 'investigation on actualization strategy of fast breeder reactor cycle'

    International Nuclear Information System (INIS)

    2000-08-01

    The Japan Nuclear Cycle Development Institute (JNC) consulted the titled middle evaluation to the Subject Evaluation Committee (SEC) according to the Schematic indication on practice procedure of evaluation common to the generalized national R and D' and so on. By receiving the consult, SEC on wastes treatment and disposal carried out evaluation of this subject on basis of documents proposed from JNC and discussions at SEC according to an evaluation procedure determined by SEC. This subject was already investigated at the third group on establishment of long term program on new nuclear energy and concluded its promotion, it can be said that aim and meaning of its R and D is clear. And, at a viewpoint of middle- and long-term business program of JNC, it can also be positioned to be at an important mission. In order to carry out flexible response to versatile needs in future for giving a meaning of development strategy of fast breeder reactor (FBR) cycle, under full understanding of variability of a promise condition for the aim and meaning of this research such as environment around development of FBR cycle development, R nad D of the program must be promoted. In this program picking-out of every subject and research program are adequately promoted to be enough evaluated for the middle results. Here was summarized by the evaluation results with documents proposed by JNC. As a result of the evaluation, it was shown that as general directionality of this program was judged to be valid. (G.K.)

  7. Neutron radiation damage studies in the structural materials of a 500 MWe fast breeder reactor using DPA cross-sections from ENDF / B-VII.1

    Science.gov (United States)

    Saha, Uttiyoarnab; Devan, K.; Bachchan, Abhitab; Pandikumar, G.; Ganesan, S.

    2018-04-01

    The radiation damage in the structural materials of a 500 MWe Indian prototype fast breeder reactor (PFBR) is re-assessed by computing the neutron displacement per atom (dpa) cross-sections from the recent nuclear data library evaluated by the USA, ENDF / B-VII.1, wherein revisions were taken place in the new evaluations of basic nuclear data because of using the state-of-the-art neutron cross-section experiments, nuclear model-based predictions and modern data evaluation techniques. An indigenous computer code, computation of radiation damage (CRaD), is developed at our centre to compute primary-knock-on atom (PKA) spectra and displacement cross-sections of materials both in point-wise and any chosen group structure from the evaluated nuclear data libraries. The new radiation damage model, athermal recombination-corrected displacement per atom (arc-dpa), developed based on molecular dynamics simulations is also incorporated in our study. This work is the result of our earlier initiatives to overcome some of the limitations experienced while using codes like RECOIL, SPECTER and NJOY 2016, to estimate radiation damage. Agreement of CRaD results with other codes and ASTM standard for Fe dpa cross-section is found good. The present estimate of total dpa in D-9 steel of PFBR necessitates renormalisation of experimental correlations of dpa and radiation damage to ensure consistency of damage prediction with ENDF / B-VII.1 library.

  8. Development of a Fast Breeder Reactor Fuel Bundle Deformation Analysis Code - BAMBOO: Development of a Pin Dispersion Model and Verification by the Out-of-Pile Compression Test

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

    2004-01-01

    To analyze the wire-wrapped fast breeder reactor fuel pin bundle deformation under bundle/duct interaction conditions, the Japan Nuclear Cycle Development Institute has developed the BAMBOO computer code. This code uses the three-dimensional beam element to calculate fuel pin bowing and cladding oval distortion as the primary deformation mechanisms in a fuel pin bundle. The pin dispersion, which is disarrangement of pins in a bundle and would occur during irradiation, was modeled in this code to evaluate its effect on bundle deformation. By applying the contact analysis method commonly used in the finite element method, this model considers the contact conditions at various axial positions as well as the nodal points and can analyze the irregular arrangement of fuel pins with the deviation of the wire configuration.The dispersion model was introduced in the BAMBOO code and verified by using the results of the out-of-pile compression test of the bundle, where the dispersion was caused by the deviation of the wire position. And the effect of the dispersion on the bundle deformation was evaluated based on the analysis results of the code

  9. Characterizing and packaging BN-350 spent fuel for long-term dry storage

    International Nuclear Information System (INIS)

    Lambert, J. D. B.; Bolshinsky, I.; Haues, S.L.; Allen, K.J.; Howden, E.A.; Hill, R.N.; Planchon, H.P.; Staples, P.; Karaulov, V.N.; Blynskij, A.P.; Yakovlev, I.K.; Maev, V.; Dumchev, I. A.

    2000-01-01

    The Republic of Kazakhstan is being assisted by the U.S. Department of Energy in preparing spent fuel from the BN-350 fast reactor for long term dry storage. Argonne National Laboratory was assigned responsibility for the physical and nuclear characterization of the spent fuel, for the design and safety analysis of 6-pac and 4-pac canisters used to contain spent fuel assemblies for storage, and for the design, testing and installation of a closure station at the reactor in which the canisters of fuel are dried, filled with inert gas and welded shut. This paper briefly describes the specialized components and equipment used, the process followed, and experience gained in packaging the spent fuel. Olsen et al and Schaefer separately discuss overall safety and criticality considerations of the packaging process in parallel papers to this conference

  10. Seismic monitoring of the Creys-Malville plant - Problems raised by the seismic behaviour of a fast breeder reactor

    International Nuclear Information System (INIS)

    Descleve, P.; Barrau, P.

    1988-01-01

    CREYS-MALVILLE reached full power in December 1986 and is presently the largest sodium cooled reactor in operation. Well established procedures of safety evaluation have been used for the design but for a large size reactor special attention must be paid to the effects of seismic disturbances. This paper describes the seismic protection and monitoring system of the plant, the core behaviour which is specific to fast reactors and the test performed to verify the analyses. Finally the seismic impact on the construction can be established as an indication for future plants. (author)

  11. Investigation of free level fluctuations in a simulated model of a sodium cooled Fast Breeder Reactor using pulsating conductance monitoring device

    International Nuclear Information System (INIS)

    Aggarwal, P.K.; Pandey, G.K.; Malathi, N.; Arun, A.D.; Ananthanarayanan, R.; Banerjee, I.; Sahoo, P.; Padmakumar, G.; Murali, N.

    2012-01-01

    Highlights: ► An innovative approach for measurement of water level fluctuation is presented. ► Measurement was conducted with a PC based pulsating type level sensor. ► Deployed the technique in monitoring level fluctuation in PFBR simulated facility. ► The technique helped in validation of hot pool design of PFBR, India. - Abstract: A high resolution measurement technique for rapid and accurate monitoring of water level using an in-house built pulsating conductance monitoring device is presented. The technique has the capability of online monitoring of any sudden shift in water level in a reservoir which is subjected to rapid fluctuations due to any external factor. We have deployed this novel technique for real time monitoring of water level fluctuations in a specially designed ¼ scale model of the Prototype Fast Breeder Reactor (PFBR) at Kalpakkam, India. The water level measurements in various locations of the simulated test facility were carried out in different experimental campaigns with and without inclusion of thermal baffles to it in specific operating conditions as required by the reactor designers. The amplitudes and the frequencies of fluctuations with required statistical parameters in hot water pool of the simulated model were evaluated from the online time versus water level plot in more convenient way using system software package. From experimental results it is computed that the maximum free level fluctuation in the hot pool of PFBR with baffle plates provided on the inner vessel is 30 mm which is considerably less than the value (∼82 mm) obtained without having any baffle plates. The present work provided useful information for assessment of appropriate design which would be adopted in the PFBR for safe operation of the reactor.

  12. Brazed thermocouple pass-through for sodium service in a liquid-metal-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Walker, D.E.

    1975-10-01

    Sensors installed in special fuel elements for the EBR-II reactor had 30-ft-long leads that would pass from the sodium environment through a sealed bulkhead. A hydrogen-atmosphere, induction-heated brazing furnace was constructed to simultaneously braze 20-26 separate sensor leads at one time. The brazed seals were leak-tight, and the sheath wall has less than 10 percent interaction with the braze alloy

  13. A COMPARISON OF (PuU)C AND (PuU)O$sub 2$ FUELED FAST BREEDER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Kazi, A. Halim; Rosen, Stanley

    1963-11-15

    BS>A conceptual evaluation is presented of a (PuU)C 500Mw(e) reactor which uses a moderate power generation of 40 kw/ft and a maximum fuel temperature of 2900 deg F, well below the 4500 deg F melting point of Pu/sub 0.2/U/sub 0.8/C. The principal design characteristics and fuel cycle costs are tabulated together with those of a (PuU)O/sub 2/ design of equal power. The comparison shows that the carbide core has significant economic and safety advantages over the oxide core. (D.L.C.)

  14. Collection and transmission of milli-volt-level data on a liquid-metal fast-breeder reactor

    International Nuclear Information System (INIS)

    Robinson, C.; Thoburn, R.

    1978-01-01

    The paper describes a high integrity data multiplexing system which caters for up to 252 input elements with remote measuring facilities, using a 12-core flexible cable. Tests carried out on a prototype unit and recommendations for effecting improvements are described. The system was developed following the identification of the problem of having to handle a large number of cables on the top of the Commercial Demonstration Fast Reactor. It has high reliability, a relatively small and conventional cable installation and is applicable to situations outside the nuclear field. (author)

  15. Investigation on analytical method for the transfer behavior of corrosion product (CP) in the fast breeder reactors

    International Nuclear Information System (INIS)

    Matsuo, Youichirou; Sasaki, Shinji

    2013-05-01

    Radioactive corrosion products (CPs) are main cause of personal radiation exposure during maintenance work without fuel failure in FBR plants. The most important CP species are 54 Mn and 60 Co. The deposited radioactive CPs cause radiation fields near the piping and components, and the CPs contribute to the radiation exposure of the plant-worker. In this review, firstly, the collected knowledge about CP transfer behavior in the fast reactor are reviewed and analyzed. Secondly, the existing analytical methods to evaluate CP transfer behavior are investigated, issues of which and their solutions are extracted and discussed. Finally, examples of the calculated results by the improved analytical method are described. The provided conclusions are as follows; (1) Collected knowledge on CP transfer behavior. CP generation is mainly due to the dissolution of CP from hot reactor core constitution materials to hot sodium. On the core materials, particle-formed structure was confirmed. The evidence of CP precipitation in the low temperature part of the primary cooling system and the lower part of reactor core was provided. Similarly, the evidence of CP particle deposition in the same domain was also provided. (2) Extracted issues on analytical methods of CP transfer and proposed solutions. In the past, radioactivity caused by CP deposition on the piping and the core materials surface is confirmed. Subsequently, analytical models were developed based on the distribution of the CPs in the reactor coolant systems and the out-pile sodium loop test. The local high radiation dosage (such as elbow part) was observed by the radiation measurement. However, this behavior cannot be evaluated by the existing model, and it is considered necessary to take into account the transfer of CP particle. (3) The recent trend of the CP behavioral analysis method. Novel CP particle generation, transfer and deposition models were developed based on existing knowledge on CP behavior. The developed

  16. Research and development long term prospects for advanced thermal and fast breeder reactors and fuel cycle activities

    International Nuclear Information System (INIS)

    Rapin, M.

    1987-01-01

    Despite the present slowing down of most of the national nuclear programmes due to various conjunctural causes, nuclear energy is presumably to have in term an increasing role worldwide for electricity production, both in industrialized countries already equipped with nuclear power stations and in countries which intent to enter the nuclear energy route in a short or mid-term perspective. Nuclear energy development will have notably to take into account the following requirements: from a strategic point of view, the major concern is to optimize the use of world available natural uranium appears abundant and inexpensive, it must not be forgotten that in a longer time perspective a growing scarcity of this material will occur unavoidably; from the technical and economical standpoints, future reactors will have to be more reliable, more available, and even more safe, with lower KWh generating costs, than units presently constructed or operated. The paper presents the approach which is followed by France to meet these goals. (author)

  17. Status of the Fast Breeder Reactor Development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    Hüper, R.; Friese, R.

    1980-01-01

    In 1979 some progress was made on the Kalkar site of the SNR 300 as a consequence of the partial construction license issued on December 20, 1978. This license covers part of the assembly work related to mechanical and electrical engineering. The civil engineering work has nearly been completed except for the cooling tower. Further progress on site will depend on the additional construction licenses being issued, especially those concerning decay heat removal and emergency core cooling systems, the reactor proper, and the regular heat transfer systems. Fabrication and testing of the components is proceeding satisfactorily. At the Karlsruhe Nuclear Research Center, the 20 MWe KNK II plant with sodium cooling and a fast neutron core has gone into operation. Valuable experience has been accumulated, e.g., from measurements of the entrapment of cover gas in the sodium. This entrainment has been reduced considerably by the installation of valves in a ventilation pipe. A failed fuel element was identified and exchanged. The experimental program has begun

  18. Simulation of the cladding freezing during the loss of flow accident in a Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chu, N.N.; Eggen, D.T.

    1977-06-01

    The accident condition of the stainless steel cladding melting, relocation and freezing, following a loss of flow situation in gas-cooled fast reactors has not been determined yet. An alloy of 50 Pb/ 50 Sn was selected to facilitate experimental procedures because of its workability and its wide phase transition temperature range (183 to 216 0 C) similar to that of stainless steel (1375 to 1500 0 C). The objective of the experiment is to simulate the motion of liquid alloy through a tube and observe the conditions where it solidifies in the tube. The objective of the theoretical model is to use transient heat transfer analysis to describe the freezing front of the moving liquid metal and the plugging of the channel. Tests were conducted in alloy tubes having inside diameters 0.5 to 0.95 cm. Molten solder was poured through the vertically held tubes. The average falling velocity of the melt was measured to be about 89 cm/sec. The distance in the tube where the molten Pb/Sn solidifed across the diameter was measured. This penetration length varies from 20 to 40 cm as the initial liquid temperature ranges from 216 (liquidus point) to 500 0 C. The plugging time is calculated for those temperatures under which data on both the penetration distance and the falling velocity were obtained

  19. Research and development long term prospects for advanced thermal and fast breeder reactors and fuel cycle activities

    International Nuclear Information System (INIS)

    Rapin, M.

    1988-01-01

    Despite the present slowing down of most of the national nuclear programmes due to various conjunctural causes, nuclear energy is presumably to have in term an increasing role worldwide for electricity production, both in industrialized countries already equipped with nuclear power stations and in countries which intent to enter the nuclear energy route in a short or mid-term perspective. Nuclear energy development will have notably to take into account the following requirements: (1) From a strategic point of view, the major concern is to optimize the use of the world available natural uranium resources. Indeed, even if presently natural uranium appears abundant and inexpensive, it must not be forgotten that in a longer time perspective a growing scarcity of this material will occur unavoidably. (2) From the technical and economical standpoints, future reactors will have to be more reliable, more available, and even more safe, with lower kWh generating costs, than units presently constructed or operated. The paper presents the approach which is followed by France to meet these goals

  20. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    1978-09-01

    Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

  1. The implication of sensitivity analysis on the safety and delayed-neutron parameters for fast breeder reactors

    International Nuclear Information System (INIS)

    Onega, R.J.; Florian, R.J.

    1983-01-01

    The delayed-neutron energy spectra for LMFBRs are not as well known as those for LWRs. These spectra are necessary for kinetics calculations which play an important role in safety and accident analyses. A sensitivity analysis was performed to study the response of the reactor power and power density to uncertainties in the delayed-neutron spectra during a rod-ejection accident. The accidents studied were central control-rod-ejections with ejection times of 2,10 and 30s. A two-energy group and two-precursor group model was formulated for the International Nuclear Fuel Cycle Evaluation (INFCE) reference design MOX-fueled LMFBR. The sensitivity analysis is based on the use of adjoints so that it is not necessary to repeatedly solve the governing (kinetics) equations to obtain the sensitivity derivatives. This is of particular importance when large systems of equations are used. The power and power-density responses were found to be most sensitive to uncertainties in the spectrum of the second delayed-neutron precursor group, resulting from the fission of 238 U, producing neutrons in the first energy group. It was found, for example, that for a rod-ejection time of 30s, and uncertainty of 7.2% in the fast components of the spectra resulted in a 24% uncertainty in the predicted power and power density. These responses were recalculated by repeatedly solving the kinetics equations. The maximum discrepancy between the recalculated and the sensitivity analysis response was only 1.6%. The results of the sensitivity analysis indicate the need for improved delayed-neutron spectral data in order to reduce the uncertainties in accident analyses. (author)

  2. UK status report on detection and localisation of leaks in steam generators of liquid metal fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Smedley, J A [Dounreay Nuclear Power Development Establishment, Caithness, Scotland (United Kingdom)

    1978-10-01

    The development of detection and location techniques applied to defects in sodium heated steam generators in Uk has been associated primarily with the PFR commissioning program. The UK position on leak detection and location studies for LMFBR steam generators may be briefly summarised as follows: a) The Initial concept of' detection using hydrogen concentration measuring equipment in secondary circuit sodium and in steam generator gas spaces has proved sound. The system used on the PFR has been shown to work well under plant conditions in both the under sodium and gas phase modes. b) Experience with small leaks in tube to tube plate welds in the PFR steam generators has indicated the majority of leaks would be detected by the installed equipment. There is, however, a very small possibility an under sodium leak in. a ferritic tube which will quickly and easily block may not be detected before it erupts as a relatively significant defect. Methods are being sought to protect against this unlikely event, although it is recognised there is no reactor hazard but there may be a plant availability problem. c) It is concluded the single barrier concept is still valid, based on the experience in the PFR steam generator units. For future units it is intended to use ferritic boiler tube materials rather than austenitic materials. In general, however, it is concluded the material choice has been satisfactory and manufacturing methods are basically sound. Improved quality assurance methods are being sought with the aim of making future steam generators more reliable than those presently in service. d) It is recognised steam generators are still in a development regime in all countries of the world working in the LMFBR field, but the time is anticipated when a utility can order an LMFBR as a commercial proposition. An attempt has been made to quantify the information necessary to achieve this state.

  3. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  4. Fast Breeder Project status report, 1974

    International Nuclear Information System (INIS)

    Hueper, R.

    A compilation of the papers read at the Status Report of the Fast Breeder Project at the Karlsruhe Nuclear Research Center on March 26, 1974 is presented. The first papers present a general survey of the present state of research and development work performed by the German, Belgian, and Netherlands research centers on the SNR 300 Prototype Fast Breeder Reactor (Kalkar Nuclear Power Station), on the SNR follow-on program, alternative fuels and coolants and basic problems, on work performed by industry with respect to the licensing procedure and construction of the SNR 300, and on commissioning and the planned conversion of KNK, the Compact Sodium Cooled Nuclear Reactor. The detailed papers deal with results elaborated at the institutes of GfK Karlsruhe, SCK/CEN Mol, RCN Petten, and TNO Apeldoorn. Most of these efforts have been concentrated upon fuel pin and materials development and on the physics and safety of fast reactors. The status report concludes with a reference to the future program under the Project. (U.S.)

  5. Analysis of self-wastage phenomena of micro leak caused by sodium-water reaction in sodium-cooled fast breeder reactor through simulant experiment

    International Nuclear Information System (INIS)

    Jang, Sunghyon; Takata, Takashi; Yamaguchi, Akira

    2014-01-01

    Self-wastage phenomena are an enlargement of a leak on the heat transfer tube caused by a corrosive sodium-water reaction (SWR) in a steam generator (SG) of sodium-cooled fast breeder reactor (SFR). If the steam generator operates for sometimes under this condition, the self-wastage phenomena start from the sodium side and advance through the tube thickness. The leak rate stays almost constant level until the wastage reaches the sodium side, however, when the thin diaphragm of the tube wall is removed, the leak rate sharply increase, and it may bring a secondary failure of the surrounding heat transfer tubes. The design and safety concern is a possibility of the secondary failure of nearby SG tubes that could cause undesirable development of the accidents. One needs to evaluate the increased resultant leak rate due to the self-wastage phenomenon. Therefore, a quantification of the diameter of enlarged leak is needed to estimate the resultant leak rate. For this purpose, a simulant self-wastage experiment was proposed to investigate the self-enlargement of the leak so that evaluate the mechanism of the Self-wastage. In the experiment, high concentrated hydrochloric acid (HCl) is injected to the reaction tank that is filled sodium hydroxide (NaOH) solution through a nozzle made by paraffin wax. The self-enlargement of the leak was evaluated by considering the melted nozzle due to the reaction heat released from the Neutralization reaction. Also, a numerical investigation has been carried out to evaluate the enlarged nozzle and validate the results of experimental methodology. Based on the experimental and computational results, it is found that despite initial leak rate, there is an upper limit in the enlarged nozzle. These results show a similar tendency with the experimental result of SWAT-4 experiment carried out by Power Reactor and Nuclear Fuel Development Corporation (PNC), Japan. Furthermore, the increased resultant leak rate is evaluated using the enlarged

  6. 3rd quarterly report 1976 of the Fast Breeder Project

    International Nuclear Information System (INIS)

    1976-12-01

    The report describes activities which were performed within the framework of the Fast Breeder Project at the Gesellschaft fuer Kernforschung mbH Karlsruhe (GfK) or on behalf of the GfK during the third quarter. It contains contributions on the following subjects: Fuel rod development, material studies and development, corrosion tests and coolant analyses, physical experiments, reactor theory, safety of fast breeders, instrumentation and signal processing for core monitoring, environmental impacts, sodium technology tests, thermo- and fluid-dynamic tests in gas, tests concerning gas-cooled breeders. (HR) [de

  7. Feasibility study on commercialization of fast breeder reactor cycle systems interim report of phase II. Technical study report for nuclear fuel cycle systems

    International Nuclear Information System (INIS)

    Sato, Koji; Amamoto, Ippei; Inoue, Akira

    2004-06-01

    As a part of the feasibility study on commercialization of fast breeder reactor cycle systems, the plant concept concerning the fuel cycle systems (combination of the reprocessing and the fuel fabrication) has been constructed to reduce their total cost by the introduction of various innovative techniques and to apply their utmost superior efficiency from such standpoints of a decrease in the environmental burden, better resource utilization and proliferation resistance improvement by the low decontamination transuranium element (TRU) recycle. This interim report of Phase II describes the results of an on-going study which will cover a five-year period. For oxide fuels, the system which combines the use of the advanced aqueous reprocessing using three main methods such as the crystallization method, the simplified solvent extraction method, and the extraction chromatography method for minor actinide (MA) recovery, as well as the simplified pelletizing fuel fabrication which rationalized a powder mixing process etc., has abundant current results and a high technical feasibility for the basic process. Though this system faces difficulties in the technical development of control technology of the extraction chromatography and the fabrication technology of low decontamination TRU fuel etc., its expected practical use is possible at an early stage. As for the super-critical direct extraction reprocessing, it is necessary to fulfill more basic data although further economical improvement of an advanced aqueous reprocessing is expected. The system which combines the advanced aqueous reprocessing and the gelation sphere packing fuel fabrication has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the fuel fabrication process. However, this system will shoulder additional cost for the reagent recovery process and the waste liquid treatment process due to need to dispose of a large bulk of process waste liquid. The system which

  8. Fast breeder fuel cycle, worldwide and French prospects

    International Nuclear Information System (INIS)

    Rapin, M.

    1982-01-01

    A review is given of fast breeder fuel cycle development from both the technological and the economical points of view. LMFBR fuel fabrication, reactor operation, spent fuel storage and transportation, reprocessing and fuel cycle economics are topics considered. (U.K.)

  9. Testing of the dual slab verification detector for attended measurements of the BN-350 dry storage casks

    Energy Technology Data Exchange (ETDEWEB)

    Santi, Peter A [Los Alamos National Laboratory; Browne, Michael C [Los Alamos National Laboratory; Williams, Richard B [Los Alamos National Laboratory; Parker, Robert F [Los Alamos National Laboratory

    2009-01-01

    The Dual Slab Verification Detector (DSVD) has been developed and built by Los Alamos National Laboratory in cooperation with the International Atomic Energy Agency (IAEA) as part of the dry storage safeguards system for the spent fuel from the BN-350 fast reactor. The detector consists of two rows of {sup 3}He tubes embedded in a slab of polyethylene which has been designed to be placed on the outer surface of the dry storage cask. The DSVD will be used to perform measurements of the neutron flux emanating from inside the dry storage cask at several locations around each cask to establish a neutron 'fingerprint' that is sensitive to the contents of the cask. The sensitivity of the fingerprinting technique to the removal of specific amount of nuclear material from the cask is determined by the characteristics of the detector that is used to perform the measurements, the characteristics of the spent fuel being measured, and systematic uncertainties that are associated with the dry storage scenario. MCNPX calculations of the BN-350 dry storage asks and layout have shown that the neutron fingerprint verification technique using measurements from the DSVD would be sensitive to both the amount and location of material that is present within an individual cask. To confirm the performance of the neutron fingerprint technique in verifying the presence of BN-350 spent fuel in dry storage, an initial series of measurements have been performed to test the performance and characteristics of the DSVD. Results of these measurements will be presented and compared with MCNPX results.

  10. The future of the Fast Breeder

    International Nuclear Information System (INIS)

    Lefevre, J.C.

    1990-01-01

    Fast Breeder Reactors (FBRs) can produce more fissile nuclei than they consume whilst, at the same time, generating energy using fast neutrons. By conversion of uranium isotope 238 into a fissionable fuel, FBRs provide over 60 times more energy than can be extracted from the uranium reserves by thermal reactors. Their development is therefore an essential objective in the next century, particularly for those industrialised countries that have little or no energy resources of their own. The European countries which have been engaged in the development of FBRs for more than 25 years have decided to collaborate in an advanced design, the European Fast Reactor (EFR) which uses the best of previous national projects and draws on extensive operating experience from FBR plants in Europe. The naturally safe characteristics and technological features of sodium-cooled Fast Reactors will be fully utilised in an EFR design which meets the same safety level as the Light Water Reactors (LWRs). Owing to technical progress and series construction effect, the EFR is expected to achieve competitiveness with contemporary LWRs with the higher capital cost of the Fast Reactor offset by its markedly lower fuel cycle cost. (author)

  11. Role of fast breeders in Japan

    International Nuclear Information System (INIS)

    Oyama, A.; Tomabechi, K.

    1978-09-01

    To meet increasing future energy demand in Japan utilization of fission energy should be promoted. In particular it is of vital importance to develop and utilize FBRs as soon as possible in order to save the natural uranium needed. If one considers the commercial introduction of FBRs in the mid-1990s in Japan, a delay of only one year will eventually result in an additional demand for natural uranium of more than 20,000 tons, because several LWRs will have to be installed instead. Ten years have passed since the development of FBRs in Japan was initiated as a national project with the highest priority and now the experimental fast reactor JOYO is successfully being operated at 50MW and the prototype fast breeder reactor MONJU has reached the stage of proceeding to construction with a schedule of operation in the mid-1980s. Following operation of MONJU, construction of a large demonstration reactor of 1000 - 15000 MW(e) will be undertaken. Some 2 - 3 years after the construction of the demonstration reactor, a series of reactors will be constructed similar in size and design to promote commercialization of LMFBRs. Strong efforts will be made to put this programme into practice. It is expected that LMFBRs will play an important role in mitigating the serious problem of energy supply in Japan foreseeable around the turn of the century

  12. Fast breeder reactor decision: an analysis of limits and the limits of analysis. A study prepared for the use of the Joint Economic Committee, Congress of the United States, Ninety-Fourth Congress, Second Session

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    Decisions on the development of the Liquid Metal Fast Breeder Reactor (LMFBR), which may involve nontransferable gains for current generations and major losses for future generations cannot be determined on a cost-benefit basis. Several cost-benefit analyses of the LMFBR are reviewed. With only one exception, all major analyses compare LMFBR futures with non-LMFBR futures and consider only the energy supply aspect of alternatives. The Manne study, however, compares peak-load pricing of electricity and points out that removal of distortions in energy pricing has greater economic benefits than economic losses due to delays in introducing the LMFBR. Revised price structures for electricity could have a larger future payoff than rapid LMFBR development. Constraints on LMFBR development, such as resource supplies, technological development, licensing, siting, and waste disposal, may have been underestimated. Appropriate institutions and procedures for nuclear decision making must be defined

  13. Dounreay: more is at stake than just the fast breeder

    International Nuclear Information System (INIS)

    Holmes, A.

    1988-01-01

    The British government's policy on nuclear power is to support and encourage an expansion of the nuclear energy programme. However, the Conservative government has said it will withdraw its funding from the fast breeder reactor research programme based at Dounreay in Scotland. Once the Electricity supply industry has been privatised it will not finance the fast breeder reactor programme either. So it seems likely that the research will be discontinued after 5 years and the reprocessing plant planned for Dounreay will never be built. This highlights the basic ambiguity that has existed for a long time over the fast breeder programme and has forced a decision. A similar problem exists generally with the United Kingdom Atomic Energy Authority which has continued to thrive despite many failures. The reasons for this are explained. Privatisation may be the spur to the demise of the UKAEA so it no longer enjoys such a favoured position. The decision to kill off the FBR is just one consequence of this. Others may be a change of attitude to the decommissioning of the Windscale prototype advanced gas cooled reactor, and the fusion project, including JET. (U.K.)

  14. Code HEX-Z-DMG for support of accounting for and control of nuclear material software system as part of international safeguards system at BN-350 site

    International Nuclear Information System (INIS)

    Bushmakin, A.G.; Schaefer, B.

    1999-01-01

    A code for the computation of the global neutron distribution in the three-dimensional hexagonal-z geometry and multi-group diffusion approximation was developed at BN-350 as the main part of the BN-350 accounting for and control of nuclear material software system. This software system includes: the model for stationary distributions of neutrons; the model to calculate isotope compositions changing; the model of refueling operations; To develop this system next two principal problems were solved: to make a micro cross sections library for all nuclides for the BN-350 reactor core; to develop the code for the computation of the global neutron distribution. To solve first task the twenty-six-energy-groups micro cross sections library for more than seventy nuclides was produced. To solve second task the three-dimensional hexagonal-z geometry and multi-group diffusion approximation code was developed. This code (HEX-Z-DMG) was based on the solution of the multi groups diffusion equation using the standard net approach. The series of calculations was performed in the twenty-six-energy-groups representation using this code. We compared eigenvalues (k eff ), a worth added during refueling operations, spatial and energy-group-dependent neutron flux distributions with results of calculation using other code (DIF3D). After the series of these calculations we can say that the HEX-Z-DMG code is well established to use as the part of the BN-350 accounting for and control of nuclear material software system. (author)

  15. 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

  16. Italian position paper on the safety analysis of liquid metal fast breeder reactors as related to sodium fires. The PEC reactor

    International Nuclear Information System (INIS)

    Gerosa, A.

    1983-01-01

    To obtain a deep understanding of physical phenomena and engineering problems connected to sodium fires, and to optimize the utilization of human and financial resources available, CNEN (now ENEA) has decided to join the French Commissariat a l'Energie Atomique (CEA) in the realization of a Franco-Italian experimental programme on sodium fires, named ESMERALDA. As for design preventions for PEC reactor (a fast flux, liquid metal cooled, fuel element testing reactor) fundamental choices were made taking into account all available knowledge, but with particular reference to the results of CEA's previous experiments on sodium fires. More detailed design analysis will be possible in the future, based on experimental results coming from the ESMERALDA programme

  17. On the history of the Fast Breeder Project

    International Nuclear Information System (INIS)

    Marth, W.

    1981-07-01

    The evolution of the Fast Breeder Project from its beginning at the Karlsruhe Nuclear Research Center to the present cooperation of various organisations especially in the Federal Republic of Germany, the Netherlands, Belgium and France is described in its historical context. Where as the emphasis was on physical studies of fast neutron cores in the early phase, technological and safety problems gained importance in the subsequent development. The increasing collaboration with industry and the support by government funds resulted in the design and start of construction of the prototype SNR 300. The objectives and the reasoning underlying important intermediate decisions are described. In the meantime, licensing and funding problems have become decisive for the project schedule. The present report also gives an account of the international and national political aspects which influence the breeder reactor development. In the annex all fast breeder publications of the Karlsruhe Nuclear Research Center are listed. (orig.) [de

  18. Fast breeder physics and nuclear core design

    International Nuclear Information System (INIS)

    Marth, W.; Schroeder, R.

    1983-07-01

    This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.) [de

  19. Alternative cycles and fast breeders, a look to the future

    International Nuclear Information System (INIS)

    Dahlberg, R.C.

    1979-01-01

    The various nuclear strategies that have been proposed to eke out available uranium are briefly summarised. A recent approach is to develop the concept of a 'transitional symbiosis strategy' which accepts the principle of the fast breeder/advanced converter for the long-range future, and to suggest that advanced converters be designed to be near-breeders. Some of the issues affecting strategic planning in the transitional period are reviewed further. The need for the thorium cycle in both advanced converters (or near breeders) and fast reactors is emphasised. This type of high temperature gas-cooled reactor appears to be the technology most suited for symbiosis with FBRs. (UK)

  20. Status of fast breeder development in Germany

    International Nuclear Information System (INIS)

    Heusener, G.

    1992-01-01

    The German Minister for Research and Technology (BMFT), Dr. Heinz Riesenhuber, announced on March 20, 1991 that SNR 300, the fast breeder power plant at Kalkar, shall be abandoned. This message followed a top level meeting between BMFT officials and senior managers of Siemens, RWE, PreuBenElektra und Bayernwerk. BMFT, vendor Siemens and the three utilities had carried the interim finance costs of DM 105 million yearly since 1989. The licensing procedure had been obstructed during a long time by the responsible authorities. For several years the licensing process for the last permits on nuclear operation of KKW Kalkar had been held up by the government of the state of North Rhine-Westphalia (NWR). Licensing of nuclear power plants is the responsibility of the states, according to the German Atomic Act. The state of NRW turned against the SNR 300 project when the Social Democratic Party (SPD) started questioning nuclear power in 1985. Until then 17 partial licenses for SNR 300 had been granted, each time including an overall project approval. One of the consequences of the demise of SNR-300 was that Interatom GmbH, a subsidiary of Siemens AG, has been integrated into the division KWU of the Siemens AG on 1 October, 1991. For SNR 300 the turn-key contracts to the supplier company were cancelled by the operator on April 10, 1991 following the political termination of the SNR-300 Project. On August 23, 1991 after the termination of the SNR project, KfK decided to shutdown the KNK II reactor for final decommissioning

  1. Status of fast breeder development in Germany

    Energy Technology Data Exchange (ETDEWEB)

    Heusener, G [Kernforschungszentrum Karlsruhe, Karlsruhe (Germany)

    1992-07-01

    The German Minister for Research and Technology (BMFT), Dr. Heinz Riesenhuber, announced on March 20, 1991 that SNR 300, the fast breeder power plant at Kalkar, shall be abandoned. This message followed a top level meeting between BMFT officials and senior managers of Siemens, RWE, PreuBenElektra und Bayernwerk. BMFT, vendor Siemens and the three utilities had carried the interim finance costs of DM 105 million yearly since 1989. The licensing procedure had been obstructed during a long time by the responsible authorities. For several years the licensing process for the last permits on nuclear operation of KKW Kalkar had been held up by the government of the state of North Rhine-Westphalia (NWR). Licensing of nuclear power plants is the responsibility of the states, according to the German Atomic Act. The state of NRW turned against the SNR 300 project when the Social Democratic Party (SPD) started questioning nuclear power in 1985. Until then 17 partial licenses for SNR 300 had been granted, each time including an overall project approval. One of the consequences of the demise of SNR-300 was that Interatom GmbH, a subsidiary of Siemens AG, has been integrated into the division KWU of the Siemens AG on 1 October, 1991. For SNR 300 the turn-key contracts to the supplier company were cancelled by the operator on April 10, 1991 following the political termination of the SNR-300 Project. On August 23, 1991 after the termination of the SNR project, KfK decided to shutdown the KNK II reactor for final decommissioning.

  2. Tridimensional ultrasonic images analysis for the in service inspection of fast breeder reactors; Analyse d'images tridimensionnelles ultrasonores pour l'inspection en service des reacteurs a neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Dancre, M

    1999-11-01

    Tridimensional image analysis provides a set of methods for the intelligent extraction of information in order to visualize, recognize or inspect objects in volumetric images. In this field of research, we are interested in algorithmic and methodological aspects to extract surface visual information embedded in volume ultrasonic images. The aim is to help a non-acoustician operator, possibly the system itself, to inspect surfaces of vessel and internals in Fast Breeder Reactors (FBR). Those surfaces are immersed in liquid metal, what justifies the ultrasonic technology choice. We expose firstly a state of the art on the visualization of volume ultrasonic images, the methods of noise analysis, the geometrical modelling for surface analysis and finally curves and surfaces matching. These four points are then inserted in a global analysis strategy that relies on an acoustical analysis (echoes recognition), an object analysis (object recognition and reconstruction) and a surface analysis (surface defects detection). Few literature can be found on ultrasonic echoes recognition through image analysis. We suggest an original method that can be generalized to all images with structured and non-structured noise. From a technical point of view, this methodology applied to echoes recognition turns out to be a cooperative approach between morphological mathematics and snakes (active contours). An entropy maximization technique is required for volumetric data binarization. (author)

  3. Comparison calculation of a large sodium-cooled fast breeder reactor using the cell code MICROX-2 in connection with ENDF/B-VI and JEF-1.1 neutron data

    International Nuclear Information System (INIS)

    Pelloni, S.

    1992-02-01

    We have obtained results for a large sodium-cooled fast breeder reactor benchmark using data from the ENDF/B-VI and from Revision 1 of the JEF-1 (JEF-1.1) evaluation. The required cross sections were processed with the NJOY code system (Version 89.62) and homogenized with the spectrum cell code MICROX-2. Multigroup transport-theory calculations in 33 neutron groups (forward and adjoint) were performed using the two-dimensional code TWODANT and kinetic parameters were determined using the first-order perturbation-theory code PERT-V. We calculated eigenvalues, neutron balance data, global and regional breeding and conversion ratios, central rate ratios and reactivity worths with and without sodium, effective delayed neutron fraction and inhour reactivity, regional sodium void reactivity, and isothermal core fuel Doppler-reactivities. In particular, it is shown that good agreement (generally within one standard deviation) is achieved between these results and the average values over sixteen benchmark solutions obtained in the past. The eigenvalues predicted with ENDF/B-VI are up to 0.7% larger than those calculated with JEF-1.1 cross sections. This discrepancy is mainly due to different inelastic scattering cross sections for 23 Na and 238 U, and to different fast fission and nubar data for 239 Pu. (author) 5 figs., 30 tabs., 24 refs

  4. Neutron cross section measurements for the Fast Breeder Program

    International Nuclear Information System (INIS)

    Block, R.C.

    1979-06-01

    This research was concerned with the measurement of neutron cross sections of importance to the Fast Breeder Reactor. The capture and total cross sections of fission products ( 101 102 104 Ru, 143 145 Nd, 149 Sm, 95 97 Mo, Cs, Pr, Pd, 107 Pd, 99 Tc) and tag gases (Kr, 78 80 Kr) were measured up to 100 keV. Filtered neutron beams were used to measure the capture cross section of 238 U (with an Fe filter) and the total cross section of Na (with a Na filter). A radioactive neutron capture detector was developed. A list of publications is included

  5. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO{sub 2}-PuO{sub 2} fuel; Contribucion al analisis del comportamiento termico de las barras combustibles de UO{sub 2}-PuO{sub 2} de los reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J; Elbel, H

    1977-07-01

    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs.

  6. Turning into carbonate the residual sodium left in BN-350 circuits may alleviate concerns over their long term safe confinement

    International Nuclear Information System (INIS)

    Rahmani, L

    2000-01-01

    After the coolant is drained from the reactor vessel and from the primary and secondary circuits of the BN-350 nuclear power plant, what sodium is left in ponds and films may amount to hundreds of kilograms. For the long term safe storage period which is to follow, preliminary safety analyses (e.g. derived from those made for French sodium cooled reactors) might show that the risks incurred through loss of leaktightness are significant. The ingress of moisture into the circuits would generate, by reaction with the sodium, two undesirable products : sodium hydroxide and hydrogene. Even when considering that water would enter the circuits progressively, so that the heat of the reaction does not give rise to over-pressure, some main risk factors remain. The most promising solution to this challenge appears to be the carbonation of the sodium residues, by progressive diffusion of an appropriate association of carbon dioxyde and water vapour through the inert gaseous medium which fills the circuits. The desired product is porous sodium hydrogenocarbonate

  7. Estimates of time-dependent fatigue behavior of Type 316 stainless steel subject to irradiation damage in fast breeder and fusion power reactor systems

    International Nuclear Information System (INIS)

    Brinkman, C.R.; Liu, K.C.; Grossbeck, M.L.

    1978-01-01

    Cyclic lives obtained from strain-controlled fatigue tests at 593 0 C of specimens irradiated in the experimental breeder reactor II (EBR-II) to a fluence of 1 to 2.63*10 26 neutrons (n)/m 2 (E>0.1 MeV) were compared with predictions based on the method of strain-range partitioning. It was demonstrated that, when appropriate tensile and creep-rupture ductilities were employed, reasonably good estimates of the influence of hold periods and irradiation damage on the fully reversed fatigue life of Type 316 stainless steel could be made. After applicability of this method was demonstrated, ductility values for 20 percent cold-worked Type 316 stainless steel specimens irradiated in a mixed-spectrum fission reactor were used to estimate fusion reactor first-wall lifetime. The ductility values used were from irradiations that simulate the environment of the first wall of a fusion reactor. Neutron wall loadings ranging from 2 to 5 MW/m 2 were used. 27 refs

  8. Estimates of time-dependent fatigue behavior of type 316 stainless steel subject to irradiation damage in fast breeder and fusion power reactor systems

    International Nuclear Information System (INIS)

    Brinkman, C.R.; Liu, K.C.; Grossbeck, M.L.

    1979-01-01

    Cyclic lives obtained from strain-controlled fatigue tests at 593 0 C of specimens irraidated in the experimental breeder reactor II (EBR-II) to a fluence of 1 to 2.63 x 10 26 neutrons (n)/m 2 E > 0.1 MeV) were compared with predictions based on the method of strain-range partitioning. It was demonstrated that, when appropriate tensile and creep-rupture ductilities were employed, reasonably good estimates of the influence of hold periods and irradiation damage on the fully reversed fatigue life of Type 316 stainless steel could be made. After applicability of this method was demonstrated, ductility values for 20% cold-worked Type 316 stainless steel specimens irradiated in a mixed-spectrum fission reactor were used to estimate fusion reactor first-wall lifetime. The ductility values used were from irradations that simulate the environment of the first wall of a fusion reactor. Neutron wall loadins ranging from 2 to 5 MW/m 2 were used. Results, although conjectural because of the many assumptions, tended to show that 20% cold-worked Type 316 stainless steel could be used as a first-wall material meeting a 7.5 go 8.5 MW-year/m 2 lifetime goal provided the neutron wall loading does not exceed more than about 2 MW/m 2 . These results were obtained for an air environment, ant it is expected that the actual vacuum environment will extend lifetime beyond 10 MW-year/m 2

  9. Evaluation for the effects of a ring plate device to eliminate free surface gradients in liquid metal fast breeder reactor vessel using multi-dimensional thermohydraulics computer code

    International Nuclear Information System (INIS)

    Gao Ming Qing.

    1997-02-01

    There is a free surface at the upper plenum in a reactor vessel of LMFBR. The free surface has spatial gradient caused by the internal coolant flow. This is a disadvantageous factor to engineering from the view point of gas entrainment into coolant. To eliminate the free surface gradients, ring plates about 20 cm wide are fitted at about 1 meter under the free surface. They interfere fluid flow, and decrease the component velocity in vertical direction. To investigate the efficiency of the ring plates, analyses with the AQUA-VOF code were carried out. For contrast, three conditions were given: Case-1: Without ring plates. Case-2: Ring plates, fitted at 1.125 m under the free surface. Case-3: Ring plates, fitted at 1.5 m under the free surface. The results shown that the ring plates have a sufficiently high potential to eliminate the free surface gradients due to disperse the momentum along reactor vessel axis to radial direction. In the calculations with ring plate (Cases-2 and -3), the maximum free surface height differences and the maximum gradients of free surface were decreased to less than 15% and 64% compared with the case without ring plates, respectively. (author)

  10. Feasibility study on commercialization of fast breeder reactor cycle system. Interim report of phase 2. Technical study report on synthetic evaluation for FBR cycle

    International Nuclear Information System (INIS)

    Shiotani, Hiroki; Ohtaki, Akira; Ono, Kiyoshi; Yasumatsu, Naoto; Kubota, Sadae; Heta, Masanori

    2004-09-01

    This report presents the outline of the development and the results of Synthetic evaluation on the candidate Fast Reactor (FR) cycle system concepts, scenario study on FR cycle deployment and cost-benefit analysis on the candidate FR cycle system concepts in the interim evaluation (FY2001 through FY2003) of the phase 2 of the Japanese 'Feasibility Study on Commercialization of Fast Reactor Cycle System (FS)'. The characteristic evaluation extended to evaluate a new view point of social acceptance besides the viewpoints of safety, economics, reduction of environmental burden, efficient utilization of uranium resource, proliferation resistance, and technical feasibility, which has been considered since the phase 1 of FS. As for the six view points, hierarchy structures and utility functions for quantitative evaluation have been developed and/or improved. Furthermore, the methodology for weighing the viewpoints, which was also developed, made it possible to examine the characteristics of the candidate concepts from all the seven viewpoints. Generally, the FR cycles with sodium-cooled FR were highly evaluated. The characteristic evaluation for alternative power supply systems was also tried in this report for the first time. FR cycle deployment scenarios clarified the necessity of FR cycle deployment and the desirable core features, etc. through the long-term mass flow analysis, which includes comparison among other nuclear fuel cycle schemes and analysis for evaluating the degree to meet future needs, on the typical FR cycle systems. Regarding cost-benefit analysis, both the amount of the cost estimated by the past R and D and the cost in the Road map of FS are used as the investment for FR cycle research and development (R and D), the results showed that the benefit derived from the commercialization of FR cycle will be more than the investment. (author)

  11. Report on the specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors organized by the international atomic energy agency at Oarai Engineering Centre of power reactor and nuclear development corporation, Japan, November 5-7, 1991

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1992-01-01

    As recommended by the International Working Group on Fast Reactors (IWGFR), the International Atomic Energy Agency organized a specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors. Specialists from all member countries of IWGFR-China, France, Germany, India, Italy, Japan, Russia, the United Kingdom, and the United States-participated in the meeting and made presentations as listed in Table 1. The Commission of European Communities also sent representatives to the meeting. Table 2 contains a list of participants. The meeting consisted of five sessions: (1) an overview, (2) safety characteristics of decay heal removal systems, (3) safely characteristics of reactor protection systems and reactor shutdown systems, (4) safely characteristics of reactor cores, and (5) general discussions antiformulation of recommendations

  12. Investigation of the three-dimensional thermoelastic deformation of the core structure of a fast breeder reactor under stationary working conditions

    International Nuclear Information System (INIS)

    Yong-Su, Hoang.

    1976-12-01

    In this study a method is described which has been developed in order to calculate three-dimensional deformation of the reactor core, taking into account thermal expansion. Two problem areas are of particular importance: 1) The spatial deflection of subassemblies in specified flexible supports and with specified clearances; 2) The investigation of the equilibrium configurations of the subassemblies in the planes of clamping (problem of clamping plane). - The elementary theory of beam deflection has been used to calculate the deformation of subassemblies. However, particular problems have been encountered as a result of flexibly designed support configurations having some spatial clearances. The problem has essentially been solved in two steps: a) Uniqueness analysis of the beam-support configuration; b) Calculation of the support loads and bending line for the unique beam-support configuration. - Basic difficulties currently prevent the problem of clamping plane being solved in a satisfactory manner. Therefore, a simplified clamping model was used for supports without spatial clearance and a parametric study was performed for supports having spatial clearance. The computation method developed is applied to the MARK I core of SNR 300. Core deformations are calculated under different support conditions for the subassemblies in the grid plate and in the upper clamping plane. (orig./HR) [de

  13. Monte-Carlo Modeling of Parameters of a Subcritical Cascade Reactor Based on MSBR and LMFBR Technologies

    CERN Document Server

    Bznuni, S A; Zhamkochyan, V M; Polanski, A; Sosnin, A N; Khudaverdyan, A H

    2001-01-01

    Parameters of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k_{eff}=0.94-0.98), is apable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10^{14} cm^{12}\\cdot s^_{-1}, in the fast booster zone is 5.12\\cdot10^{15} cm^{12}\\cdot s{-1} at k_{eff}=0.98 and proton beam current I=2.1 mA.

  14. Monte-Carlo modeling of parameters of a subcritical cascade reactor based on MSBR and LMFBR technologies

    International Nuclear Information System (INIS)

    Bznuni, S.A.; Zhamkochyan, V.M.; Khudaverdyan, A.G.; Barashenkov, V.S.; Sosnin, A.N.; Polanski, A.

    2001-01-01

    Parameters are investigated of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k eff = 0.94 - 0.98), is capable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10 14 cm 12 · s -1 , in the fast booster zone is 5.12 · 10 15 cm 12 · s -1 at k eff = 0.98 and proton beam current I = 2.1 mA. (author)

  15. The uranium market and its relevance for the fast breeder

    International Nuclear Information System (INIS)

    Marth, W.

    1982-01-01

    Uranium is a very particular element: the technology of light water nuclear power plants is based on its existence, and the doubts with respect to the long-term availability form the basis for the development of fast breeder reactor. At the same time uranium is a merchandise obedient to the laws of the market, which are based on offer and demand, which are however also affected by influences outside the market, particularly in the last time. Here an attempt is made to describe these peculiarities of the uranium market for the past, the presence and the visual future, and to give some insight into the economic field around the breeder reactor, as the uranium market is of great relevance for the introduction of the breeder as an electric power producing system. (orig.) [de

  16. Liquid Metal Fast Breeder Reactors: a bibliography

    International Nuclear Information System (INIS)

    Raleigh, H.D.

    1980-11-01

    This bibliogralphy includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2

  17. Absorber element for fast breeder reactor

    International Nuclear Information System (INIS)

    Verset, L.

    1987-01-01

    This absorber element is characterized by a new head which avoids an accident disconnection of the mobil absorber. This head is made by a superior piece which can take shore up an adjusting ring on an adjusting bearing on the inferior piece. The intermediate piece is catched at the superior piece by a link of chain [fr

  18. Liquid Metal Fast Breeder Reactors: a bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Raleigh, H.D. (ed.)

    1980-11-01

    This bibliography includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  19. Liquid Metal Fast Breeder Reactors: a bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Raleigh, H.D. (ed.)

    1980-11-01

    This bibliogralphy includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  20. Liquid Metal Fast Breeder Reactors: a bibliography

    International Nuclear Information System (INIS)

    Raleigh, H.D.

    1980-11-01

    This bibliography includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2