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Sample records for metal fast breeder

  1. 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.)

  2. 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)

  3. 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.)

  4. Advanced liquid metal fast breeder reactor designs

    International Nuclear Information System (INIS)

    Sayles, C.W.

    1978-01-01

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

  5. 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

  6. 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

  7. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Seed, G.

    1980-01-01

    In a liquid metal cooled nuclear reactor in which the reactor core is submerged in a pool of liquid metal coolant in a primary vessel housed in a concrete vault the core is surrounded by an impermeable barrier bounding an inner or hot region of the pool and an outer or cool region of the pool. The object of the present invention is the provision of a construction in which the complexity of design and manufacture of the barrier for bounding the inner and outer pools of coolant is reduced. (UK)

  8. 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

  9. 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

  10. Alloys for a liquid metal fast breeder reactor

    Science.gov (United States)

    Rowcliffe, Arthur F.; Bleiberg, Melvin L.; Diamond, Sidney; Bajaj, Ram

    1979-01-01

    An essentially gamma-prime precipitation-hardened iron-chromium-nickel alloy has been designed with emphasis on minimum nickel and chromium contents to reduce the swelling tendencies of these alloys when used in liquid metal fast breeder reactors. The precipitation-hardening components have been designed for phase stability and such residual elements as silicon and boron, also have been selected to minimize swelling. Using the properties of these alloys in one design would result in an increased breeding ratio over 20% cold worked stainless steel, a reference material, of 1.239 to 1.310 and a reduced doubling time from 15.8 to 11.4 years. The gross stoichiometry of the alloying composition comprises from about 0.04% to about 0.06% carbon, from about 0.05% to about 1.0% silicon, up to about 0.1% zirconium, up to about 0.5% vanadium, from about 24% to about 31% nickel, from 8% to about 11% chromium, from about 1.7% to about 3.5% titanium, from about 1.0% to about 1.8% aluminum, from about 0.9% to about 3.7% molybdenum, from about 0.04% to about 0.8% boron, and the balance iron with incidental impurities.

  11. Heavy liquid metal cooled fast breeder reactor. Results in 1999

    International Nuclear Information System (INIS)

    Mihara, Takatsugu; Enuma, Yasuhiro; Tanaka, Yoshihiko; Umetsu, Youichirou; Ichimiya, Masakazu

    2000-07-01

    Based on the medium and long-term program of JNC, the feasibility study for fast breeder reactors (FBRs) including related nuclear fuel cycles has been started from the 1999 fiscal year. Various options of FBR plant systems have been studied and a concept of Heavy Liquid Metal cooled FBRs is one of these options. The purpose of this paper is to research and evaluate Heavy Liquid Metal cooled FBRs on the basis of literatures. First, we selected four types of plant concepts listed below. Concept 1: Large-scale pond type reactor with Pb cooled. Concept 2: Large-scale loop type reactor with Pb cooled. Concept 3: Medium-scale module tank type reactor with Pb cooled. Concept 4: Small-scale module tank type reactor with Pb-Bi cooled. Concept 1 and 2 are selected to seek for scale merit on economical aspect. In Concept 3 and 4, we tried to reduce the inventory of HLMC and to ease the load conditions on structures and seek for competitiveness with module effect such as mass production and learning effect. Through a preliminary design study, we identified some technical features of each concept and roughly evaluated economical competitiveness based on total weight of the NSSSs. From this study, we concluded. In general, the large-scale type concepts have little economical advantage because of its huge amount of material needed for its severe load conditions. (Concept 1 and 2). Even for the large-scale pond type reactor, the conclusion seems to be the same. Total amount of the thermal shielding material became huge. Aseismatic structure makes the amount of material increase under the Japanese seismic condition. (Concept 1) For the large-scale loop type reactor, we selected side entry and dual walled piping concept with slide-joint inner wall to cope with thermal expansion of piping system. However, there seemed to be difficulty with compatibility between slide-joint and oxide film corrosion prevention measures. (Concept 2) The medium and small modular type seemed to be

  12. 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

  13. 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

  14. 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

  15. Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

    Science.gov (United States)

    Brehm, Jr., William F.; Colburn, Richard P.

    1982-01-01

    An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

  16. 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

  17. 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

  18. 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

  19. 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.

  20. Assessing the economics of the liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Farmer, A.A.

    1984-01-01

    The main purpose of this paper is to examine the economics of fast reactors but, before doing so, it describes briefly some of their characteristics and states their main attraction, namely to utilize to the maximum the available low-cost uranium resources. This particularly makes fast reactors desirable for nations without large indigenous uranium reserves. Turning to economics, the components that go to make up the cost in a fast reactor, such as capital, fuel fabrication, reprocessing, etc. are considered first. The chapter then deals with the costs of generating electricity from stations taken in isolation (i.e. single station generating costs) and identifies those factors which can help to reduce them to a minimum. Finally, the expenditure of a whole system of thermal and fast reactors is considered over an extended period, where it will be shown that an optimum fast reactor design, based on system costs may differ from one based on single station generating costs. (author)

  1. 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.)

  2. Reprocessing technology of liquid metal cooled fast breeder reactor fuel

    International Nuclear Information System (INIS)

    Baetsle, L.H.; Broothaerts, J.; Heylen, P.R.; Eschrich, H.; Geel, J. van

    1974-11-01

    All the important aspects of LMFBR fuel reprocessing are critically reviewed in this report. Storage and transportation techniques using sodium, inert gas, lead, molten salts and organic coolants are comparatively discussed in connection with cooling time and de-activation techniques. Decladding and fuel disaggregation of UO 2 -PuO 2 fuel are reviewed according to the present state of R and D in the main nuclear powers. Strong emphasis is put on on voloxidation, mechanical pulverization and molten salt disaggregation in connection with volatilization of gaseous fission products. Release of fission gases and the resulting off-gas treatment are discussed in connection with cooling time, burn up and dissagregation techniques. The review is limited to tritium, iodine xenon-krypton and radioactive airborne particulates. Dissolution, solvent extraction and plutonium purification problems specifically connected to LMFBR fuel are reviewed with emphasis on the differences between LWR and fast fuel reprocessing. Finally the categories of wastes produced by reprocessing are analysed according to their origin in the plant and their alpha emitters content. The suitable waste treatment techniques are discussed in connection with the nature of the wastes and the ultimate disposal technique. (author)

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. 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)

  8. Liquid metal fast breeder reactor: decision process and issues. Special report

    International Nuclear Information System (INIS)

    1975-11-01

    The Liquid Metal Fast Breeder Reactor (LMFBR) Program involves many issues of both a policy and a technical nature that need continuous assessment. A suggested structure is presented for national decision making on the timing of the LMFBR Program, taking into consideration the principal uncertainties and variables. The process examines the key issues that affect current decisions. In summary, an analysis of the issues indicates that the key parameters in the decision process are electricity demand, the availability of uranium, and the availability of a clean coal technology. The predictive uncertainties in these three result in eight possible future energy situations. The analysis indicates that six of these outcomes support the acceleration of the LMFBR and two favor slowing the program

  9. 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

  10. 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

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Issa Cherif [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Bose, Tanmoy [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Pekpe, Komi Midzodzi, E-mail: midzodzi.pekpe@univ-lille1.fr [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Cassar, Jean-Philippe [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Mohanty, A.R. [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Paumel, Kévin [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    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.

  13. Mixing requirements for the limiting fuel-coolant interactions in liquid metal fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lenz, Jr, W F

    1976-11-01

    An estimation of the mixing requirements for the limiting fuel-coolant interactions in two specific liquid metal cooled fast reactors, the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR), has been undertaken. The mixing requirements were represented in terms of the limiting mixing time constants. These constants were determined with the Argonne parametric FCI Computer Model for a range of core involvements. Specifically, fuel masses used ranged from as low as one-seventh of the core to a full core involvement. In general, conservative values for additional FCI input parameters were assumed such that the results would be conservative. With the results in hand, several mechanisms were investigated to determine what limiting effects they could have on the mixing rates of the fuel and coolant during an FCI. The energy requirements for mixing were investigated. The results, however, provided no limiting effects. A solidification limited fragmentation model was also investigated. Although this model provided no absolute limiting effects, it did show that fuel particle sizes of a certain size could indeed limit the fuel-coolant mixing rates. Additionally, the limiting effects were found to be much less significant for UC fuel. The third mechanism that was investigated concerned the limiting effects of the finite fuel release rates as a result of TOP accidents in the FFTF. Equivalent mixing time constants based on the fuel release rates were shown to be greater than the limiting values. Thus, this mechanism was shown to be limiting for the particular accident sequence investigated.

  14. A fuel freezing model for liquid-metal fast breeder reactor hypothetical core disruptive accidents

    Energy Technology Data Exchange (ETDEWEB)

    Best, F.R.; Erdman, C.; Wayne, D.

    1985-01-01

    A proposed fuel freezing mechanism for molten UO2 fuel penetrating a steel channel was investigated in the course of liquid-metal-cooled fast breeder reactor hypothetical core disruptiv accident safety studies. The fuel crust deposited on an underlying melting steel wall was analyzed as being subjected to two stresses one due to the pressure difference between the flowing fuel and the stagnant molten steel layer, and the other resulting from the temperature variation through the crust thickness. Analyses based on the proposed freezing mechanism and comparisons with fuel freezing experiments confirmed that fuel freezing occurs in three modes. For initially low steel wall temperatures, the fuel crust was stable and grew to occlude the channel. At high steel wall temperatures (above 1070 K), instantaneous wall melting leading to steel entrainment was calculated to occur with final penetration depending on the refreezing of the entrained steel. Between these two extremes, the stress developed within the crust at the steel melting front exceeds the critical buckling value, the crust ruptures, and steel is injected into the fuel flow. Freezing is dominated by the fuel/steel mixture. The theoretical penetration distances and freezing times were in good agreement with the experimental results with no more than 20% error involved.

  15. 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

  16. Instrumentation for core and coolant monitoring in liquid-metal fast breeder reactors (LMFBR)

    International Nuclear Information System (INIS)

    Hess, B.; Ruppert, E.

    1975-01-01

    The review on core and coolant instrumentation for liquid metal fast breeders aims to give a short survey of measurement methods and the variety of appropriate instrumentation developed and tested for reactor application throughout the world. The introductory part gives a general outline of instrumentation development, partly as the refinement of well-known thermal reactor instrumentation and partly as the special instrumentation demanded for LMFBR safety requirements, some aspects of which are also discussed briefly. The in-core LMFBR instrumentation is surveyed, classifying the measurement or monitoring of coolant properties such as temperature, pressure, flow and acoustic emission and the measurement of core-kinetic quantities such as neutron flux and reactivity. Without considering the fundamentals of the measurements, the state of instrument development is reviewed and, where known, future aspects are indicated. An additional review on fuel failure detection methods and the related instrumentation distinguishes between global or whole-core detection methods and those used for localization of failures. Special attention is paid to the aspect of reactor safety and its reliability as one of the major objectives of these detection methods. A summary of the protective systems and instrumentation already used or foreseen for LMFBR plants forms a transition to a very brief discussion of handling and interpretation of the multitude of data derived from the rather comprehensive LMFBR instrumentation. This state of the art review claims neither to be complete at the time published nor to be a detailed guide to special problems of instrumentation development, the solutions to which are normally part of industrial knowhow. (author)

  17. 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

  18. Thermal-hydraulic characteristics in a liquid-metal fast breeder reactor hot plenum

    International Nuclear Information System (INIS)

    Tanaka, N.; Moriya, S.; Wada, A.

    1984-01-01

    The most important problem in the thermal-hydraulic designs of the pool-type fast breeder reactor is to estimate the thermal conditions affecting the vessel and/or internal structures during both steady and transient operations. The severity of these conditions in the Japanese pool-type reactor, which will be reinforced and equipped with special devices for seismic demands, is apt to be much greater than for other countries. Water tests and thermal-hydraulic analyses have been performed to study such conditions. The effects of the elevations of upper internal structure discharge and intermediate heat exchanger intakes on flow patterns, free surface disturbance, and thermal stratification in the hot plenum have been estimated. From the results of the experiments, suitable elevations could be recommended by comparing some thermal-hydraulic characteristics. The calculations agreed well with the experimental results for the steady-state flow patterns and thermal transients, with the exception of thermal stratification

  19. Fast breeder reactors

    International Nuclear Information System (INIS)

    1978-01-01

    The subject of this invention is a liquid metal cooled nuclear reactor construction in which a concrete pit is lagged to protect it from the heat radiated from the reactor in normal operation but in which the efficiency of the lagging is reduced in case of emergency to allow the excess heat generated by the reactor to be dissipated throughout the pit. The lagging is in two layers, the first covering the internal surface of the pit wall is impermeable to the liquid metal, whilst the second layer over the first is permeable [fr

  20. 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

  1. Carbon transport in a bimetallic sodium loop simulating the intermediate heat transport system of a liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Hampton, L.V.; Spalaris, C.N.; Roy, P.

    1980-04-01

    Carbon transport data from a bimetallic sodium loop simulating the intermediate heat transport system of a Liquid Metal Fast Breeder Reactor are discussed. The results of bulk carbon analyses after 15,000 hours' exposure indicate a pattern of carburization of Type 304 stainless steel foils which is independent of loop sodium temperature. A model based on carbon activity gradients accounting for this behavior is proposed. Data also indicate that carburization of Type 304 stainless steel is a diffusion-controlled process; however, decarburization of the ferritic 2 1/4 Cr-1Mo steel is not. It is proposed that the decarburization of the ferritic steel is controlled by the dissolution of carbides in the steel matrix. The differences in the sodium decarburization behavior of electroslag remelted and vacuum-arc remelted 2 1/4 Cr-1Mo steel are also highlighted

  2. 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

  3. Mass-transfer behavior of materials for liquid-metal fast breeder reactor steam-generators, (1)

    International Nuclear Information System (INIS)

    Matsumoto, Keiji; Ohta, Yoshio; Kataoka, Tadayuki; Suzuki, Katsumi; Yagi, Shigeji

    1976-01-01

    To examine the compatibility of structural materials with liquid sodium in the Liquid-Metal Fast Breeder Reactor (LMFBR) steam generator environments, tests were conducted, in a sodium loop facility, on selected steels mainly of ferrite 2 1/4Cr-1Mo, 2 1/4Cr-1Mo-Nb, high Cr-1Mo, and austenitic AISI type 316. As a result of the above tests, it was found that the decarburization behavior is conditioned by the heat treatment history of steel, and the carbon loss is more rapid in annealed steel than in normalized and tempered steels. Other noteworthy findings were as follows: the carbon concentration immediately beneath the specimen surface increased consistently with the chromium content of steel, and within the range of chromium content of 3 to 5 % weight the decarburization shown by the low-chromium specimens changed to carburization. The tensile strength of annealed 2 1/4Cr-1Mo steel at 500 0 C was markedly reduced by thermal aging rather than by sodium exposure. The influence of sodium exposure on the decrease of tensile strength was slight. The creep rupture strength at 550 0 C was decreased by some 10% by sodium exposure. (auth.)

  4. Development of a methodology for the quantification of the initiating event frequencies in liquid metal fast breeder reactor PSA

    International Nuclear Information System (INIS)

    Yang, Joon Eon; Hahn, Do Hee

    2001-03-01

    One of the major tasks of the Probabilistic Safety Assessment (PSA) is to identify various sequences of events which could lead to the release of radioactivity. In general, each such sequence of events is the result of an initiating event (IE), followed by the failures of other functions or systems designed to mitigate its effects. Thus one of the first requirements of the PSA is to identify the various initiating events requiring analysis, and later to quantify their frequencies. In general, the IEs to be analyzed in PWR/PHWR PSA are well established . The design of Liquid Metal Fast Breeder Reactor (LMFBR) is much different from that of the water reactors, so the IEs to be analyzed in LMFBR PSA also will be different from them of water reactor PSA. However, due to the limited operating and LMFBR PSA experiences, it will be difficult to derive IEs and to quantify their frequencies for LMFBR PSA. Hence, in this report, we describe the methods that can be used to identify the IEs and quantify their frequencies for LMFBR PSA

  5. 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

  6. Advanced breeder cycle uses metallic fuel

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1991-01-01

    Scientists from Argonne National Laboratory have been developing a concept called the Integral fast Reactor (IFR). This fast breeder reactor could effectively increase Uranium resources a hundred fold making nuclear power essentially an inexhaustible energy source. The IFR is outlined. In the IFR, the inherent properties of liquid metal cooling are combined with a new metallic fuel which is allowed to swell and gives an improved burnup level and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics and waste management. (author)

  7. 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.

  8. 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

  9. Thermal-performance study of liquid metal fast breeder reactor insulation

    International Nuclear Information System (INIS)

    Shiu, K.K.

    1980-09-01

    Three types of metallic thermal insulation were investigated analytically and experimentally: multilayer reflective plates, multilayer honeycomb composite, and multilayer screens. Each type was subjected to evacuated and nonevacuated conditions, where thermal measurements were made to determine thermal-physical characteristics. A variation of the separation distance between adjacent reflective plates of multilayer reflective plates and multilayer screen insulation was also experimentally studied to reveal its significance. One configuration of the multilayer screen insulation was further selected to be examined in sodium and sodium oxide environments. The emissivity of Type 304 stainless steel used in comprising the insulation was measured by employing infrared technology. A comprehensive model was developed to describe the different proposed types of thermal insulation. Various modes of heat transfer inherent in each type of insulation were addressed and their relative importance compared. Provision was also made in the model to allow accurate simulation of possible sodium and sodium oxide contamination of the insulation. The thermal-radiation contribution to heat transfer in the temperature range of interest for LMFBR's was found to be moderate, and the suppression of natural convection within the insulation was vital in preserving its insulating properties. Experimental data were compared with the model and other published results. Moreover, the three proposed test samples were assessed and compared under various conditions as viable LMFBR thermal insulations

  10. Thermal-performance study of liquid metal fast breeder reactor insulation

    Energy Technology Data Exchange (ETDEWEB)

    Shiu, Kelvin K.

    1980-09-01

    Three types of metallic thermal insulation were investigated analytically and experimentally: multilayer reflective plates, multilayer honeycomb composite, and multilayer screens. Each type was subjected to evacuated and nonevacuated conditions, where thermal measurements were made to determine thermal-physical characteristics. A variation of the separation distance between adjacent reflective plates of multilayer reflective plates and multilayer screen insulation was also experimentally studied to reveal its significance. One configuration of the multilayer screen insulation was further selected to be examined in sodium and sodium oxide environments. The emissivity of Type 304 stainless steel used in comprising the insulation was measured by employing infrared technology. A comprehensive model was developed to describe the different proposed types of thermal insulation. Various modes of heat transfer inherent in each type of insulation were addressed and their relative importance compared. Provision was also made in the model to allow accurate simulation of possible sodium and sodium oxide contamination of the insulation. The thermal-radiation contribution to heat transfer in the temperature range of interest for LMFBR's was found to be moderate, and the suppression of natural convection within the insulation was vital in preserving its insulating properties. Experimental data were compared with the model and other published results. Moreover, the three proposed test samples were assessed and compared under various conditions as viable LMFBR thermal insulations.

  11. 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,

  12. 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

  13. Development of fast-breeder reactors with high conversion ratios

    International Nuclear Information System (INIS)

    Bobrov, S.B.; Danilychev, A.V.; Eliseev, V.A.

    1983-01-01

    Fast power reactors (breeders) burning oxide fuel have moderate breeding characteristics. The highest conversion ratios can be obtained in breeders using metallic fuel, but unfortunately these are inferior to oxide breeders in such important characteristics as thermal load and burnup. An effective method of building fast reactors with good breeding characteristics is the use of heterogeneous oxide-metal cores. The paper reports on parametric studies of various models of heterogeneous cores. A number of physical aspects of such reactors are analysed: breeding gain, specific doubling time; and safety parameters: sodium vacuum and Doppler effects of reactivity. The paper quotes dependences of these quantities on various parameters which will make it possible to identify the optimum design of an oxide-metal heterogeneous core

  14. Core catcher cooling for a gas-cooled fast breeder

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Schretzmann, K.

    1976-01-01

    Water, molten salts, and liquid metals are under discussion as coolants for the core catcher of a gas-cooled fast breeder. The authors state that there is still no technically mature method of cooling a core melt. However, the investigations carried out so far suggest that there is a solution to this problem. (RW/AK) [de

  15. Trend of development of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, S. (Science and Technology Agency, Tokyo (Japan). Atomic Energy Bureau)

    1982-01-01

    The development of nuclear power is indispensable as the core of the substitute energy for petroleum. It is the urgent subject for world advanced countries to develop fast breeder reactors which can utilize uranium resources efficiently, to breed nuclear fuel resources, and to secure the stable supply of energy for long term in future. In Japan, the development of fast breeder reactors has been advanced independently and efficiently as a national project mainly by the Power Reactor and Nuclear Fuel Development Corp. The experimental reactor ''Joyo'' attained the criticality in April, 1977, and has been operated at the thermal output of 75,000 kW. As for the prototype reactor ''Monju'', the application for the permission to install it was submitted in December, 1980, and now, the safety examination is in progress. The present state of the development of fast breeder reactors in USA, Great Britain, France, West Germany, USSR and Japan is explained. In order to advance fast breeder reactors to the stage of full-scale practical use, a number of the reactors of 1 million kW class including the demonstration reactor will be constructed and operated to demonstrate and learn the technology of power generation plants in practical scale, to improve the performance, and to establish the economical efficiency. The schedule of development, the organization and the sharing of roles, the research and development in the demonstration stage are described.

  16. 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

  17. Creep behaviour of austenitic stainless steels, base and weld metals used in liquid metal fast breeder reactors, during temperature variations

    International Nuclear Information System (INIS)

    Felsen, M.F.

    1982-07-01

    Creep rupture and deformation during temperature variations have been studied for 316 austenitic steel, base and weld metals. Loaded specimens were heated to 900 0 C or 1000 0 C and maintained at this temperature for different durations. The heating rate to these temperatures was between 5 and 50 0 C h -1 , whilst the cooling rate was between 5 and 20 0 C h -1 . The above tests were coupled with short time creep and tensile tests (straining rate 10 -2 h -1 to 10 3 h -1 ) at constant temperature. These tests were used for predicting the creep behaviour of the materials under changing temperature condition. The predictions were in good agreement with the changing temperature and creep experimental results. In addition, a correlation between certains tensile properties, such as the rupture time as a function of stress was observed at high temperature

  18. 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

  19. 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

  20. 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.)

  1. Safety design of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chetal, S.C.; Singh, Om Pal

    2004-01-01

    The basic design and safety design of Prototype Fast Breeder Reactor (PFBR) is presented. Design aspects covered include safety classification, seismic categorization, design basis conditions, design safety limits, core physics, core monitoring, shutdown system, decay heat removal system, protection against sodium leaks and tube leaks in steam generator, plant layout, radiation protection, event analysis, beyond design basis accidents, integrity of primary containment, reactor containment building and design pressure resulting from core disruptive accident. The measures provided in the design represent a robust case of the safety of the reactor. (authors)

  2. 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.)

  3. 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.)

  4. Alternative breeder reactor technologies

    International Nuclear Information System (INIS)

    Spinrad, B.I.

    1978-01-01

    The significance of employing breeder reactors to stretch the world resources of nuclear fuels is briefly discussed, and the various types of breeder concepts are described. General descriptions, advantages, and disadvantages of the liquid metal cooled fast breeder, gas cooled fast breeder, molten salt breeder, thermal breeders, and spectral-shift control reactors are presented. Aspects of safeguarding fissile material connected with breeder operation are examined. 31 references

  5. 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)

  6. 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)

  7. Advanced fuels for fast breeder reactors

    International Nuclear Information System (INIS)

    Potter, P.E.; Spear, K.E.

    1979-01-01

    In this paper we have assessed critically six ternary systems of great significance to the preparation, fabrication and performance of advanced fuels for use in fast breeder nuclear reactors. The systems which have been considered are uranium-carbon-oxygen, plutonium-carbon-oxygen, uranium-carbon-nitrogen, plutonium-carbon-nitrogen, uranium-nitrogen-oxygen and plutonium nitrogen-oxygen. All the systems are characterized by partial or complete solid solutions and a major task of this assessment has been to develop simple models for these solutions which allow consistency between the known thermodynamic and phase equilibria data of the binary systems and the known condensed and gaseous phase equilibria of the ternary systems. Either ideal or regular solution models have been employed to describe the behaviour of the various solutions. (orig.) [de

  8. 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

  9. Reprocessing of spent fuel, Dounreay and fast breeder reactors

    International Nuclear Information System (INIS)

    Lingjaerde, R.

    1986-11-01

    In the light of the public interest in Norway in the breeder reactor fuel reprocessing plant projected in Dounreay, Scotland, the report gives a description of the research center in Dounreay and the planned joint European demonstration facility (EDRP). Certain aspects of the fast breeder reactor are also explained

  10. SIEX: a correlated code for the prediction of liquid metal fast breeder reactor (LMFBR) fuel thermal performance

    International Nuclear Information System (INIS)

    Dutt, D.S.; Baker, R.B.

    1975-06-01

    The SIEX computer program is a steady state heat transfer code developed to provide thermal performance calculations for a mixed-oxide fuel element in a fast neutron environment. Fuel restructuring, fuel-cladding heat conduction and fission gas release are modeled to provide assessment of the temperature. Modeling emphasis has been placed on correlations to measurable quantities from EBR-II irradiation tests and the inclusion of these correlations in a physically based computational scheme. SIEX is completely modular in construction allowing the user options for material properties and correlated models. Required code input is limited to geometric and environmental parameters, with a ''consistent'' set of material properties and correlated models provided by the code. 24 references. (U.S.)

  11. Empirical process modeling in fast breeder reactors

    International Nuclear Information System (INIS)

    Ikonomopoulos, A.; Endou, A.

    1998-01-01

    A non-linear multi-input/single output (MISO) empirical model is introduced for monitoring vital system parameters in a nuclear reactor environment. The proposed methodology employs a scheme of non-parametric smoothing that models the local dynamics of each fitting point individually, as opposed to global modeling techniques--such as multi-layer perceptrons (MLPs)--that attempt to capture the dynamics of the entire design space. The stimulation for employing local models in monitoring rises from one's desire to capture localized idiosyncrasies of the dynamic system utilizing independent estimators. This approach alleviates the effect of negative interference between old and new observations enhancing the model prediction capabilities. Modeling the behavior of any given system comes down to a trade off between variance and bias. The building blocks of the proposed approach are tailored to each data set through two separate, adaptive procedures in order to optimize the bias-variance reconciliation. Hetero-associative schemes of the technique presented exhibit insensitivity to sensor noise and provide the operator with accurate predictions of the actual process signals. A comparison between the local model and MLP prediction capabilities is performed and the results appear in favor of the first method. The data used to demonstrate the potential of local regression have been obtained during two startup periods of the Monju fast breeder reactor (FBR)

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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

  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. 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

  20. DeBeNe Test Facilities for Fast Breeder Development

    International Nuclear Information System (INIS)

    Storz, R.

    1980-10-01

    This report gives an overview and a short description of the test facilities constructed and operated within the collaboration for fast breeder development in Germany, Belgium and the Netherlands. The facilities are grouped into Sodium Loops (Large Facilities and Laboratory Loops), Special Equipment including Hot Cells and Reprocessing, Test Facilities without Sodium, Zero Power Facilities and In-pile Loops including Irradiation Facilities

  1. Overview of pool hydraulic design of Indian prototype fast breeder ...

    Indian Academy of Sciences (India)

    Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations ...

  2. Design of the fast breeder reactor single wall inner vessel

    International Nuclear Information System (INIS)

    Terny, P.; Blaix, J.C.; Del Beccaro, R.

    1987-01-01

    This paper deals with the design of a fast breeder reactor single wall inner vessel. It describes first the studies that led to the optimization of the structure geometry with respect to stability behavior, and subsequently the analyses required to justify the equipment with respect to risks of fatigue, creep and progressive deformation. (orig.)

  3. Thermal analysis of biological shield of fast breeder test reactor

    International Nuclear Information System (INIS)

    Saha, D.; Sarda, V.

    1976-01-01

    A design optimisation of the biological shield of fast breeder test reactor was carried out using computer code HEATING. The effect of different heat sources, variation of coolant tube pitch circle radius, coolant temperature, angular pitch of coolant tubes and thermal conductivity of concrete on the temperature distribution within the shield has been studied. (author)

  4. 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 ...

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

    Indian Academy of Sciences (India)

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of ...

  6. 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

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

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    technology for the energy security of India in the 21st century. Keywords. Fast breeder reactor .... The design of the nuclear fuel is an important aspect which has to be optimised for efficient, economic and safe ..... Recycling of plutonium present in the irradiated fuel with minimum delay is important to ensure rapid growth of ...

  8. 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

  9. Fast Breeder Project. Second quarterly report, 1974

    International Nuclear Information System (INIS)

    Hueper, R.

    Research progress is reported on fuel pin development, material studies and development, corrosion tests and coolant analysis, fuel cycle studies, physics experiments, fast reactor safety, instrumentation development, environmental studies, and sodium technology tests. Much of the work had SNR-300 design applications. (U.S.)

  10. Development of a self-actuated shutdown system for a liquid-metal breeder reactor

    International Nuclear Information System (INIS)

    Cooper, M.H.; Tupper, R.B.; Bernard, A.M.

    1984-01-01

    A comprehensive development program for a Self-Actuated Shutdown System for a Liquid-Metal Fast Breeder Reactor has been completed. The development program included component tests of a temperature sensitive electromagnet, prototype test including the absorber in a sodium loop, accelerated life tests of the high temperature electric coils, and an irradiation test of a miniaturized coil and magnetic circuit materials. The results of these tests have demonstrated that the self-actuated shutdown system is sufficiently developed for application in the next generation Liquid-Metal Fast Breeder Reactor. (author)

  11. TOR: reprocessing of fuel from fast-breeder reactors

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    There have been three stages in the development of the fast-breeder and three stages in fuel reprocessing development. An experimental stage where the cycle was closed with Rapsodie and the fast-breeder fuel processing workshop a La Hague. A demonstration stage -operation TOR- which made it possible to close the Phenix cycle. An industrial prototype stage on an industrial scale which will be achieved with the coming onstream of the Rapid Reprocessing Plant (PURR) (Usine de Retraitement Rapide). The research and construction were carried out in such a way that the move from one stage to another was achieved with the maximum amount of continuity with the benefit of all the experience achieved previously [fr

  12. 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

  13. 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

  14. 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

  15. 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.)

  16. Economic aspects of the fast breeder project

    International Nuclear Information System (INIS)

    Man, R. de; Schenk, H.

    1975-01-01

    A cost-benefit analysis is done for the fast sodium-cooled reactor project (SNR) as compared to power generation by light water-cooled reactors (LWR). The method is comparable with the calculations of the Netherlands Economic Institue (NEI). The conclusions are as follows: 1. Comparison of the costs of LWR and SNR teaches that even in case of a rise in the price of uranium to $ 200/kg and a high estimate of specific investment costs of LWR of fl. 1600/kWc, it will be necessary for the specific investment costs of the SNR to decrease from fl. 6000/kWc to fl. 2000/kWc which, due to scale and learning effects, is not likely to happen before the turn of the century. 2. The cost-benefit analysis, not including external effects (benefits only from power generation) leads to a cost-benefit ratio of 40,15 and 5% for a self developed SNR, assuming that the commercial introduction of the SNR will be possible at once, after 10 and after 20 years respectively. This means that participation in the SNR project should be discontinued on the basis of this analysis. 3. Inclusion of spin-off effects benefits for enterprises (more than 80% of the total benefits as compared to a maximum of 20% from power production according to NEI) also requires the calculation of other external costs and benefits. In the absence of concise data for these costs and benefits, a decision can only be made on the basis of conclusions 1. and 2

  17. Plutonium Worlds. Fast Breeders, Systems Analysis and Computer Simulation in the Age of Hypotheticality

    OpenAIRE

    Sebastian Vehlken

    2014-01-01

    This article examines the media history of one of the hallmark civil nuclear energy programs in Western Germany – the development of Liquid Metal Fast Breeder Reactor (LMFBR) technology. Promoted as a kind of perpetuum mobile of the Atomic Age, the "German Manhattan Project" not only imported big science thinking. In its context, nuclear technology was also put forth as an avantgarde of scientific inquiry, dealing with the most complex and critical technological endeavors. In the face of the ...

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. Plutonium Worlds. Fast Breeders, Systems Analysis and Computer Simulation in the Age of Hypotheticality

    Directory of Open Access Journals (Sweden)

    Sebastian Vehlken

    2014-09-01

    Full Text Available This article examines the media history of one of the hallmark civil nuclear energy programs in Western Germany – the development of Liquid Metal Fast Breeder Reactor (LMFBR technology. Promoted as a kind of perpetuum mobile of the Atomic Age, the "German Manhattan Project" not only imported big science thinking. In its context, nuclear technology was also put forth as an avantgarde of scientific inquiry, dealing with the most complex and critical technological endeavors. In the face of the risks of nuclear technology, German physicist Wolf Häfele thus announced a novel epistemology of "hypotheticality". In a context where traditional experimental engineering strategies became inappropiate, he called for the application of advanced media technologies: Computer Simulations (CS and Systems Analysis (SA generated computerized spaces for the production of knowledge. In the course of the German Fast Breeder program, such methods had a twofold impact. One the one hand, Häfele emphazised – as the "father of the German Fast Breeder" – the utilization of CS for the actual planning and construction of the novel reactor type. On the other, namely as the director of the department of Energy Systems at the International Institute for Applied Systems Analysis (IIASA, Häfele advised SA-based projections of energy consumption. These computerized scenarios provided the rationale for the conception of Fast Breeder programs as viable and necessary alternative energy sources in the first place. By focusing on the role of the involved CS techniques, the paper thus investigates the intertwined systems thinking of nuclear facilities’s planning and construction and the design of large-scale energy consumption and production scenarios in the 1970s and 1980s, as well as their conceptual afterlives in our contemporary era of computer simulation.

  4. 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

  5. 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)

  6. Fast breeder reactor primary pump discharge sphere support device

    International Nuclear Information System (INIS)

    Terny, P.; Blaix, J.C.

    1987-01-01

    In their lower part, the fast breeder reactor primary pump are connected to a sphere-shaped discharge capacity from which the diagrid feedpipes emerge. This discharge capacity is rigidly set on the strongback by means of a device that bears the forces resulting from the bottom effect. This last being generated by the pressure of the sodium discharged from primary pump. When the reactor operating conditions undergo some rapid changes combined to temperature changes, the structures differential expansions result in high stresses in the sphere support as well as in the feedpipes. This paper aims at describing a sphere support which is provided with flexibility under horizontal forces and a higher stiffness under vertical forces for a better adaptation to the various loading conditions. (orig.)

  7. A linear model of the Fast Breeder Test Reactor Plant

    International Nuclear Information System (INIS)

    Kumar, S.S.; Vaidyanathan, G.; Rajakumar, A.

    1979-02-01

    A linear analysis of the Fast Breeder Test Reactor System, consisting of the reactor, intermediate heat exchanger, steam generator and connected piping is presented. The problem of variable boundaries in the steam generator is reduced to a problem of fixed boundaries by dividing the steam generator into six zones. Based upon this, one can obtain the transfer function of any input/output combination. Starting with the time domain non-linear partial differential equations, the problem is reduced to a system of linear equations in complex variables, which can be solved basically by Gaussian elimination process. The results of this work will be useful in determining a suitable control scheme for waterflow in the steam generator and the control parameters. (auth.)

  8. Designing a SCADA system simulator for fast breeder reactor

    Science.gov (United States)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  9. Fluid-structure interaction in fast breeder reactors

    Science.gov (United States)

    Mitra, A. A.; Manik, D. N.; Chellapandi, P. A.

    2004-05-01

    A finite element model for the seismic analysis of a scaled down model of Fast breeder reactor (FBR) main vessel is proposed to be established. The reactor vessel, which is a large shell structure with a relatively thin wall, contains a large volume of sodium coolant. Therefore, the fluid structure interaction effects must be taken into account in the seismic design. As part of studying fluid-structure interaction, the fundamental frequency of vibration of a circular cylindrical shell partially filled with a liquid has been estimated using Rayleigh's method. The bulging and sloshing frequencies of the first four modes of the aforementioned system have been estimated using the Rayleigh-Ritz method. The finite element formulation of the axisymmetric fluid element with Fourier option (required due to seismic loading) is also presented.

  10. 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

  11. 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)

  12. 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

  13. Risk-assessment methodology for fast breeder reactors

    International Nuclear Information System (INIS)

    Ott, K.O.

    1976-04-01

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

  14. 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

  15. 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)

  16. 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.)

  17. Fast breeder cladding tubes provided with helical fins

    International Nuclear Information System (INIS)

    Baumgaertner, E.; Hoffmann, H.; Miller, H.; Jacobi, O.; Bojarski, E.; Freund, D.; Reiser, H.

    1977-04-01

    Development of fast breeder tubes with helical fins started around 1966. The development was initiated mainly by the search for a more favorable alternative to the conventional spacers such as grids and spiral wires. First of all the possibility was investigated of fabricating them on an industrial scale. The first thermohydraulic and fluiddynamic tests were made in parallel by AEG, GfK and Interatom. Besides the possibility of industrial fabrication the problems of non-destructive testing for geometry and material defects had to be treated with priority. In this report the different stages of development are shown and, finally, two rather large projects are described. These projects are the study ready for the licensing procedure concerning the irradiation of finned tube oxide fuel elements in KNK II and the already completed 19-rod fuel element bundle irradiation experiment Mol 7D of GfK performed in the 500 kW sodium loop of the Belgian BR2 reactor. In this BR2 experiment the maximum target burnup of 85,000 MWd/t of oxide was attained without trouble. (orig.) [de

  18. 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

  19. 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)

  20. 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.

  1. 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.

  2. 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

  3. 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

  4. 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)

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. 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

  8. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  9. 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.

  10. 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)

  11. 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

  12. 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)

  13. 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.)

  14. 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

  15. 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

  16. 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

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

    Indian Academy of Sciences (India)

    Keywords. Fast breeder reactors; closed fuel cycle; fuel production and depletion; Pu239 equivalence; multiple recycling; fuel reactivity effects ... With these modifications and also with PHWR Pu as external feed, the study on PFBR fuel recycling is repeated. It is observed that the core-1 initial Pu inventory increases by 3.5% ...

  18. 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.)

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

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

  1. 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

  2. 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)

  3. The history of the construction und operation of the KNK II German Fast Breeder Power Plant

    International Nuclear Information System (INIS)

    Marth, W.

    1993-08-01

    This report describes the German KNK fast breeder project from its beginnings in 1957 until permanent shutdown in 1991. The initial design provided for a sodium-cooled, but thermal reactor. Already during the commissioning of KNK I on the premises of the Karlsruhe Nuclear Research Center modification into a fast nuclear power plant was decided. Considerable difficulties in licensing had to be overcome. KNK II reached high burnup values in the fuel elements and closing of the fuel cycle was achieved. A number of technical problems concerning individual components are described in detail. After the politically motivated discontinuation of the SNR 300 fast breeder project at Kalkar, KNK II was shut down for good in August 1991. (orig.) [de

  4. The History of the Construction and Operation of the German KNK II Fast Breeder Power Plant

    International Nuclear Information System (INIS)

    Marth, W.

    1994-11-01

    The report gives a historical review of the German KNK fast breeder project, from its beginnings in 1957 up to permanent plant shutdown in 1991. The original design was for the sodium cooled thermal reactor KNK I, which was commissioned on the premises of the Karlsruhe Nuclear Research Center. The conversion into a fast nuclear power plant however was a process, which had to overcome considerable licensing difficulties. KNK II attained high fuel element burnups, and the completion of the fuel cycle was achieved. Various technical problems encountered in specific components are described in detail. After the termination of the SNR 300 fast breeder project in Kalkar for political reasons, KNK II was shutdown in August 1991

  5. Space--time dynamics of a fast breeder reactor for localized disturbances

    International Nuclear Information System (INIS)

    Doshi, J.B.

    1976-01-01

    Sodium voiding accidents in Liquid Metal Fast Breeder reactors may cause a positive reactivity insertion due to a lack of balance between two competing effects following the voiding: (1) the shift in neutron energies to higher values which leads to a positive reactivity addition, and (2) the increased neutron diffusion which generates a negative reactivity. Since these two effects have strong spatial dependence, the amount of reactivity added in a voiding accident depends strongly on the position in the reactor core where the voiding accident has occurred. A method of analysis is developed for the localized voiding perturbation based on a flux factorization technique, accounting for the flux shape changes taking place near the region of the voided subassembly. Heat transfer equations are solved in conjunction with the neutronic equations and the two are coupled via feedback effects like the Doppler effect. For the purpose of analysis, a typical large (1000 MWe) LMFBR with two radial core zones of different enrichments is considered. Time-dependent void fractions, which are the input quantities in the analysis, are taken to be typical of a channel flow blockage accident. The transients are analyzed for voidings taking place at three different radial locations in the core and for two different rates of voiding. Results of the calculations show that the voiding transients occurring at the center of the reactor core cause most severe effects in terms of the positive reactivity addition and the effects become progressively less severe for voidings occurring outward toward the periphery of the reactor

  6. 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)

  7. Updated reference design of a liquid metal cooled tandem mirror fusion breeder

    International Nuclear Information System (INIS)

    Berwald, D.H.; Whitley, R.H.; Garner, J.K.

    1985-09-01

    Detailed studies of key techinical issues for liquid metal cooled fusion breeder (fusion-fission hybrid blankets) have been performed during the period 1983-4. Based upon the results of these studies, the 1982 reference liquid metal cooled tandem mirror fusion breeder blanket design was updated and is described. The updated reference blankets provides increased breeding and lower technological risk in comparison with the original reference blanket. In addition to the blanket design revisions, a plant concept, cost, and fuel cycle economics assessment is provided. The fusion breeder continues to promise an economical source of fissile fuel for the indefinite future

  8. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1978-11-01

    As for the experimental fast reactor ''Joyo'', the power increase test has been carried out since April, and the power output was raised stepwise up to 40 MW. The power output, core behavior, plant characteristics as well as shielding integrity were measured at each power level. The examination for licensing the power increase to 75 and 100 MW is still continued by the Committee No. 130. The preparation of various codes required for the core characteristic analysis is in progress. As for the development of the prototype fast reactor ''Monju'', the Construction Preliminary Design (1) was evaluated, and the studies on the specifications of the Construction Preliminary Design (2) are carried out. In respect to the analysis for the Safety Licensing, the analysis of decay heat, the development of an analytical code regarding the rupture propagation in heat transfer tubes for steam generators and others are under way. Technological investigation is carried out to obtain the overseas informations on the safety standards for FBRs and LMFBR technologies. The technical specifications for the preliminary design of the demonstration fast reactor are being prepared. The researches and developments of reactor physics, the structural components of ''Joyo'' and ''monju'', instrumentation and control, sodium technology, fuel materials, structural materials, safety and steam generators are reported, respectively. (Kako, I.)

  9. 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.)

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. Fast breeder reactor kinetics. An inverse problem - 135

    International Nuclear Information System (INIS)

    Seleznev, E.F.; Belov, A.A.; Mushkaterov, A.A.; Matveenko, I.P.; Zhukov, A.M.; Raskatch, K.F.

    2010-01-01

    An analysis of solution of the spatial reactor kinetics neutron-transfer equation using a fast reactor as an example is presented in this paper. To solve the spatial reactor kinetics equation in three-dimensional geometry in multi-energy group-diffusion approximation, a program calculated module TIMER was created. The number of groups of delayed neutron precursor concentrations varies from 6 to 8. When solving a transient neutron-transfer problem, two specific tasks are distinguished. The first of them is a direct problem wherein the neutron flux density and its derivatives such as reactor power etc. are determined at each time step. The second (inverse) problem exists for the point-kinetics equation where the reactor reactivity is calculated using the known dependence of reactor power on time. The paper presented focuses on solution of the inverse problem using experiment calculations for BFS-105 critical assembly. (authors)

  16. Nuclear data needs for fast breeder reactor shielding

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  17. 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

  18. 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)

  19. 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.)

  20. 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

  1. 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)

  2. Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

    International Nuclear Information System (INIS)

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

    1990-01-01

    Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  4. 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

  5. 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

  6. 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

  7. 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

  8. The Kalkar fast breeder: Decision of the Bundestag concerning start-up

    International Nuclear Information System (INIS)

    1983-01-01

    This survey informs on declarations and recommendations of the official inquiry commission ''Future Nuclear Energy Policies'' of the 9th Bundestag concerning the question whether or not commissioning of the fast breeder prototype plant SNR 300 in Kalkar is politically justifiable as well as on how the commission's recommendations were discussed in Parliament. The commission recommended by majority of votes to commission the SNR 300. The research and technology commission in charge accepted this recommendation by majority of votes. On December 3, 1982, the Bundestag voted in favour of the recommendation by majority of votes. Thus the political reservations of the Bundestag against commissioning the SNR-300 are revoked. (orig./HSCH) [de

  9. 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

  10. 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

  11. 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

  12. The SNR 300 fast breeder in the ups and downs of its history

    International Nuclear Information System (INIS)

    Marth, W.

    1994-12-01

    The Fast Breeder Project was founded in Karlsruhe in 1960. After an initial period of fundamental research, industry assumed responsibility for designing the SNR 300. Construction of the Kalkar Nuclear Power Station was hampered by a variety of political influences, but finally completed in 1985. As a consequence of the North Rhine-Westphalian party-in-government's opting out of nuclear power, no startup permit was issued for the SNR 300. Consequently, the Kalkar Nuclear Power Station project was discontinued for political reasons in March 1991. The report is the English translation of KFK--4466. (orig./HP) [de

  13. Control rod calibration methods for fast breeder reactors applied to Phenix

    International Nuclear Information System (INIS)

    Lecourt, G.

    1998-01-01

    The control and the emergency shutdown of a fast breeder reactor depends essentially on control rods. For this reason, it is imperative to know exactly how much anti reactivity is introduced with the rods in the reactor core. Different methods have been compared in order to see if they are compatible with Phenix reactor. Their limits have been studied. The shadow and anti shadow effects that can the rods make one to the other and then their effective weight of the rods screen have been clarified. (N.C.)

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

    Science.gov (United States)

    Honekamp, John R.; Fryer, Richard M.

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

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

    International Nuclear Information System (INIS)

    Ishiguro, Y.

    1983-01-01

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

  16. U.S. and foreign breeder reactors

    International Nuclear Information System (INIS)

    Hill, E.H.

    1977-01-01

    The running battle between Congress and the Administration over the Clinch River Breeder Reactor Plant (CRBRP) Project has provoked an increased interest in domestic and foreign breeder reactor programs. Perhaps an understanding of the history of breeders here and abroad will serve to place the CRBRP in perspective and allow some analysis of how the U.S. appears on the global canvas. Breeder reactor technology has, for the most part, settled down to concentration on the liquid metal fast breeder reactor (LMFBR). This is the result of 32 years of experience with reactors employing a fast neutron flux and even longer experience with liquid metal coolants. However, a number of U.S. utilities are sponsoring a gas cooled fast reactor program as an alternative technology to the LMFBR. This development program is supported by the U.S. Department of Energy

  17. Comparison of lithium and the eutectic lead lithium alloy, two candidate liquid metal breeder materials for self-cooled blankets

    International Nuclear Information System (INIS)

    Malang, S.; Mattas, R.

    1994-06-01

    Liquid metals are attractive candidates for both near-term and long-term fusion applications. The subjects of this comparison are the differences between the two candidate liquid metal breeder materials Li and LiPb for use in breeding blankets in the areas of neutronics, magnetohydrodynamics, tritium control, compatibility with structural materials, heat extraction system, safety, and required R ampersand D program. Both candidates appear to be promising for use in self-cooled breeding blankets which have inherent simplicity with the liquid metal serving as both breeders and coolant. The remaining feasibility question for both breeder materials is the electrical insulation between liquid metal and duct walls. Different ceramic coatings are required for the two breeders, and their crucial issues, namely self-healing of insulator cracks and radiation induced electrical degradation are not yet demonstrated. Each liquid metal breeder has advantages and concerns associated with it, and further development is needed to resolve these concerns

  18. 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

  19. 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)

  20. 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.)

  1. 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

  2. Test run of 50 MW steam generator for prototype fast breeder reactor of PNC

    International Nuclear Information System (INIS)

    Tanaka, Kenji

    1974-01-01

    A sodium heated steam generator is one of the most important machinery in a fast reactor plant, influencing the aspects of economy and reliability. In the Power Reactor and Nuclear Fuel Development Corporation (PNC), the 50 MW steam generator and associated testing facility have been developed, which lead to the demonstration of the steam generator and the establishment of plant control technology for the fast breeder reactor rackled by PNC. The steam generator is separate helical-coil type with an evaporator, a superheater and a reheater. Taking two to three years in construction, in June 1974, the steam generator with the testing facility was successfully operated at rated power of 50 MW continuously for 72 hours. The steam generator planned for the prototype FBR ''Monju'' is first outlined, together with its devlopment program. The 50 MW steam generator and the testing facility are then described, including construction and operation; finally, the results of test and future test plans are explained. (Mori, K.)

  3. 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.)

  4. Economic analysis of fusion breeders

    International Nuclear Information System (INIS)

    Delene, J.G.

    1985-01-01

    This paper presents a study of the economic performance of Fission/Fusion Hybrid devices. This work takes fusion breeder cost estimates and applies methodology and cost factors used in the fission reactor programs to compare fusion breeders with Liquid Metal Fast Breeder Reactors (LMFBR). The results of the analysis indicate that the Hybrid will be in the same competitive range as proposed LMFBRs and have the potential to provide economically competitive power in a future of rising uranium prices. The sensitivity of the results to variations in key parameters is included

  5. 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

  6. 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

  7. IAMBUS, a computer code for the design and performance prediction of fast breeder fuel rods

    International Nuclear Information System (INIS)

    Toebbe, H.

    1990-05-01

    IAMBUS is a computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fast breeder fuel rods. The code deals with steady, non-steady and transient operating conditions and enables to predict in-pile behavior of fuel rods in power reactors as well as in experimental rigs. Great effort went into the development of a realistic account of non-steady fuel rod operating conditions. The main emphasis is placed on characterizing the mechanical interaction taking place between the cladding tube and the fuel as a result of contact pressure and friction forces, with due consideration of axial and radial crack configuration within the fuel as well as the gradual transition at the elastic/plastic interface in respect to fuel behavior. IAMBUS can be readily adapted to various fuel and cladding materials. The specific models and material correlations of the reference version deal with the actual in-pile behavior and physical properties of the KNK II and SNR 300 related fuel rod design, confirmed by comparison of the fuel performance model with post-irradiation data. The comparison comprises steady, non-steady and transient irradiation experiments within the German/Belgian fuel rod irradiation program. The code is further validated by comparison of model predictions with post-irradiation data of standard fuel and breeder rods of Phenix and PFR as well as selected LWR fuel rods in non-steady operating conditions

  8. Fission gas pressure build-up and fast-breeder economy

    International Nuclear Information System (INIS)

    Engelmann, P.

    1962-01-01

    Fuel-cycle costs and doubling time of fast-breeder reactors are strongly affected by the fuel-burn-up obtainable. Use of oxide or carbide fuel offers the possibility of reaching a burn-up of 100 000 MWd/t. In fuel-clad elements, a limiting factor is the fission-gas-pressure build-up. At the high burn-up considered, an appreciable fraction of the fission gases gets into the pores and thus contributes to the pressure on the can. Starting from the known fission-product yields and decay chains, gas production and pressure build-up have been calculated. Three physical models have been employed in calculating the pressure acting upon the can : the gas is contained either in interconnected pores, in separate pores, or in a central hole. The pressure-dependence upon free volume (fuel density) and temperature will be discussed. Cans made of high-strength materials as Ineonel-X and molybdenum could stand the fission-gas pressure at operating temperatures. Unfortunately, these materials have higher absorption cross-sections than stainless steel. Results of a multi-group calculation are given, showing the effect of using these can materials and of decreasing the fuel density on critical mass and breeding ratio in small and medium-size breeders. (author) [fr

  9. 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

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

    International Nuclear Information System (INIS)

    Sinha, N.K.; Raj, Baldev

    2015-01-01

    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

  11. Fast Breeder Blanket Facility. Quarterly progress report, April 1, 1978--June 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Clikeman, F. M. [ed.

    1978-06-01

    The work performed in the reporting period was primarily concerned with the initial measurements and checkout of the Purdue Fast Breeder Blanket Facility (TASK C). The safety calculations for the ''as built'' are also presented and compared with the initial safety calculations (TASK B). Safety calculations on the FBBR facility have been completed for a variety of new flooded-dry combinations. Specifically the effect of adding boron carbide to the transformer regions was investigated. The investigation showed the stabilizing effect of the boron carbide in the flooded cases while the effect in the dry cases is minimal. Measurements of the k/sub eff/ of the FBBF facility, of the effectiveness of the shielding, and of the radiation levels in the rooms surrounding the laboratory have been completed and compare well with the calculations presented in the license application.

  12. 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)

  13. 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.

  14. 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)

  15. 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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  17. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. 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

  20. 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

  1. 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

  2. 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

  3. Design optimization of backup seal for sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Sinha, N.K.; Ghosh, P.; Saha, A.; Mukhopadhyay, R.; Raj, Baldev; Chetal, S.C.

    2012-01-01

    Highlights: ► Design arrived from fourteen geometric options by finite element analysis. ► Seal geometry, size, compression, contact pressure, stress and compression load optimized. ► Effects of reduced fluoroelastomer strength at 110 °C, strain rate and stress-softening incorporated. ► Ageing, friction, tolerances, batch-to-batch/production variations in fluoroelastomer considered. ► Procedure applicable to other elastomeric seals of Fast Breeder Reactors. -- Abstract: Design optimization of static, fluoroelastomer backup seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) is depicted. 14 geometric variations of a solid trapezoidal cross-section were studied by finite element analysis (FEA) to arrive at a design with hollowness and double o-ring contours on the sealing face. The seal design with squeeze of 5 mm assures failsafe operation for at least 10 years under a differential pressure of 25 kPa and ageing influences of fluid (air), temperature (110 °C) and γ radiation (23 mGy/h) in reactor. Hybrid elements of 1 mm length, regular integration, Mooney–Rivlin material model and Poisson’s ratio of 0.493 were used in axisymmetric analysis scheme. Possible effects of reduced fluoroelastomer strength at 110 °C, ageing, friction, tolerances in reactor scale, testing conditions during FEA data generation and batch-to-batch/production variations in seal material were considered to ensure adequate safety margin at the end of design life. The safety margin and numerical prediction accuracy could be improved further by using properties of specimens extracted from seal. The approach is applicable to other low pressure, moderate temperature elastomeric sealing applications of PFBR, mostly operating under maximum strain of 50%.

  4. 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)

  5. 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)

  6. Comparison of oxide- and metal-core behavior during CRBRP [Clinch River Breeder Reactor Plant] station blackout

    International Nuclear Information System (INIS)

    Polkinghorne, S.T.; Atkinson, S.A.

    1986-01-01

    A resurrected concept that could significantly improve the inherently safe response of Liquid-Metal cooled Reactors (LMRs) during severe undercooling transients is the use of metallic fuel. Analytical studies have been reported on for the transient behavior of metal-fuel cores in innovative, inherently safe LMR designs. This paper reports on an analysis done, instead, for the Clinch River Breeder Reactor Plant (CRBRP) design with the only innovative change being the incorporation of a metal-fuel core. The SSC-L code was used to simulate a protected station blackout accident in the CRBRP with a 943 MWt Integral Fast Reactor (IFR) metal-fuel core. The results, compared with those for the oxide-fueled CRBRP, show that the margin to boiling is greater for the IFR core. However, the cooldown transient is more severe due to the faster thermal response time of metallic fuel. Some additional calculations to assess possible LMR design improvements (reduced primary system pressure losses, extended flow coastdown) are also discussed. 8 refs., 13 figs., 2 tabs

  7. 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

  8. 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

  9. 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

  10. 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

  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. Liquid Metal Fast Breeder Reactor Program: Argonne facilities

    Energy Technology Data Exchange (ETDEWEB)

    Stephens, S. V. [comp.

    1976-09-01

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

  13. Liquid Metal Fast Breeder Reactor program. Volume IV. Environmental statement

    International Nuclear Information System (INIS)

    1974-12-01

    A broad overview is presented of the many implications of LMFBR program implementation, up to and encompassing a fully developed LMFBR power plant economy, including the secondary impacts, the unavoidable adverse environmental impacts, cumulative environmental impacts, and cost-benefit analyses, and alternative energy strategies. Under the heading of secondary impacts, the national implications of the availability of electricity from LMFBRs, and the specific economic impacts of the LMFBR program are examined. The currently feasible alternatives and potential future alternatives for mitigating adverse environmental impacts of the LMFBR fuel cycle are described. The problems of safeguarding special nuclear material from potential diversion to unauthorized purposes are analyzed. The cumulative environmental effects of LMFBR operation to the Year 2020, the decommissioning of LMFBRs and fuel cycle facilities upon the completion of their useful life, the irreversible and irretrievable commitments of resources that will accompany implementation of an LMFBR economy, and an analysis of the costs and benefits of implementing the LMFBR Program are included. (U.S.)

  14. Breeders: operational experience with fast power reactors in five states - more intensive German-French breeder cooperation

    International Nuclear Information System (INIS)

    Hueper, R.

    1978-01-01

    In the past year contracts have been signed and implemented for German-French cooperation in LMFBR development and commercialization. - The first German nuclear power station with a sodium cooled fast reactor, KNK II in Karlsruhe, is going into operation. - Construction of the prototype SNR 300 at Kalkar (Lower Rhine) is slowing down awaiting a decision of the German Federal Constitutional Court. - On the international level, remarkable experience in the operation of fast power reactors has accumulated. - Possible fuel cycle alternatives are being evaluated by an international committee. (orig.) [de

  15. 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)

  16. 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

  17. Structure and experimental program for SNEAK 12 - an assembly for fast breeder safety experiments

    International Nuclear Information System (INIS)

    Helm, F.

    1979-01-01

    The critical assembly SNEAK 12 and its foreseen experimental program have the main purpose to check the validity of neutron physics calculational methods used in the analysis of accidental situations of fast breeder reactors. In the investigation of accidental courses configurations have to be considered, which are caused by assembly deformation and meltdown and which are characterized by irregular fuel and structural material arrangements with cavities and empty channels. The reactivity differences between the unperturbed core and a series of perturbed configurations have to be determined. The individual configurations have to be chosen in such a way, that the calculational methods for the different aspects of the accident sequence (formation of cavities and channels, relocation of fissile and fertile material and steel) can be tested one by one. Two different cores are foreseen: SNEAK 12A as a one-zone core with enriched uranium fuel and SNEAK 12B with a central test zone with plutonium-uranium mixed-oxide fuel surrounded by a driver zone of enriched uranium. The report describes these cores and their assemblies, and the experimental program is outlined

  18. 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%

  19. 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)

  20. Design features of installed radiation monitoring system in Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Sivasailanathan, Vidhya; Ananth, Allu; Sarangapani, R.; Jose, M.T.

    2016-01-01

    The Prototype Fast Breeder Reactor (PFBR), the first of its kind reactor in our country, which is presently under construction at Kalpakkam is a pool type reactor which uses mixed oxide fuel (UO 2 +PuO 2 ) and sodium as coolant. In nuclear facilities, it is mandatory to monitor the radiological status of all active areas on continuous basis to control radiation exposure and contamination. The installed Radiation Monitoring System (RMS) of PFBR is designed to meet radiological protection and reactor safety requirements. RMS consists of both online and offline monitors. The signals from the online monitors are available in Radiation Protection Panel at Main Control Room (MCR). The online monitors provide information to control room personnel to initiate further actions whenever the radiation levels exceed the alarm thresholds. The radiation levels around the zone-1 areas of the plant and specific activity of Cat II and Cat III liquid wastes sent through pipelines are also available at Main Control Room. The system helps in minimizing the collective dose and ensures that the discharges from the plant are well below the technical specification limits. RMS also provides information to health physicists during the clearance of radiological work permits and is a very useful tool to review the trend, time of escalation of radiation field, etc., of radiation levels in case of any incident in the plant. The paper brings out the design features of RMS of PFBR

  1. C-scope under-sodium viewer for sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Miyazawa, Tatsuo; Uesugi, Nobuo; Iguchi, Tatsuo; Taguchi, Junzo; Takagi, Nobuyuki

    1976-01-01

    A C-scope under-sodium viewer has been developed for monitoring the interior of sodium-cooled fast breeder reactors. Consisting of a transducer that emits and receives ultrasonic waves under liquid sodium, a mechanism that drives the transducer under liquid sodium and an image displaying section, it inspects the fuel assembly through its image in optically opaque high-temperature (300 0 C) liquid sodium. The results of its evaluation test are: (1) The transducer could continue satisfactory operation under 350 0 C (at the highest) sodium for more than a month. (2) The driving mechanism, though it was the first of the kind appearing in Japan, has been proved that it could continue operation for a week under 300 0 C sodium. (3) The image displaying section, in spite of the low speed of the transducer (below 20 rpm), could display stable and clear images. (4) The image in 300 0 C was as clear as that in room-temperature water. (auth.)

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

    International Nuclear Information System (INIS)

    Sinha, N.K.; Raj, Baldev

    2014-01-01

    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 −3 /10 −2 scc/s/m length of seal) and argon inflation gas (10 −3 scc/s/m length of seal) is described

  3. A dynamic model for the steam generator of the Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Kumar, S.S.; Kothaandaraman, A.L.; Ravi Prasan, G.R.

    1978-11-01

    A computer code for the thermal-hydraulic dynamics of the steam generator of the Fast Breeder Test Reactor Plant has been written. This report describes the physical, mathematical and computational models used. Some results are listed which bring out the usefulness of this code as the prerequisite for the estimation of thermal stresses and determination of safe operating modes. Problems of slip and non-uniform radial distribution are neglected because of the high pressure of operation and low pressure drop. By suitable decoupling of some of the equations of the model, the computational effort is reduced to the solution of a system of linear equations or one non-linear equation. The non-linear equation is solved by the method of false position and the method of binary division. Results are listed for the perturbations in sodium temperature, sodium flow, water flow and steam pressure. The computer code is written in Fortran IV and is operational on IBM 370/155 system. (auth.)

  4. 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.

  5. 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)

  6. 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)

  7. 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)

  8. Advanced methods comparisons of reaction rates in the Purdue Fast Breeder Blanket Facility

    International Nuclear Information System (INIS)

    Hill, R.N.; Ott, K.O.

    1988-01-01

    A review of worldwide results revealed that reaction rates in the blanket region are generally underpredicted with the discrepancy increasing with penetration; however, these results vary widely. Experiments in the large uniform Purdue Fast Breeder Blanket Facility (FBBF) blanket yield an accurate quantification of this discrepancy. Using standard production code methods (diffusion theory with 50 group cross sections), a consistent Calculated/Experimental (C/E) drop-off was observed for various reaction rates. A 50% increase in the calculated results at the outer edge of the blanket is necessary for agreement with experiments. The usefulness of refined group constant generation utilizing specialized weighting spectra and transport theory methods in correcting this discrepancy was analyzed. Refined group constants reduce the discrepancy to half that observed using the standard method. The surprising result was that transport methods had no effect on the blanket deviations; thus, transport theory considerations do not constitute or even contribute to an explanation of the blanket discrepancies. The residual blanket C/E drop-off (about half the standard drop-off) using advanced methods must be caused by some approximations which are applied in all current methods. 27 refs., 3 figs., 1 tab

  9. 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)

  10. 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.

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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.)

  17. 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

  18. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. A review. Pt. I. Key areas

    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 studies 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 ready-made supply. Part I addresses key areas of design shortlisting, strategy, development and unification with a backdrop of international evolution. (orig.)

  19. 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

  20. 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

  1. Carter's breeder policy has failed, claims Westinghouse manager

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    Nuclear nations developing liquid metal fast breeder reactor (LMFBR) technology have not been dissuaded by President Carter's efforts to stop the breeder program as a way to control the proliferation of nuclear weapons. There is no evidence that Carter's policy of moral persuasion has had any impact on their efforts. A review of the eight leading countries cites their extensive progress in the areas of breeder technology and fuel reprocessing, while the US has made only slight gains. The Fast Flux Test Facility at Hanford is near completion, but the Clinch River project has been slowed to a minimum

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  3. 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

  4. List of reports and conference papers on the Fast Breeder Project published by the Nuclear Research Center Karlsruhe from 1972 to 1975

    International Nuclear Information System (INIS)

    Hueper, R.

    1976-12-01

    This list is a continuation of the list of publications contained in the final report on work completed by GfK during Phase IIa of the Fast Breeder Project (KFK-Ext. 25/72-1). In addition, short accounts of the results of the research work done in the framework of the Fast Breeder Project of GfK have been published in the quarterly reports KFK 1972/1 through 1975/4. These reports have not been included in the present list. (orig.) [de

  5. Fast and efficient charge breeding of the Californium rare isotope breeder upgrade electron beam ion source

    International Nuclear Information System (INIS)

    Ostroumov, P. N.; Barcikowski, A.; Dickerson, C. A.; Perry, A.; Sharamentov, S. I.; Vondrasek, R. C.; Zinkann, G. P.; Pikin, A. I.

    2015-01-01

    The Electron Beam Ion Source (EBIS), developed to breed Californium Rare Isotope Breeder Upgrade (CARIBU) radioactive beams at Argonne Tandem Linac Accelerator System (ATLAS), is being tested off-line. A unique property of the EBIS is a combination of short breeding times, high repetition rates, and a large acceptance. Overall, we have implemented many innovative features during the design and construction of the CARIBU EBIS as compared to the existing EBIS breeders. The off-line charge breeding tests are being performed using a surface ionization source that produces singly charged cesium ions. The main goal of the off-line commissioning is to demonstrate stable operation of the EBIS at a 10 Hz repetition rate and a breeding efficiency into single charge state higher than 15%. These goals have been successfully achieved and exceeded. We have measured (20% ± 0.7%) breeding efficiency into the single charge state of 28+ cesium ions with the breeding time of 28 ms. In general, the current CARIBU EBIS operational parameters can provide charge breeding of any ions in the full mass range of periodic table with high efficiency, short breeding times, and sufficiently low charge-to-mass ratio, 1/6.3 for the heaviest masses, for further acceleration in ATLAS. In this paper, we discuss the parameters of the EBIS and the charge breeding results in a pulsed injection mode with repetition rates up to 10 Hz

  6. 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)

  7. Role of chemistry in analysis on the leakage incident of sodium from the fast breeder reactor of 'Monju'

    International Nuclear Information System (INIS)

    Takahashi, Masao

    2008-01-01

    The chemical reactions of Na-Fe-O 2 -H 2 O system and their equilibrium constants (600-1200 K) were expressed in terms of standard equilibrium O 2 pressure P 0 and standard potential E 0 . The results were graphically illustrated on the equilibrium diagrams of log P (O2) (O 2 partial pressure) vs. log (Na 2 O) (basicity) or E(potential) vs. log(O 2- ) (basicity). These diagrams visually and comprehensively show the equilibrium relations in the reactions among molten Na with O 2 , H 2 O and Fe with their reaction products and give a useful information for analysis of the leakage incident of sodium from the fast breeder reactor of 'Monju'. (author)

  8. 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)

  9. 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

  10. 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)

  11. Liquid metal reactor deactivation as applied to the experimental breeder reactor - II

    International Nuclear Information System (INIS)

    Earle, O. K.; Michelbacher, J. A.; Pfannenstiel, D. F.; Wells, P. B.

    1999-01-01

    The Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-W) was shutdown in September, 1994. This sodium cooled reactor had been in service since 1964, and by the US Department of Energy (DOE) mandate, was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility (SPF) was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the SPF

  12. An estimate of the radiological consequence of notional accidental releases of radioactivity from a fast breeder reactor

    International Nuclear Information System (INIS)

    Kelly, G.N.; Jones, J.A.; Hunt, B.W.

    1977-08-01

    In this report an assessment is made of the radiological consequences of notional accidental releases of activity from a fast breeder reactor under certain circumstances. It was prepared under contract to the Nuclear Installations Inspectorate (Health and Safety Executive) to assist them in making a preliminary safety assessment of fast reactors. The range of releases considered in the report was specified by the Inspectorate and comprises the vaporisation and release of varying fractions of the core of a 1300 MW(e) reactor. Two cases are evaluated depending on assumptions relating to the remainder of the core. No attempt is made to assign any probability to the occurrence of a given release; the report provides no more than a part of the information necessary for a safety assessment and is to be considered only within this limited context. The subject is dealt with under the following headings: introduction; parameters used in the assessment; atmospheric dispersion; pathways of exposure and dosimetric models; doses associated with the release of one tonne of fuel; biological effects; consequences of releases in which the nuclide composition differs from that in the fuel. (U.K.)

  13. Binary breeder reactor with annular core

    International Nuclear Information System (INIS)

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

    1988-01-01

    Characteristics of a 1200 MWe binary breeder reactor with annular core fueled with metallic 233 U- 238 U-Zr, Pu- 238 U-Zr and Th in the blankets have been analyzed. The Doppler effect is small as expected in a metal fueled fast reactor. The sodium void reactivity is, in general, smaller than in homogeneous fast reactors fueled with metallic fuel and with 1 m core height. The worths of available control is high and there is a large shutdown margin throughout the operational cycle. There are flexibility in blankets fueling in the two cycles, uranium and thorium, with doubling times of about 20 years. (author) [pt

  14. 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

  15. Preliminary study on nano- and micro-composite sol-gel based alumina coatings on structural components of lead-bismuth eutectic cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Dou, Peng; Kasada, Ryuta

    2011-01-01

    In order to protect the structural components of lead-bismuth eutectic cooled fast breeder reactors from liquid metal corrosion, Al 2 O 3 nano- and micro-composite coatings were developed using an improved sol-gel process, which includes dipping specimens in a sol-gel solution dispersed with fine α-Al 2 O 3 powders prepared by mechanical milling. Accelerated corrosion tests were conducted on coated specimens in liquid lead-bismuth eutectic at 500 o C under dynamic conditions. Scanning electron microscopy (SEM) and X-ray diffraction (XRD) analyses revealed that the coatings are composed of α-Al 2 O 3 and they are about 10 μm thick. After the corrosion tests, no spallation occurred on the coatings, and neither Pb nor Bi penetrated into the coatings, which indicates that the coatings possess an enhanced dynamic LBE corrosion resistance to lead-bismuth eutectic corrosion. The nano-structured composite particles integrated into the coatings play an important role in achieving such superior lead-bismuth eutectic corrosion resistance.

  16. 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

  17. 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

  18. Evaluation of molten fuel containment concepts for gas-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Kang, C.S.; Torri, A.

    1979-10-01

    Four in-vessel molten fuel containment concepts for the GCFR were compared, namely, (1) a ceramic crucible, (2) a borax bath, (3) a heavy metal bath, and (4) a steel bath. The ceramic crucible is the simplest but depends on substantial upward heat removal. The borax bath and the heavy metal bath concepts offer better performance but would require design changes and an increased experimental effort. The steel bath concept is a good compromise and has potential for further improvement by combining it with the essential features of other concepts, i.e., the crucible or the heavy metal bath. It is concluded that several concepts could potentially exploit the normally provided cooled liner barrier in the PCRV cavity for post-accident fuel containment

  19. 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)

  20. 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

  1. 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.)

  2. A preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

    Directory of Open Access Journals (Sweden)

    Tohru Suzuki

    2015-04-01

    Full Text Available In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss of flow (ULOF, which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of in-vessel retention for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of in-vessel retention against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.

  3. Application of a fault-tolerant microprocessor-based core-surveillance system in a German fast breeder reactor

    International Nuclear Information System (INIS)

    Voges, U.

    1986-01-01

    For the fast breeder reactor KNK II at Karlsruhe, Germany, a microprocessor-based safety shut-down system is built. Analogue to the triple modular instrumentation it consists of TMR hardware. Functionally it is split into four blocks which operate in cascade-like fashion. The main functions are mean value calculation, current limit control, trend control, and final evaluation. In order to secure correctness, several constructive and analytical methods are applied for fault avoidance, like formal specification languages, programming guidelines, software quality assurance plan, validation, verification, and testing. Since additional means for correct and safe operation are still necessary, fault-tolerance and error-detection techniques are applied. These include self-checking programs, plausibility checks, control data, information exchange and control between the redundancies, and especially diversity. This diversity refers to different teams for the different development phases as well as to different tools and environments, like different programming languages for the application software. Three separate but functional identical programs will be implemented in Iftran, Pascal and PL/M. These will not only be used during the extensive testing period, but also during final operation

  4. 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)

  5. 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

  6. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO2-PuO2 fuel

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Elbel, H.

    1977-01-01

    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

  7. 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)

  8. 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

  9. 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

  10. Life forecasting of welded elements of fast breeder reactor according to evaluation of creep damage

    Energy Technology Data Exchange (ETDEWEB)

    Tanabe, Tatsuhiko; Yamazaki, Masayoshi; Hongo, Hiromichi; Watabe, Takashi; Kinugawa, Junichi; Nakazawa, Takanori [National Research Inst. for Metals, Tsukuba, Ibaraki (Japan)

    2000-02-01

    This study aims at precision upgrading of construction soundness evaluation at abnormal and accidental times, foundation establishment on life forecasting of welded elements at middle and later periods of the FBR life, and contribution to upgrading of its safety, by quantifying time dependent type creep damage accumulated at welded portions in reactor vessel or primary pipings under steady and unsteady operation conditions of the FBR. Therefore, on an object of welded joints of austenitic stainless steel (304 and 306FR) and ferritic heat resistant steel (2.25Cr-1Mo and improved 9Cr-1Mo) constituting of the primary system in FBR, the following three theme were carried out: (1) development on a creep strain equation using macroscopic continuous body dynamics, (2) development on detection and quantification technologies of microscopic creep damage, and (3) development on simulation of the creep damage. In 1998 fiscal year, together with detail investigation on creep deformation behavior of the mother materials for the 340HP steel thick plate welded joint, elucidation on creep fracture of the 2.25Cr-1Mo steel welded joint was carried out. And, relationship between creep damage process and supersonic property of the 304 steel was investigated at a viewpoint of metallic texture and transition density. In addition, in order to conduct creep damage simulation more precisely, local change on creep strain of the 316FR steel multi-layer welded joint was tried to observe by the moire interference method using a CCD camera developed by the National Research Institute for Metals. (G.K.)

  11. 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

  12. 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

  13. Experience and topical problems of surveillance and diagnosis of sodium-cooled fast breeders in the period of introducing prototype units

    International Nuclear Information System (INIS)

    Kochetkov, L.A.; Petrenko, A.A.

    1984-01-01

    The solution of the problem of increasing the safety and economy of sodium-cooled fast reactors is impossible unless appropriate surveillance and diagnostic systems have been developed. In the past, improvement of surveillance and diagnostic systems took the following directions: centralization of surveillance, increase of safety, coupling to computer and control systems. It is reported on experience gained in developing and operating individual surveillance and diagnostic systems for fast breeders. Basic objectives of further developing methods and measuring instruments, diagnostic procedures and standards of surveillance equipment are presented. (author)

  14. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-07

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

  15. 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...

  16. 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

  17. 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

  18. Localized corrosion studies on materials proposed for a safety-grade sodium-to-air decay-heat removal system for fast breeder reactors

    International Nuclear Information System (INIS)

    Mudali, U.K.; Khatak, H.S.; Dayal, R.K.; Gnanamoorthy, J.B.

    1993-01-01

    The present investigation was carried out to assess the localized corrosion resistance of materials proposed for the construction of the safety-grade sodium-to-air decay-heat removal system for fast breeder reactors. The materials, such as Alloy 800, 9Cr-1 Mo steel, and type 316LN stainless steel, in different microstructural conditions were assessed for pitting and stress-corrosion cracking resistances in a chloride medium. The results indicated that 9Cr-1Mo steel in the normalized and tempered condition can be considered for the above application from the standpoint of corrosion resistance

  19. Optimization of binary breeder reactor V - Binary breeder reactors with two and four zones and a conventional LMFBR - (Pu/U) of two zones

    International Nuclear Information System (INIS)

    Dias, A.F.; Ishiguro, Y.

    1986-04-01

    Comparative analyses of a commercial-size Pu/U-fueled liquid metal fast breeder reactor and two binary breeder reactors with different numbers of enrichment zones have been done. Principal parameters of comparison are safety and breeding characteristics and reactivity losses during an operational cycle. The comparison shows that in a binary breeder reactor, good breeding characteristics in both cycles, Pu/U and U/Th, in addition to a possibility of an efficient utilization of thorium, and superior inherent safety than current LMFBRs can be achieved. (Author) [pt

  20. 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.)

  1. New small sodium leak detection system. (2) Experiments of moving fluid ionization detector and laser leak detector in fast breeder reactor Monju

    International Nuclear Information System (INIS)

    Okazaki, Koki; Naito, Susumu; Nakayama, Kunihiko; Ito, Jun; Muto, Keitaro

    2011-01-01

    In fast breeder reactor (FBR), it is important to detect a small sodium leak for safety. As the coolant sodium used in FBR is low pressure, a small leak occurs from a crack before a pipe breaks, therefore it is possible to detect a sodium leak so as to shutdown a reactor before a pipe completely breaks. The sodium leaking from piping of a cooling system reacts with oxygen etc. including with the ambient gas to form aerosols. In the second coolant system of the prototype liquid metal FBR Monju, the ambient air around piping is drawn into radiative ionization detectors (RID). When sodium aerosols are detected by RID, chemical analysis of aerosols and the viewing of the leak point are carried out, then the sodium leak is confirmed. As the signal of RID depends strongly on temperature and RID detects the aerosols which do not include sodium (smoke, mist, volatile materials, etc.), the improvements of RID are demanded for reliability. Moving fluid ionization detectors (MID) and laser leak detectors (LLD) have been developed to improve these points. MID can reduce the dependence for temperature of the signal strength, and LLD can detect sodium atoms which are included in aerosols by means of laser-induced breakdown spectroscopy. An experiment was carried out for validating sodium leak detection system with MID and LLD under a real environment installed in the second coolant system of Monju. In the experiment, sodium was not leaked, and ambient air around piping of the secondary coolant system was introduced through actual sampling line and measured by MID and LLD. In addition a sodium level in the ambient air was measured by means of ion chromatography. The experiment showed that the background signals of MID and LLD were stable and low and that the sodium level due to salt particles contained in the ambient air was lower than three orders of magnitude than the sodium level for a small sodium leak. The plant applicability of MID and LLD was validated by this experiment

  2. Fusion breeder

    International Nuclear Information System (INIS)

    Moir, R.W.

    1982-01-01

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outline specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs

  3. Effects of fast breeders' characteristics on consumptions and expenditures related to electric power generation. An assessment for the European Community

    International Nuclear Information System (INIS)

    Graziani, G.; Zanantoni, C.

    1976-12-01

    The effects of the physical characteristics of the FBR (breeding ratio, inventory) and of the associated fuel cycle (out-of-pile time) on the results of strategy calculations performed for an electric power system are discussed. Attention is focused on the penetration of FBR, on the demand of uranium and separative work and on the economical performance of the breeders. It is pointed out that uncertainties about the out-of-pile time are as important as those concerning FBR breeding ratio and inventory

  4. Annual report of the working group 'fuel pin and fuel element mechanics' of the Institut fuer Reaktortechnik (IRT) of the Technische Hochschule Darmstadt for the Fast Breeder Project

    International Nuclear Information System (INIS)

    Fabian, H.; Humbach, W.; Lassmann, K.; Mueller, J.J.; Preusser, T.; Schmelz, K.

    1978-09-01

    This report comprises six single lectures given at an information meeting organized by the Institut fuer Reaktortechnik der Technischen Hochschule Darmstadt (IRT) in Darmstadt on April 24, 1978. The lectures are an account of work performed at IRT on the mechanics of fuel pins and fuel elements and supported by the Fast Breeder Project (PSB) of KfK. These activities can be broken down into studies of the integral fuel pin (URANUS computer code) and into multidimensional studies of the fuel pin using the finite-element method (FINEL and ZIDRIG computer codes). Moreover, a report is presented of the status of the test facility for simulation of out-of-pile cladding tube loads and of the IRT project on the simulation and analysis of radiation damage. (orig./GL) [de

  5. 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

  6. The development of a model to study the thermal behaviour of the coolant in the blind elements of a fast sodium-cooled breeder in the case of a severe hypothetical accident during the initial phase

    International Nuclear Information System (INIS)

    Genter, G.

    1981-03-01

    The enthalpy level of the coolant is studied in the interior of gaps and special elements of a fast sodium coded breeder reactor during the initial and the final stages of a hypothetical accident. For this purpose numerical models are presented to calculate the heat transport in the special element on the basis of heat conduction and axial convection. (orig./RW) [de

  7. Tubular gage for a liquid-metal-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Hutter, E.; Tuma, L.A.

    1977-01-01

    Spring-loaded plungers are arranged about a housing for insertion into a polygonal tube, one plunger for each side of the tube. Each plunger has a locking cam and sliding wedge mechanism which can overcome the spring force associated with the plunger and lock it in any position. The wedges are operated by a rod movable axially in the housing. Several housings with their associated plungers can be stacked. The stack is lowered into the polygonal tube with all of the plungers locked in a fully inward position. When the stack is in the tube, each wedge is moved to release its locking cam, allowing each of the plungers to spring outward against an inner side of the tube. Each housing will thus gage the internal dimensions of the tube at its elevation. The plungers are locked in position, the entire stack is rotated to bring the plungers into the corners described by the intersections of the flat sides, and the stack is removed from the tube whereupon the dimensions across opposite locked plungers may be read by a micrometer

  8. Rotating plug size study for liquid-metal fast breeder reactors

    International Nuclear Information System (INIS)

    Nemeth, L.J.

    1980-01-01

    A study was performed to evaluate possible rotating plug arrangements. The three-, two-, and one-rotating plug schemes were developed using a set of established restrictions and component sizes. The three-rotating plug configuration is the recommended reference design

  9. Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 3

    International Nuclear Information System (INIS)

    1975-12-01

    Included are copies of thirty-four comment letters on the Proposed Final Environmental Statement together with the ERDA replies to these letters. The letters were received from Federal, State, and local agencies, environmental and public interest groups, members of the academic and industrial communities, and individual citizens

  10. Mathematical modeling of turbulent stratified flows. Application of liquid metal fast breeders

    International Nuclear Information System (INIS)

    Villand, M.; Grand, D.

    1983-01-01

    Mathematical model of turbulent stratified flow was proposed under the following assumptions: Newtonian fluid; incompressible fluid; coupling between temperature and momentum fields according to Boussinesq approximation; two-dimensional invariance for translation or rotation; coordinates cartesian or curvilinear. Solutions obtained by the proposed method are presented

  11. Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 2

    International Nuclear Information System (INIS)

    1975-12-01

    Included are copies of fifty-six comment letters on the Proposed Final Environmental Statement together with the ERDA replies to these letters. The letters were received from Federal, State, and local agencies, environmental and public interest groups, members of the academic and industrial communities, and individual citizens

  12. Liquid-metal fast-breeder reactors: Preliminary safety and environmental information document. Volume VI

    International Nuclear Information System (INIS)

    1980-01-01

    Information is presented concerning LMFBR design characteristics; uranium-plutonium/uranium recycle homogeneous core; uranium-plutonium/uranium spiked recycle heterogeneous core; uranium-plutonium/uranium spiked recycle homogeneous core; uranium-plutonium/thorium spiked recycle heterogeneous core; uranium-plutonium/thorium spiked recycle homogeneous core; thorium-plutonium/thorium spiked recycle homogeneous core; denatured uranium-233/thorium cycle homogeneous core; safety consideration for the LMFBR; and environmental considerations

  13. Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 1

    International Nuclear Information System (INIS)

    1975-12-01

    Information is presented under the following section headings: LMFBR program options and their compatibility with the major issues affecting commercial development, Proposed Final Environmental Statement for the LMFBR program, December 1974, WASH-1535, supplemental material, and material relating to Proposed Final Environmental Statement review

  14. Design of the core and subassemblies of a liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Chaumont, J.M.; Clauzon, Pierre; Delpeyroux, Paul; Estavoyer, M.; Ginier, R.; Marmonier, Pierre; Mougniot, J.-C.

    1975-01-01

    It is shown that the main objective in designing a power station is a minimum cost of the kW-h produced and that the choice of the main parameters of a reactor is the result of a compromise. The determination of the core architecture, the shape and size of fuel pins, the thermal and hydraulic parameters and the fuel assembly design is discussed [fr

  15. US Liquid Metal Fast Breeder Reactor man-machine interface program

    International Nuclear Information System (INIS)

    Vaurio, J.K.; Change, S.A.

    1982-01-01

    The US LMFBR Man-Machine Interface Program is supportive to and an integral part of the LMFBR Safety Program. This paper describes the goal and objectives of the program, and the necessary research and development efforts with a logical structure for the orderly and timely implementation of the prgoram. Current status and near-term and long-term priority activities are also summarized

  16. AB INITIO STUDY OF ADVANCED METALLIC NUCLEAR FUELS FOR FAST BREEDER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Landa, A; Soderlind, P; Grabowski, B; Turchi, P A; Ruban, A V; Vitos, L

    2012-04-23

    Density-functional formalism is applied to study the ground state properties of {gamma}-U-Zr and {gamma}-U-Mo solid solutions. Calculated heats of formation are compared with CALPHAD assessments. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components. The decomposition curves for {gamma}-based U-Zr and U-Mo solid solutions are derived from Ising-type Monte Carlo simulations. We explore the idea of stabilization of the {delta}-UZr{sub 2} compound against the {alpha}-Zr (hcp) structure due to increase of Zr d-band occupancy by the addition of U to Zr. We discuss how the specific behavior of the electronic density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. The mechanism of possible Am redistribution in the U-Zr and U-Mo fuels is also discussed.

  17. 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)

  18. Status and prospects of thermal breeders

    International Nuclear Information System (INIS)

    1978-08-01

    The main objective of this cooperative study and of this report is to evaluate the extent to which thermal breeders might complement or serve as an alternative to fast breeders in solving the long-term nuclear fuel supply problem. A secondary objective is to consider in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability, and fuel cycle versatility to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

  19. Building on knowledge base of sodium cooled fast breeder reactors to develop materials technology for fusion reactors

    International Nuclear Information System (INIS)

    Raj, B.; Rao, B.S.

    2007-01-01

    Full text of publication follows: There are strong synergies between fission and fusion research. Similarities extend to candidate structural materials, methods of theoretical and experimental investigations, high operating temperatures, fast neutron exposure and similar underlying physical phenomena. Enhancing the cross-fertilization of fission and fusion R and D programmes would mutually benefit the development of both kinds of nuclear reactors for sustainable nuclear energy. Materials used in fusion energy systems must be fully code qualified for high-temperature applications that include creep, low cycle fatigue, thermo-mechanical fatigue and creep-fatigue interaction. Evaluation of various mechanical properties of reduced activation ferritic-martensitic alloys are underway for establishing reliable data bases for development of a suitable code and further optimization and redesign of the alloys for fusion reactor components. The alloys 316L (N) and derived variants of Mod.9Cr1Mo steel are anticipated for use in fusion systems and the RCC-MR code would serve as a backbone for the initial design of various components. In this respect, the efforts made in India for developing these materials, their characterization and applicability of RCC-MR code are highlighted. The salient features of consortium approach and the experience that has led to manufacturing of main and safety vessel components from 316L (N) with close dimensional tolerances, thick section forged components and seamless tubes of Modified 9Cr-1Mo steel for steam generators will be presented. The finite element analysis has been used for optimal design of die and punch for the near net shape forming of petals of main vessel of PFBR. The safe temperature and strain rate domains were established using dynamic materials model for forming of 316L (N) and 9Cr1Mo steels by various industrial processes. Welds are weak links in the design of both fission and fusion systems. Detailed investigations have led to

  20. 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)

  1. Fast thermal cycling-enhanced electromigration in power metallization

    NARCIS (Netherlands)

    Nguyen, Van Hieu; Salm, Cora; Krabbenborg, B.H.; Krabbenborg, B.H.; Bisschop, J.; Mouthaan, A.J.; Kuper, F.G.

    Fast thermal nterconnects used in power ICs are susceptible to short circuit failure due to a combination of fast thermal cycling and electromigration stresses. In this paper, we present a study of electromigration-induced extrusion short-circuit failure in a standard two level metallization

  2. Clinch River Breeder Reactor Plant: a building block in nuclear technology

    International Nuclear Information System (INIS)

    McCormack, M.

    1979-01-01

    Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP

  3. Clinch River Breeder Reactor Plant: a building block in nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    McCormack, M.

    1979-01-01

    Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP.

  4. The breeder reactor and Europe

    International Nuclear Information System (INIS)

    Daglish, J.

    1979-01-01

    A report is given of a conference on the breeder reactor and Europe held in Lucerne, Switzerland from 14 - 17 October 1979 sponsored by the Swiss Association for Atomic Energy and the Association of European Atomic Forums. The underlying theme of the conference was the question that if nuclear power is to play a major role in meeting world energy needs in the long term, thermal reactors must in time be complemented with more advanced reactor systems that conserve uranium resources which are huge but not unlimited. This is not questioned; disagreement begins with discussion of the desirability of the breeder, and how fast and how far the introduction of such reactors should go. Aspects considered at the conference which are especially dealt with in this review are; why breed, commercial aspects, alternatives to the LMFBR, how to build a fast reactor, the breeder programmes in Europe, Britain, the Soviet Union, Japan and the United States. (U.K.)

  5. 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.

  6. 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)

  7. Metallic uranium as fuel for fast reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de

    1988-01-01

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

  8. Breeder reactors over the world

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    After a short recall on the development of research programs, this paper reviews the fast breeder reactor operating, in construction or in project over the world (USA, France, Italy, RFG, India, Japan and U.K.). Thermal and electrical power output, and operation data are given [fr

  9. Fast Flux Test Facility metal fuel pin fabrication

    International Nuclear Information System (INIS)

    Benecke, M.W.; Dittmer, J.O.; Feigenbutz, L.V.

    1989-05-01

    A new initiative to develop, irradiate, and qualify a binary uranium/zirconium metal fuel system in the Fast Flux Test Facility (FFTF) has been implemented by the Westinghouse Hanford Company for the US Department of Energy. Metal fuel test assemblies have been designed and fabricated and are now being irradiated in FFTF to provide the data needed to support the potential use of binary fuels in FFTF and other liquid metal reactors. These development efforts support licensing activities for metal fuel use in near-term advanced liquid metal reactors. New metal fuel pin design features, fabrication development, and fabrication processes for three metal fuel tests will be described and their irradiation status reported in this paper. 11 figs

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  11. 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)

  12. Development of metallic fuels for fast reactors

    International Nuclear Information System (INIS)

    Singh, R.P.

    2011-01-01

    A very rapid growth rate of 100-fold increase in the next 50 years has been targeted for nuclear energy in India for which metallic fuel will be introduced in FBRs after 2020. Many innovative and challenging fuel design concepts for metal fuels are under consideration, which require extensive research and development work to generate database for the designers. The designs under active considerations are one in which the fuel will either be sodium bonded ternary U-15Pu-6Zr alloy or mechanically/sodium bonded binary U-15Pu alloy with a Zr liner between the fuel and the clad. The decision on the choice of the fuel will be based on test fuel irradiations in FBTR, subsequent PIE results, modeling studies and closing the fuel cycle through pyrometallurgical route. The development of metallic fuel is being pursued jointly by BARC and IGCAR. While the mechanical bonding concept with Pu-U alloy as fuel is being pursued at BARC, sodium bonding concept is being developed at IGCAR. BARC has already established the fabrication route for the mechanical bonding of U-Zr alloy, through co-swaging route. The properties of the fuels are evaluated. The fabrication of U-Pu alloy fuel pins through this route is now under development. It is now proposed to carry out the irradiation of test fuel pins, fabricated through both the routes, in FBTR in order to select the route for further development. This will be followed by subassembly level irradiation in FBTR to obtain experience on large scale fabrication of metallic fuel and to establish its performance

  13. Advanced methods for the fabrication of mixed uranium plutonium oxide, monocarbide and mononitride fuels for fast breeder reactors

    International Nuclear Information System (INIS)

    Ganguly, C.

    1988-01-01

    Mixed uranium plutonium monocarbide (MC) and mononitride (MN) are considered as advanced LMFBR fuels. 'Powder-pellet (POP) is the conventional method for production of MOX, MC and MN fuel pellets starting from UO 2 and PuO 2 powders. The POP route involves generation and handling of plutonium bearing fine powder or dust particles which has the problem of radiotoxic dust hazard. Further, fine powders have poor flowability which makes automation and remote fabrication difficult. The combination of ammonium uranyl plutonyl carbonate (AUPuC) process and low temperature oxidative sintering (LTS) minimises the radiotoxic dust hazard and the fabrication cost of MOX fuel and is considered as advanced POP route. Vibro-sol (or sphere-pac) and sol-gel microsphere pelletisation (SGMP) are advanced methods for fabrication of MOX, MC and MN fuels. Here, dust-free and free-flowing gel microspheres of the oxide or oxide-carbon are prepared from heavy metal nitrate feed solutions by 'internal' or 'external' gelation processes. The gel microspheres are subjected to controlled calcination for MOX or carbothermic reduction for MC and MN. Thereafter, the microspheres are either vibropacked in fuel pins or directly pelletised and sintered. These processes minimise radiotoxic dust hazard, facilitate automation and remote processing and ensure excellent microhomogeneity. The present paper summarises the state-of-art of the POP, vibrosol and SGMP processes for the fabrication of MOX, MC and MN fuels, highlighting the author's experience in SGMP process. (author). 21 refs., 11 figs

  14. Volatile Elements Retention During Injection Casting of Metallic Fuel Slug for a Recycling Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Jong-Hwan; Song, Hoon; Kim, Hyung-Tae; Oh, Seok-Jin; Kuk, Seoung-Woo; Keum, Chang-Woon; Lee, Jung-Won; Kim, Ki-Hwan; Lee, Chan-Bock

    2015-01-01

    The as-cast fuels prepared by injection casting were sound and the internal integrities were found to be satisfactory through gamma-ray radiography. U and Zr were uniform throughout the matrix of the slug, and the impurities, i.e., oxygen, carbon, and nitrogen, satisfied the specification of the total impurities of less than 2000 ppm. The losses of the volatile Mn were effectively controlled using argon over pressures, and dynamic pumping for a period of time before injection showed no detrimental effect on the Mn loss by vaporization. This result suggests that volatile minor actinide-bearing fuels for SFRs can be prepared by improved injection methods. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, several injection casting methods were applied in order to prepare metallic fuel for an fast reactor that control the transport of volatile elements during fuel melting and casting. Mn was selected as a surrogate alloy since it possesses a total vapor pressure equivalent to that of a volatile minor actinide-bearing fuel. U.10Zr and U.10Zr.5Mn (wt%) metallic fuels were injection cast under various casting conditions and their soundness was characterized

  15. 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

  16. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    1999-04-01

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

  17. Present state of new technologies of nuclear power generation, and technological development of fast-breeder reactor and next-generation light water reactor

    International Nuclear Information System (INIS)

    Aoyagi, Toshihiro; Ando, Masato

    2014-01-01

    This paper introduces the present state of development of FBR in Japan and international cooperation, the development of HP-ABWR and HP-APWR as the next-generation light water reactors, and SMR development in the United States. As for FBR, the following situations are described: (1) history of development in Japan in the past, (2) history of change due to the readjustment of development plan caused by the accident of Fukushima Daiichi Nuclear Power Station, in which shift to FaCT phase 2 was suspended, and the approach to the establishment of safety standards for sodium-cooled FBR and its international standardization was adopted, and (3) future challenges. As for the Japan - France fast-breeder reactor development cooperation, the conclusion of the Japan - France inter-government agency agreement, and Japan's cooperation plan and system are described. Next, as for HP-ABWR and HP-APWR, the development goal and concept of each plant, and the element technologies required for the success are described. On the other hand, the small reactor development in the United States started with the aim of the securement of domestic technology base, contribution to reduction in carbon dioxide emissions, and its export to new entry countries for nuclear energy. This project aimed the practical use of SMR, and started 'financial support program for small reactors' to allocate about 452 million dollars to maximum two units of SMRs in the next five years. This project is outlined. (A.O.)

  18. A review of fusion breeder blanket technology, part 1

    International Nuclear Information System (INIS)

    Jackson, D.P.; Selander, W.N.; Townes, B.M.

    1985-01-01

    This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

  19. Unconventional liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

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

    1989-06-01

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

  20. 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)

  1. 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

  2. Coated ceramic breeder materials

    Science.gov (United States)

    Tam, Shiu-Wing; Johnson, Carl E.

    1987-01-01

    A breeder material for use in a breeder blanket of a nuclear reactor is disclosed. The breeder material comprises a core material of lithium containing ceramic particles which has been coated with a neutron multiplier such as Be or BeO, which coating has a higher thermal conductivity than the core material.

  3. Control rod calibration methods for fast breeder reactors applied to Phenix; Les methodes d'etalonnage des barres de commande des reacteurs a neutrons rapides application a Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Lecourt, G

    1998-06-18

    The control and the emergency shutdown of a fast breeder reactor depends essentially on control rods. For this reason, it is imperative to know exactly how much anti reactivity is introduced with the rods in the reactor core. Different methods have been compared in order to see if they are compatible with Phenix reactor. Their limits have been studied. The shadow and anti shadow effects that can the rods make one to the other and then their effective weight of the rods screen have been clarified. (N.C.)

  4. Characteristics of Matrix Metals in Which Fast Diffusion of Foreign Metallic Elements Occurs

    Science.gov (United States)

    Mae, Yoshiharu

    2018-04-01

    A few foreign elements are known to diffuse faster than the self-diffusion of the matrix metal. However, the characteristics of the matrix metal, which contribute to such fast diffusion remain unknown. In this study, the diffusion coefficients of various elements were plotted on a TC-YM diagram. The matrix metals that show fast diffusion are located in the low thermal conductivity range of the TC-YM diagram, while diffuser elements that undergo fast diffusion are mainly gulf elements such as Fe, Ni, Co, Cr, and Cu. The gulf elements are those that show the largest combination of thermal conductivity and Young's modulus. The great difference in the electron mobility between the matrix metal and diffuser elements generates a repulsive force between them, and the repulsive force—acting between the soft and large atoms of the matrix metal and the hard and small atoms of the diffuser elements—deforms the atoms of the matrix metal to open passageways for fast diffusion of diffuser elements.

  5. Alternative Fabrication of Recycling Fast Reactor Metal Fuel

    International Nuclear Information System (INIS)

    Kim, Ki-Hwan; Kim, Jong Hwan; Song, Hoon; Kim, Hyung-Tae; Lee, Chan-Bock

    2015-01-01

    Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. In order to develop innovative fabrication method of metal fuel for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, metal fuel slugs for SFR have been fabricated by modified casting method, and characterized to evaluate the feasibility of the alternative fabrication method. In order to prevent evaporation of volatile elements such as Am and improve quality of fuel slugs, alternative fabrication methods of metal fuel slugs have been studied in KAERI. U-10Zr-5Mn fuel slug containing volatile surrogate element Mn was soundly cast by modified injection casting under modest pressure. Evaporation of Mn during alternative casting could not be detected by chemical analysis. Mn element was most recovered with prevention of evaporation by alternative casting. Modified injection casting has been selected as an alternative fabrication method in KAERI, considering evaporation prevention, and proven benefits of high productivity, high yield, and good remote control

  6. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 1

    International Nuclear Information System (INIS)

    Malang, S.; Reimann, J.; Sebening, H.; Barleon, L.; Bogusch, E.; Bojarsky, E.; Borgstedt, H.U.; Buehler, L.; Casal, V.; Deckers, H.; Feuerstein, H.; Fischer, U.; Frees, G.; Graebner, H.; John, H.; Jordan, T.; Kramer, W.; Krieg, R.; Lenhart, L.; Malang, S.; Meyder, R.; Norajitra, P.; Reimann, J.; Schwenk-Ferrero, A.; Schnauder, H.; Stieglitz, R.; Oschinski, J.; Wiegner, E.

    1991-12-01

    A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary, Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated R and D-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required R and D-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.) [de

  7. Breeding nuclear fuels with accelerators: replacement for breeder reactors

    International Nuclear Information System (INIS)

    Grand, P.; Takahashi, H.

    1984-01-01

    One application of high energy particle accelerators has been, and still is, the production of nuclear fuel for the nuclear energy industry; tantalizing because it would create a whole new industry. This approach to producing fissile from fertile material was first considered in the early 1950's in the context of the nuclear weapons program. A considerable development effort was expended before discovery of uranium ore in New Mexico put an end to the project. Later, US commitment to the Liquid Metal Fast Breeder Reactors (LMFBR) killed any further interest in pursuing accelerator breeder technology. Interest in the application of accelerators to breed nuclear fuels, and possibly burn nuclear wastes, revived in the late 1970's, when the LMFBR came under attack during the Carter administration. This period gave the opportunity to revisit the concept in view of the present state of the technology. This evaluation and the extensive calculational modeling of target designs that have been carried out are promising. In fact, a nuclear fuel cycle of Light Water Reactors and Accelerator Breeders is competitive to that of the LMFBR. At this time, however, the relative abundance of uranium reserves vs electricity demand and projected growth rate render this study purely academic. It will be for the next generation of accelerator builders to demonstate the competitiveness of this technology versus that of other nuclear fuel cycles, such as LMFBR's or Fusion Hybrid systems. 22 references, 1 figure, 5 tables

  8. Special topics reports for the reference tandem mirror fusion breeder: liquid metal MHD pressure drop effects in the packed bed blanket. Vol. 1

    International Nuclear Information System (INIS)

    McCarville, T.J.; Berwald, D.H.; Wong, C.P.C.

    1984-09-01

    Magnetohydrodynamic (MHD) effects which result from the use of liquid metal coolants in magnetic fusion reactors include the modification of flow profiles (including the suppression of turbulence) and increases in the primary loop pressure drop and the hydrostatic pressure at the first wall of the blanket. In the reference fission-suppressed tandem mirror fusion breeder design concept, flow profile modification is a relatively minor concern, but the MHD pressure drop in flowing the liquid lithium coolant through an annular packed bed of beryllium/thorium pebbles is directly related to the required first wall structure thickness. As such, it is a major concern which directly impacts fissile breeding efficiency. Consequently, an improved model for the packed bed pressure drop has been developed. By considering spacial averages of electric fields, currents, and fluid flow velocities the general equations have been reduced to simple expressions for the pressure drop. The averaging approach results in expressions for the pressure drop involving a constant which reflects details of the flow around the pebbles. Such details are difficult to assess analytically, and the constant may eventually have to be evaluated by experiment. However, an energy approach has been used in this study to bound the possible values of the constant, and thus the pressure drop. In anticipation that an experimental facility might be established to evaluate the undetermined constant as well as to address other uncertainties, a survey of existing facilities is presented

  9. Accelerator breeder concept

    International Nuclear Information System (INIS)

    Bartholomew, G.A.; Fraser, J.S.; Garvey, P.M.

    1978-10-01

    The principal components and functions of an accelerator breeder are described. The role of the accelerator breeder as a possible long-term fissile production support facility for CANDU (Canada Deuterium Uranium) thorium advanced fuel cycles and the Canadian research and development program leading to such a facility are outlined. (author)

  10. Results of fundamental research and development of partitioning and transmutation technology of long-lived-nuclides for fast breeder reactor cycle (from 2001 to 2004 Japanese Fiscal Years)

    International Nuclear Information System (INIS)

    Yamashita, Kiyonobu; Ozawa, Masaki; Ikegami, Tetsuo; Osaka, Masahiko; Ohki, Shigeo; Tachi, Yoshiaki; Furutaka, Kazuyoshi; Nakamura, Shoji; Harada, Hideo

    2005-03-01

    Research Evaluation Committee carried out a pre-evaluation of 'Research and Development of Partitioning and Transmutation Technology of Long-Lived-Nuclides' in Aug 2000. Following results are obtained from the research and development. Two extractant systems, capable of recovering all actinides in spent fuel, were newly nominated. Tertiary pyridine resin process possesses high abilities on Am/Cm as well as Miner Actinides (MA)/Lanthanides (Ln) separations. Neutron capture cross sections of 7 nuclides of MA and Fission Products (FP) were determined in nuclear data measurements. Some of those measurements are for the first time in the world. An advanced measurement system of a full solid angle Bi 4 Ge 3 O 12 detector etc., was developed to measure the energy dependence of the neutron capture cross sections. Uncertainty in isometric ratio of 241 Am neutron capture reactions was reduced through analyses of MA samples irradiated in the experimental fast reactor 'JOYO'. Main production processes of inert-matrix fuels were developed through trial-production experiments with mock materials. A complex process fuel cycle concept was proposed that was able to burn LWR spent fuels up to 400 GWd/t only with heat treatment without reprocessing. Following results are also obtained from the research and development newly added after the pre-evaluation of Research Evaluation Committee, to establish Partitioning and Transmutation Technology of Long-Lived-Nuclides. Catalytic-electrolytic extraction was successfully developed to recover rare metal fission products (RMFP) as Ru, Rh, etc. RMFP-deposit electrodes showed high catalytic ability on the hydrogen production by water electrolysis. Material properties of 6 promising Iodine compounds such as CuI, BaI 2 were selected for 129 I-transmutation targets. An environmental Impact index was newly introduced instead of conventional radio-toxicity for quantitative evaluation of geological repository effect. These achievements in the first

  11. 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

  12. Leak detection of steam or water into sodium in steam generators of liquid-metal fast breeder reactors

    International Nuclear Information System (INIS)

    Hans, R.; Dumm, K.

    1977-01-01

    The leakage of water or steam into sodium in LMFBR steam generators, including a study of how leaks are detected and located as well as the potential damage that could be caused by such leaks, is surveyed. The most interesting steam generator designs evolving in those countries that develop and construct LMFBRs are presented. The relevant protection measures are described. Fault conditions are defined and descriptions given of possible sequences of events leading to abnormal conditions in a steam generator. Taking into account theory, the potential of the hydrogen and oxygen detection systems is discussed. Different hydrogen and oxygen detection systems are fully described. In so far as interesting technical solutions are concerned, previously developed devices have also been taken into account. The way oxygen detection supplements hydrogen detection is described by listing the available oxygen measuring devices and the relevant theory. Only a few sonic and accelerometer measurements have been made on complete steam generator units so there is little system data available. Descriptions, however, have been included to give the state of the art achieved for the sensors and the achieved sensitivities or band widths. The potential of this monitoring method is made evident by adding the technical data of the sensors. Furthermore, the available systems for monitoring medium and large leakages are described. Finally, recommendations are made concerning steam generator development and the application of hydrogen and oxygen detection systems, as well as acoustic measuring methods for small-leakage detection

  13. Status and prospects of thermal breeders and their effect on fuel utilization

    International Nuclear Information System (INIS)

    1979-01-01

    The report evaluates the extent to which thermal breeders and near-breeders might complement fast breeders or serve as an alternative in solving the long-term nuclear fuel supply problem. It considers in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability and fuel cycle versatility in order to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

  14. Ceramic breeder materials

    International Nuclear Information System (INIS)

    Johnson, C.E.

    1990-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 32 refs., 1 fig., 1 tab

  15. Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview

    International Nuclear Information System (INIS)

    Wei, B.C.; Cooper, W.L. Jr.; Dhalla, A.K.

    1987-09-01

    Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work

  16. Liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Scott, D.

    1981-01-01

    An improved method of constructing the diagrid used to support fuel assemblies of liquid metal fast breeder reactors, is described. The functions of fuel assembly support and coolant plenum are performed by discrete components of the diagrid each of which can serve the function of the other in the event of failure of one of the components. (U.K.)

  17. Fast and scalable synthesis of uniform zirconium-, hafnium-based metal-organic framework nanocrystals.

    Science.gov (United States)

    He, Ting; Xu, Xiaobin; Ni, Bing; Wang, Haiqing; Long, Yong; Hu, Wenping; Wang, Xun

    2017-12-14

    Metal-organic frameworks based on zirconium or hafnium possess tantalizing commercial prospects due to their high stability but require a long reaction time to form crystals. The fast synthesis of uniform Zr-, Hf-MOF nanocrystals at scale is of key importance in the potential commercial application of MOFs. In this work, we have developed a versatile strategy through controlling the hydrolysis and nucleation of metal salts in the presence of acetic acid and water; up to 24 grams of UiO-66-NH 2 nanocrystals with a uniform octahedron could be synthesized within 15 minutes using a one step method. The current synthetic strategy could be extended to other Zr-, Hf-MOF nanocrystals [UiO-66-Fast, UiO-66-(OH) 2 -Fast, UiO-66-2,6-NDC-Fast, UiO-67-Fast, BUT-12-Fast, PCN-222-Ni-Fast, PCN-222-Co-Fast, Hf-UiO-66-Fast, Hf-UiO-66-NH 2 -Fast, Hf-UiO-66-(OH) 2 -Fast, Hf-UiO-66-2,6-NDC-Fast and Hf-BUT-12-Fast]. Significantly, when noble metal nanoparticles (NPs) are introduced into MOF precursors, NPs encapsulated in MOFs with excellent dispersion have also been obtained and show outstanding performance in catalysis. This facile procedure is expected to pave the way to expand the commercial applications of MOFs.

  18. Pyrometallurgical processing of Integral Fast Reactor metal fuels

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  19. Thermal-hydraulic methods in fast reactor safety

    International Nuclear Information System (INIS)

    Weber, D.P.; Briggs, L.L.

    1985-01-01

    Methods for the solution of thermal-hydraulic problems in liquid metal fast breeder reactors (LMFBRs) arising primarily from transient accident analysis are reviewed. Principal emphasis is given to the important phenomenological issues of sodium boiling and fuel motion. Descriptions of representative phenomenological and mathematical models, computational algorithms, advantages and limitations of the approaches, and current research needs and directions are provided

  20. Uranium resources and their implications for fission breeder and fusion hybrid development

    International Nuclear Information System (INIS)

    Max, C.E.

    1984-01-01

    Present estimates of uranium resources and reserves in the US and the non-Communist world are reviewed. The resulting implications are considered for two proposed breeder technologies: the liquid metal fast breeder reactor (LMFBR) and the fusion hybrid reactor. Using both simple arguments and detailed scenarios from the published literature, conditions are explored under which the LMFBR and fusion hybrid could respectively have the most impact, considering both fuel-supply and economic factors. The conclusions emphasize strong potential advantages of the fusion hybrid, due to its inherently large breeding rate. A discussion is presented of proposed US development strategies for the fusion hybrid, which at present is far behind the LMFBR in its practical application and maturity

  1. Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

    International Nuclear Information System (INIS)

    Cho, S.M.; Zury, H.L.; Cook, M.E.; Fair, C.E.

    1978-12-01

    The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

  2. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1985-01-01

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

  3. 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.

  4. 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)

  5. Improvements in the fabrication of metallic fuels

    International Nuclear Information System (INIS)

    Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

    1989-01-01

    Argonne National Laboratory (ANL) is currently developing a new liquid-metal-cooled breeder reactor known as the Integral Fast Reactor (IFR). The IFR represents the state of the art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, discussed in this paper, will support ANL-West's (ANL-W) fully remote fuel cycle facility, which is an integral part of the IFR concept

  6. Improvements in fabrication of metallic fuels

    International Nuclear Information System (INIS)

    Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

    1989-12-01

    Argonne National Laboratory is currently developing a new liquid- metal cooled breeder reactor known as the Integral Fast Reactor (IFR). IFR fuels represent the state-of-the-art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, to be discussed below, will support the fully remote fuel cycle facility that as an integral part of the IFR concept will be demonstrated at the EBR-II site. 3 refs

  7. Ceramic breeder materials

    International Nuclear Information System (INIS)

    Johnson, C.E.; Kummerer, K.R.; Roth, E.

    1987-01-01

    Ceramic materials are under investigation as potential breeder material in fusion reactors. This paper will review candidate materials with respect to fabrication routes and characterization, properties in as-fabricated and irradiated condition, and experimental results from laboratory and inpile investigations on tritium transport and release. Also discussed are the resources of beryllium, which is being considered as a neutron multiplier. The comparison of ceramic properties that is attempted here aims at the identification of the most-promising material for use in a tritium breeding blanket. 82 refs., 12 figs., 5 tabs

  8. Optimization of the binary breeder reactor. VIII annular core fueled with 233U - 238U and Pu-238U

    International Nuclear Information System (INIS)

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

    1988-04-01

    First cycle burnup characteristics of a 1200 MWe binary breeder reactor with annular core fueled with metallic 233 U- 238 U-Zr, Pu- 238 U-Zr and Th in the blankets have been analysed. The Doppler effect is small as expected in a metal fueled fast reactor. The sodium void reactivity is, in general, smaller than in metal fueled homogeneous fast reactors of 1 m core height. The estimates of the required and available control rod worths show a large shutdown margin throughout the operational cycle. There are flexibilities in the blanket fueling and well balanced breeding in the two cycles, uranium and thorium, with doubling times of about 20 years are possible. (author) [pt

  9. Review of fast reactor activities in India

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1982-01-01

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

  10. R + D work on gas-cooled breeder development

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Jacobs, G.; Meyer, L.; Rehme, K.; Schumacher, G.; Wilhelm, D.

    1978-01-01

    The development work for the gas-cooled breeder in the Karlsruhe Nuclear Research Center may be assigned to two different groups: a) Investigations on fuel elements. b) Studies concerning the safety of gas-cooled fast breeder reactors. To the first group there belongs the work related to the: - heat transfer between fuel elements and coolant gas, - influence of increased content of water vapor in helium or the fuel rods. The second group concerns: - establishing a computer code for transient calculations in the primary and secondary circuit of a gas-cooled fast breeder reactor, - steam reactivity coefficients, - the core destruction phase of hypothetical accidents, - the core-catcher using borax. (orig./RW) [de

  11. Automated breeder fuel fabrication

    International Nuclear Information System (INIS)

    Goldmann, L.H.; Frederickson, J.R.

    1983-01-01

    The objective of the Secure Automated Fabrication (SAF) Project is to develop remotely operated equipment for the processing and manufacturing of breeder reactor fuel pins. The SAF line will be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at the Department of Energy's (DOE) Hanford site near Richland, Washington, and is operated by the Westinghouse Hanford Company (WHC). The fabrication and support systems of the SAF line are designed for computer-controlled operation from a centralized control room. Remote and automated fuel fabriction operations will result in: reduced radiation exposure to workers; enhanced safeguards; improved product quality; near real-time accountability, and increased productivity. The present schedule calls for installation of SAF line equipment in the FMEF beginning in 1984, with qualifying runs starting in 1986 and production commencing in 1987. 5 figures

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-20

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. Chemistry R and D for back end fuel cycle for fast reactors in India

    International Nuclear Information System (INIS)

    Vasudeva Rao, P.R.

    2011-01-01

    The development of fast reactors and the associated fuel cycle is one of the focal programmes of vital interest for India to meet its energy requirements in the 21st Century in a sustainable manner. The targets for Indian fast reactor fuel cycle include simplification of the fabrication route, irradiation of fuel to high burn-up of over 150 GWd/Te, reprocessing of fuel with minimum cooling period consistent with fuelling cycle of the reactor and with near quantitative recovery, and minimizing of waste volume for ultimate disposal. In the last year, the closure of the fuel cycle for Fast Breeder Test Reactor was established by the reprocessing of the mixed carbide fuel of FBTR and re-fabrication of the fuel. Simultaneously with the construction of the 500 MWe Prototype Fast Breeder Reactor, steps have also been taken for setting up of a fuel cycle facility to close its fuel cycle. For enhancing the growth of the fast reactors in India, and at the same time optimally utilizing its uranium resources, metal fuelled fast breeder reactors are proposed to be established in the next decade. Since the metallic fuels will be reprocessed through the pyrochemical route, R and D for establishing the molten salt electrorefining process for metal fuel has been taken up in a comprehensive manner. The presentation would highlight the chemistry R and D related to back end of fast reactor fuel cycle, being pursued in IGCAR. (author)

  15. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    Sep 4, 2015 ... The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the ...

  16. Economic impact of fuel subassembly design and management in breeder reactors

    International Nuclear Information System (INIS)

    Wehmann, U.K.; Pilate, S.

    1988-01-01

    The fuel cycle costs of sodium cooled fast breeder reactors are strongly influenced by the design of fuel subassemblies and their use during reactor operation. The optimization of these two aspects is therefore an important task of the core design. The impact of maximum burnup as a design limit, fuel element management, and fuel element design on fuel cycle costs and plant availability of fast breeder reactors is illustrated for the reactor SNR 2. (orig.) [de

  17. Canadian accelerator breeder system development

    International Nuclear Information System (INIS)

    Schriber, S.O.

    1982-11-01

    A shortage of fissile material at a reasonable price is expected to occur in the early part of the twenty-first century. Converting fertile material to fissile material by electronuclar methods is an option that can extend th world's resources of fissionable material, supplying fuel for nuclear power stations. This paper presents the rationale for electronuclear breeders and describes the Canadian development program for an accelerator breeder facility that could produce 1 Mg of fissile material per year

  18. Validation of models for the analysis of the transient behavior of metallic fast reactor fuel

    International Nuclear Information System (INIS)

    Kramer, J.M.; Hughes, T.H.; Gruber, E.E.

    1989-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in U-Pu-Zr metal alloys as a fuel for sodium-cooled fast reactors. Part of the attractiveness of the IFR concept is the improvement in reactor safety margins through inherent features of a metal-fueled LMR core. In order to demonstrate these safety margins it is necessary to have computer codes available to analyze the detailed response of metallic fuel to a wide range of accident initiators. Two of the codes that play a key role in assessing this response are the STARS fission gas behavior code and the FPIN2 fuel pin mechanics code. Verification and validation are two important components in the development of models and computer codes. Verification demonstrates through comparison of calculations with analytical solutions that the methodology and algorithms correctly solve the equations that govern the phenomena being modeled. Validation, on the other hand, demonstrates through comparison with data that the phenomena are being modeled correctly. Both components are necessary in order to have the confidence to extrapolate the calculations to reactor accident conditions. This paper presents the results of recent progress in the validation of models for the analysis of the behavior of metallic fast reactor fuel. 9 refs., 7 figs

  19. An overview of the Indian program related to fast reactor core mechanical behaviour

    International Nuclear Information System (INIS)

    Govindarajan, S.; Bhoje, S.B.; Paranjpe, S.R.

    1984-01-01

    This Indian review paper presents the evolution of the fast breeder program which began with fast breeder test reactor (FBTR) commencing in 1972. The state-of-art in the field of core mechanical behaviour is reviewed

  20. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

    Greenspan, E.; Miley, G.H.

    1983-08-01

    This report undertakes a preliminary assessment of the prospects of tritium-assisted D-D fuel cycle fusion breeders. Two well documented fusion power reactor designs - the STARFIRE (D-T fuel cycle) and the WILDCAT (Cat-D fuel cycle) tokamaks - are converted into fusion breeders by replacing the fusion electric blankets with 233 U producing fission suppressed blankets; changing the Cat-D fuel cycle mode of operation by one of the several tritium-assisted D-D-based modes of operation considered; adjusting the reactor power level; and modifying the resulting plant cost to account for the design changes. Three sources of tritium are considered for assisting the D-D fuel cycle: tritium produced in the blankets from lithium or from 3 He and tritium produced in the client fission reactors. The D-D-based fusion breeders using tritium assistance are found to be the most promising economically, especially the Tritium Catalyzed Deuterium mode of operation in which the 3 He exhausted from the plasma is converted, by neutron capture in the blanket, into tritium which is in turn fed back to the plasma. The number of fission reactors of equal thermal power supported by Tritium Catalyzed Deuterium fusion breeders is about 50% higher than that of D-T fusion breeders, and the profitability is found to be slightly lower than that of the D-T fusion breeders

  1. SXR Continuum Radiation Transmitted Through Metallic Filters: An Analytical Approach To Fast Electron Temperature Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Delgado-Aparicio, L.; Tritz, K.; Kramer, T.; Stutman, D.; Finkentha, M.; Hill, K.; Bitter, M.

    2010-08-26

    A new set of analytic formulae describes the transmission of soft X-ray (SXR) continuum radiation through a metallic foil for its application to fast electron temperature measurements in fusion plasmas. This novel approach shows good agreement with numerical calculations over a wide range of plasma temperatures in contrast with the solutions obtained when using a transmission approximated by a single-Heaviside function [S. von Goeler, Rev. Sci. Instrum., 20, 599, (1999)]. The new analytic formulae can improve the interpretation of the experimental results and thus contribute in obtaining fast teperature measurements in between intermittent Thomson Scattering data.

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. FFTF [fast flux tests facility] metal fuel pin sodium bond quality verification

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1989-01-01

    The Fast Flux Tests Facility (FFTF) Series III driver fuel design consists of U-10 Zr fuel slugs contained in a ferritic alloy cladding. A liquid-metal, sodium bond between the fuel and cladding is required to prevent unacceptable temperatures during operation. Excessive voiding or porosity in the sodium thermal bond could result in localized fuel melting during irradiation. It is therefore imperative that bond quality be verified during fabrication of these metal fuel pins prior to irradiation. Cladding stripping of two spare metal fuel pins was successfully performed to inspect the quality of the sodium thermal bond and to compare observed defects in the bond with eddy current measurements. Bond quality was found to be generally good, thereby confirming the fabrication process used to bond the fuel and cladding with liquid-metal sodium to be acceptable. Observed porosity distribution in the sodium bond correlated well with eddy current indications

  4. Results of thermal test of metallic molybdenum disk target and fast-acting valve testing

    Energy Technology Data Exchange (ETDEWEB)

    Virgo, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Chemerisov, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Gromov, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Jonah, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, G. F. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-01

    This report describes the irradiation conditions for thermal testing of helium-cooled metallic disk targets that was conducted on March 9, 2016, at the Argonne National Laboratory electron linac. The four disks in this irradiation were pressed and sintered by Oak Ridge National Laboratory from molybdenum metal powder. Two of those disks were instrumented with thermocouples. Also reported are results of testing a fast-acting-valve system, which was designed to protect the accelerator in case of a target-window failure.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  7. International cooperation on breeder reactors

    International Nuclear Information System (INIS)

    Gray, J.E.; Kratzer, M.B.; Leslie, K.E.; Paige, H.W.; Shantzis, S.B.

    1978-01-01

    In March 1977, as the result of discussions which began in the fall of 1976, the Rockefeller Foundation requested International Energy Associates Limited (IEAL) to undertake a study of the role of international cooperation in the development and application of the breeder reactor. While there had been considerable international exchange in the development of breeder technology, the existence of at least seven major national breeder development programs raised a prima facie issue of the adequacy of international cooperation. The final product of the study was to be the identification of options for international cooperation which merited further consideration and which might become the subject of subsequent, more detailed analysis. During the course of the study, modifications in U.S. breeder policy led to an expansion of the analysis to embrace the pros and cons of the major breeder-related policy issues, as well as the respective views of national governments on those issues. The resulting examination of views and patterns of international collaboration emphasizes what was implicit from the outset: Options for international cooperation cannot be fashioned independently of national objectives, policies and programs. Moreover, while similarity of views can stimulate cooperation, this cannot of itself provide compelling justification for cooperative undertakings. Such undertakings are influenced by an array of other national factors, including technological development, industrial infrastructure, economic strength, existing international ties, and historic experience

  8. Reprocessing in breeder fuel cycles

    International Nuclear Information System (INIS)

    Burch, W.D.; Groenier, W.S.

    1982-01-01

    Over the past decade, the United States has developed plans and carried out programs directed toward the demonstration of breeder fuel reprocessing in connection with the first breeder demonstration reactor. A renewed commitment to moving forward with the construction of the Clinch River Breeder Reactor (CRBR) has been made, with startup anticipated near the end of this decade. While plans for the CRBR and its associated fuel cycle are still being firmed up, the basic research and development programs required to carry out the demonstrations have continued. This paper updates the status of the reprocessing plans and programs. Policies call for breeder recycle to begin in the early to mid-1990's. Contents of this paper are: (1) evolving plans for breeder reprocessing (demonstration reprocessing plant, reprocessing head-end colocated at an existing facility); (2) relationship to LWR reprocessing; (3) integrated equipment test (IET) facility and related hardware development activities (mechanical considerations in shearing and dissolving, remote operations and maintenance demonstration phase of IET, integrated process demonstration phase of IET, separate component development activities); and (4) supporting process R and D

  9. Liquid metal corrosion considerations in alloy development

    International Nuclear Information System (INIS)

    Tortorelli, P.F.; DeVan, J.H.

    1984-01-01

    Liquid metal corrosion can be an important consideration in developing alloys for fusion and fast breeder reactors and other applications. Because of the many different forms of liquid metal corrosion (dissolution, alloying, carbon transfer, etc.), alloy optimization based on corrosion resistance depends on a number of factors such as the application temperatures, the particular liquid metal, and the level and nature of impurities in the liquid and solid metals. The present paper reviews the various forms of corrosion by lithium, lead, and sodium and indicates how such corrosion reactions can influence the alloy development process

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

    International Nuclear Information System (INIS)

    Tanaka, Nobuo

    2017-01-01

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

  11. A fast response hydrogen sensor with Pd metallic grating onto a fiber's end-face

    Science.gov (United States)

    Yan, Haitao; Zhao, Xiaoyan; Zhang, Chao; Li, Qiu-Ze; Cao, Jingxiao; Han, Dao-Fu; Hao, Hui; Wang, Ming

    2016-01-01

    We demonstrated an integrated hydrogen sensor with Pd metallic grating fabricated on a fiber end-face. The grating consists of three thin metal layers in stacks, Au, WO3 and Pd. The WO3 is used as a waveguide layer between the Pd and Au layer. The Pd layer is etched by using a focused ion beam (FIB) method, forming a Pd metallic grating with period of 450 nm. The sensor is experimentally exposed to hydrogen gas environment. Changing the concentration from 0% to 4% which is the low explosive limit (LEL), the resonant wavelength measured from the reflection experienced 28.10 nm spectral changes in the visible range. The results demonstrated that the sensor is sensitive for hydrogen detection and it has fast response and low temperature effect.

  12. Stripping scattering of fast atoms on surfaces of metal-oxide crystals and ultrathin films

    International Nuclear Information System (INIS)

    Blauth, David

    2010-01-01

    In the framework of the present dissertation the interactions of fast atoms with surfaces of bulk oxides, metals and thin films on metals were studied. The experiments were performed in the regime of grazing incidence of atoms with energies of some keV. The advantage of this scattering geometry is the high surface sensibility and thus the possibility to determine the crystallographic and electronic characteristics of the topmost surface layer. In addition to these experiments, the energy loss and the electron emission induced by scattered projectiles was investigated. The energy for electron emission and exciton excitation on Alumina/NiAl(110) and SiO 2 /Mo(112) are determined. By detection of the number of projectile induced emitted electrons as function of azimuthal angle for the rotation of the target surface, the geometrical structure of atoms forming the topmost layer of different adsorbate films on metal surfaces where determined via ion beam triangulation. (orig.)

  13. Supplement to Final Environmental Statement related to construction and operation of Clinch River Breeder Reactor Plant, Docket No. 50-537

    International Nuclear Information System (INIS)

    1982-10-01

    In February 1977, the Office of Nuclear Reactor Regulation issued a Final Environmental Statement (FES) (NUREG-0139) related to the construction and operation of the proposed Clinch River Breeder Reactor Plant (CRBRP). Since the FES was issued, additional data relative to the site and its environs have been collected, several modifications have been made to the CRBRP design, and its fuel cycle, and the timing of the plant construction and operation has been affected in accordance with deferments under the DOE Liquid Metal Fast Breeder Reactor (LMFBR) program. These changes are summarized and their environmental significance is assessed in this document. The reader should note that this document generally does not repeat the substantial amount of information in the FES which is still current; hence, the FES should be consulted for a comprehensive understanding of the staff's environmental review of the CRBRP project

  14. SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control

    OpenAIRE

    Lindström, Tobias

    2015-01-01

    In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristic...

  15. Fusion Breeder Program interim report

    International Nuclear Information System (INIS)

    Moir, R.; Lee, J.D.; Neef, W.

    1982-01-01

    This interim report for the FY82 Fusion Breeder Program covers work performed during the scoping phase of the study, December, 1981-February 1982. The goals for the FY82 study are the identification and development of a reference blanket concept using the fission suppression concept and the definition of a development plan to further the fusion breeder application. The context of the study is the tandem mirror reactor, but emphasis is placed upon blanket engineering. A tokamak driver and blanket concept will be selected and studied in more detail during FY83

  16. FAST

    DEFF Research Database (Denmark)

    Zuidmeer-Jongejan, Laurian; Fernandez-Rivas, Montserrat; Poulsen, Lars K.

    2012-01-01

    ABSTRACT: The FAST project (Food Allergy Specific Immunotherapy) aims at the development of safe and effective treatment of food allergies, targeting prevalent, persistent and severe allergy to fish and peach. Classical allergen-specific immunotherapy (SIT), using subcutaneous injections...... with aqueous food extracts may be effective but has proven to be accompanied by too many anaphylactic side-effects. FAST aims to develop a safe alternative by replacing food extracts with hypoallergenic recombinant major allergens as the active ingredients of SIT. Both severe fish and peach allergy are caused...... in depth serological and cellular immune analyses will be performed, allowing identification of novel biomarkers for monitoring treatment efficacy. FAST aims at improving the quality of life of food allergic patients by providing a safe and effective treatment that will significantly lower their threshold...

  17. Lessons learned from the licensing process for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Dickson, P.W.; Clare, G.H.

    1991-01-01

    This paper presents the experience of licensing a specific liquid-metal fast breeder reactor (LMFBR), the Clinch River Breader Reactor Plant (CRBRP). It was a success story in that the licensing process was accomplished in a very short time span. The actions of the applicant and the actions of the US Nuclear Regulatory Commission (NRC) in response are presented and discussed to provide guidance to future efforts to license unconventional reactors. The history is told from the perspective of the authors. As such, some of the reasons given for success or lack of success are subjective interpretations. Nevertheless, the authors' positions provided them an excellent viewpoint to make these judgements. During the second phase of the licensing process, they were the CRBRP Technical Director and the Licensing Manager, respectively, for the Westinghouse Electric Corporation, the prime contractor for the reactor plant

  18. Dependence of heavy metal burnup on nuclear data libraries for fast reactors

    CERN Document Server

    Ohki, S

    2003-01-01

    Japan Nuclear Cycle Development Institute (JNC) is considering the highly burnt fuel as well as the recycling of minor actinide (MA) in the development of commercialized fast reactor cycle systems. Higher accuracy in burnup calculation is going to be required for higher mass plutonium isotopes ( sup 2 sup 4 sup 0 Pu, etc.) and MA nuclides. In the framework of research and development aiming at the validation and necessary improvements of fast reactor burnup calculation, we investigated the differences among the burnup calculation results with the major nuclear data libraries: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2, and JENDL-3.3. We focused on the heavy metal nuclides such as plutonium and MA in the central core region of a conventional sodium-cooled fast reactor. For main heavy metal nuclides ( sup 2 sup 3 sup 5 U, sup 2 sup 3 sup 8 U, sup 2 sup 3 sup 9 Pu, sup 2 sup 4 sup 0 Pu, and sup 2 sup 4 sup 1 Pu), number densities after 1-cycle burnup did not change over one or two percent. Library dependence was re...

  19. Study on Doppler coefficient for metallic fuel fast reactor added hydrogeneous moderator

    Energy Technology Data Exchange (ETDEWEB)

    Hirakawa, Naohiro; Iwasaki, Tomohiko; Tsujimoto, Kazuhumi [Tohoku Univ., Sendai (Japan). Faculty of Engineering; Osugi, Toshitaka; Okajima, Shigeaki; Andoh, Masaki; Nemoto, Tatsuo; Mukaiyama, Takehiko

    1998-01-01

    A series of mock-up experiments for moderator added metallic fast reactor core was carried out at FCA to obtain the experimental verification for improvement of reactivity coefficients. Softened neutron spectrum increases Doppler effect by a factor of 2, and flatter adjoint neutron spectrum decreases Na void effect by a factor of 0.6 when hydrogen to heavy metal atomic number ratio is increased from 0.02 to 0.13. The experimental results are analyzed with SLALOM and CITATION-FBR, which is the standard design code system for a fast reactor at JAERI, and SRAC95 and CITATION-FBR. The present code system gives generally good agreement with the experimental results, especially by the use of the latter, the dependence of the Doppler effect to the hydrogen to fuel element atomic number density ratio is disappeared. Therefore, it looks possible to use the present code system for the conceptual design of a fast reactor system with hydrogeneous materials. (author)

  20. Study of mechanisms and kinetics of Sodium-CO2 interactions. Contribution to the evaluation of an energy conversion system with supercritical CO2 for sodium fast breeder reactors

    International Nuclear Information System (INIS)

    Gicquel, L.

    2010-01-01

    This PhD study consisted in studying reactive mechanisms and kinetics of sodium-CO 2 interactions, in the frame of the assessment of an energy conversion system with supercritical CO 2 for fast breeder reactors cooled by sodium. The approach was the following. First of all, the interactions between sodium and CO 2 have been brought to light by laboratory experiments associated with products analysis. They have enabled the establishment of a coherent mechanism, in agreement with literature data, and gave preliminary indications on the reaction kinetics. In order to estimate a more detailed reaction kinetics, we tried to approach the phenomenon that appears in the case of a leak in a sodium-CO 2 heat exchanger. Geometry of such heat exchangers is not fixed for the moment, even if the development of compact exchangers is foreseen. Then, free jets of CO 2 in liquid sodium have been modeled in order to obtain, by identification, kinetics parameters of the reaction. Those parameters, estimated with such a geometry, will remain valid with a much complex geometry, that will better represent the real exchanger. An experimental bench has been defined and built to realize those jets. The first laboratory experiments have concluded in the existence of different reactive mechanisms according to the temperature level. A threshold has been brought to light around 500 C. Below this one, reaction appears moderated, or even, slow, with a medium exothermicity, and appears after an induction period that depends on the temperature,and which duration could reach several hours. At contrary, above this threshold, it seems rapid and more exothermic. Below 500 C, sodium oxalate is produced, and then reacts with sodium in an exothermic way, following the reactions: CO 2 + Na →1/4 Na 2 C 2 O 4 + 1/4 CO + 1/4 Na 2 CO 3 (5) 4 Na + Na 2 C 2 O 4 → 3 Na 2 O + CO + C (6) Above 500 C, sodium carbonate is produced, and can then possibly react with sodium in an endothermic way, following the

  1. Simple Attenauation Models of Metallic Cables Suitable for G.fast Frequencies

    Directory of Open Access Journals (Sweden)

    Pavel Lafata

    2015-01-01

    Full Text Available Recently, a new xDSL successor called G.fast, which can occupy frequencies up to 106 or 212~MHz, has been introduced in ITU-T G.9700 series of recommendations. Moreover, a new model of transmission characteristics suitable for various types of metallic cables has been designed and described as well. The model is based on 9 parameters specified for each type of metallic cable and can provide accurate estimations. However, its complexity together with the number of required parameters makes its practical application questionable, since the most important metallic cable characteristic, the attenuation, can be estimated using much simpler models. Therefore, two innovative attenuation models suitable for frequencies up to 250 MHz were designed and they will be introduced in this paper. The main motivation was to achieve an accurate approximation of attenuation character for various types of metallic cables, while maintaining low mathematical complexity and a number of necessary parameters. Both models were compared with attenuation characteristics measured for variety types of real metallic cables and also with other standard attenuation models. The results are included in this article as well.

  2. Technology of Fabrication for Sodium-cooled Fast Reactor Metallic Fuel

    International Nuclear Information System (INIS)

    Oh, S. J.; Kim, K. H.; Lee, C. T.; Ryu, H. J.; Ko, Y. M.; Woo, W. M.; Jang, S. J.; Lee, Y. S.; Lee, C. B.

    2008-02-01

    The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr- Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized

  3. Electrolyte additive enabled fast charging and stable cycling lithium metal batteries

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Jianming; Engelhard, Mark H.; Mei, Donghai; Jiao, Shuhong; Polzin, Bryant J.; Zhang, Ji-Guang; Xu, Wu

    2017-03-01

    Lithium (Li) metal battery is an attractive energy storage system owing to the ultrahigh specific capacity and the lowest redox potential of Li metal anode. However, safety concern associated with dendrite growth and limited cycle life especially at a high charge current density are two critical challenges hindering the practical applications of rechargeable Li metal batteries. Here, we report for the first time that an optimal amount (0.05 M) of LiPF6 as additive in the LiTFSI-LiBOB dual-salt/carbonate-based electrolyte can significantly enhance the charging capability and the long-term cycle life of Li metal batteries with a moderately high cathode loading of 1.75 mAh cm-2. Unprecedented stable-cycling (97.1% capacity retention after 500 cycles) along with very limited increase in electrode over-potential has been achieved at a high current density of 1.75 mA cm-2. This unparalleled fast charging and stable cycling performance is contributed from both the stabilized Al cathode current collector, and, more importantly, the robust and conductive SEI layer formed on Li metal anode in the presence of the LiPF6 additive.

  4. Consideration on risk reduction of future breeder reactors

    International Nuclear Information System (INIS)

    Vossebrecker, H.

    1990-09-01

    An overall concept of risk minimization of future sodium-cooled fast breeder reactors is presented in this report. Since shutdown reliability is of vital importance for the breeder safety, a so-called third shutdown level is proposed in addition to the two independent fast shutdown systems. It is basically a group of passive and active measures, which are capable to bring the reactor to safe conditions in all conceivable accident-initiating events and in case of total failure of the two actual shutdown systems. Core disruptions as a result of shutdown failure are therefore beyond the scope of technical imagination. Measures are also foreseen to combat other conceivable causes of core disruption, in particular to achieve residual heat removal with essentially passive systems by making use of the good natural circulation capacity of sodium. On top of that, since absolute safety can never be claimed, damage-limiting containment measures are discussed

  5. Distinguishing Pu Metal from Pu Oxide and Determining alpha-ratio using Fast Neutron Counting

    Energy Technology Data Exchange (ETDEWEB)

    Verbeke, J. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chapline, G. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nakae, L. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Prasad, M. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sheets, S. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Snyderman, N. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-01-07

    We describe a new method for determining the ratio of the rate of (α, n) source neutrons to the rate of spontaneous fission neutrons, the so called α-ratio. This method is made possible by fast neutron counting with liquid scintillator detectors, which can determine the shape of the fast neutron spectrum. The method utilizes the spectral difference between fission spectrum neutrons from Pu metal and the spectrum of (α, n) neutrons from PuO2. Our method is a generalization of the Cifarelli-Hage method for determining keff for fissile assemblies, and also simultaneously determines keff along with the α-ratio.

  6. Status of safety technology for radiological consequence assessment of postulated accidents in liquid metal fast breeder reactors, Canoga Park, California, 29 July--31 July 1975

    International Nuclear Information System (INIS)

    1975-07-01

    State-of-the-art capabilities are examined for prediction and mitigation of radiological consequences of postulated LMFBR accidents. The following topics are treated: radioactive source terms, sodium reactions, aerosol behavior, radiological dose assessment, and engineered safeguards. (U.S.)

  7. Liquid metal fast breeder reactor steam generator: behaviour of heat exchange tubes in face of a through crack resulting in a contact between sodium and water

    International Nuclear Information System (INIS)

    Quinet, J.L.; Lannou, L.

    1978-01-01

    The results of a survey made Electricite de France on the behaviour of cracked tubes under operating conditions of an industrial steam generator are submitted in this communication. A comparison is made of the tube material: INCOLOY 800, 2 1/4 Cr-1 Mo, 9 Cr-2 Mo land to the initial leak. Finally, a description is given of the self-development process of a water leak into sodium. (author)

  8. Spallator - accelerator breeder

    International Nuclear Information System (INIS)

    Steinberg, M.

    1985-01-01

    The concept involves the use of spallation neutrons produced by interaction of a high energy proton (1 to 2 GeV) from a linear accelerator (LINAC) with a heavy metal target (uranium). The principal spallator concept is based on generating fissile fuel for use in LWR nuclear power plants. The spallator functions in conjunction with a reprocessing plant to regenerate and produce the Pu-239 or U-233 for fabrication into fresh LWR reactor fuel elements. Advances in proton accelerator technology has provided a solid base for predicting performance and optimizing the design of a reliable, continuous wave, high-current LINAC required by a fissile fuel production machine

  9. Primary Damage Characteristics in Metals Under Irradiation in the Cores of Thermal and Fast Reactors

    International Nuclear Information System (INIS)

    Pechenkin, V.A.

    2012-01-01

    For an analysis and forecasting of radiation-induced phenomena in structural materials of WWERs, PWRs and BN reactors the fast neutron fluence is usually used (for structural materials of the reactor cores and internals the fluence of neutrons with energy > 0.1 MeV, for WWER and PWRs vessel steels the fluence of neutrons with energy > 0.5 MeV in Russia and East Europe, and with energy > 1.0 MeV in USA and France). Displacements per atom (dpa) seem to be a more appropriate correlation parameter, because it allows comparing the results of materials irradiation in different neutron energy spectra or with different types of particles (neutrons, ions, fast electrons). Energy spectra of primary knocked atoms (PKA) and 'effective' dpa, which are introduced to take into account the point defect recombination during the relaxation stage of a displacement cascade, can be still better representation of the effect of irradiation on material properties. In this work the results of calculating dose rates (dpa/s, NRT-model), PKA energy spectra and PKA mean energies in metals under irradiation in the cores of Russian reactors WWER-440, WWER-1000 (both power thermal reactors) and BN-600 (power fast reactor) and BR-10 (test fast reactor) are presented. In all the reactors Fe and Zr are considered, with addition of Ti and W in BN-600. 'Effective' dose rates in these metals are calculated. Limitations and uncertainties in the standard dpa formulation (the NRT-dpa) are discussed. IPPE activities in the fields related to the TM subject are considered

  10. Preparations for the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Griffin, Jeffrey W.; Peters, Timothy J.; Posakony, Gerald J.; Chien, Hual-Te; Bond, Leonard J.; Denslow, Kayte M.; Sheen, Shuh-Haw; Raptis, Paul

    2009-03-27

    This current report is a summary of information obtained in the "Information Capture" task of the U.S. DOE-funded "Under Sodium Viewing (USV) Project." The goal of the multi-year USV project is to design, build, and demonstrate a state-of-the-art prototype ultrasonic viewing system tailored for periodic reactor core in-service monitoring and maintenance inspections. The study seeks to optimize system parameters, improve performance, and re-establish this key technology area which will be required to support any new U.S. liquid-metal cooled fast reactors.

  12. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Jeong, C. J.; Yang, M. S

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

  13. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    International Nuclear Information System (INIS)

    Park, Jee Won; Jeong, C. J.; Yang, M. S.

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs

  14. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    OpenAIRE

    Chan Bock Lee; Jin Sik Cheon; Sung Ho Kim; Jeong-Yong Park; Hyung-Kook Joo

    2016-01-01

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU)–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochem...

  15. FAST

    Science.gov (United States)

    Nathavitharana, R R; Daru, P; Barrera, A E; Mostofa Kamal, S M; Islam, S; Ul-Alam, M; Sultana, R; Rahman, M; Hossain, Md S; Lederer, P; Hurwitz, S; Chakraborty, K; Kak, N; Tierney, D B; Nardell, E

    2017-09-01

    National Institute of Diseases of the Chest and Hospital, Dhaka; Bangladesh Institute of Research and Rehabilitation in Diabetes, Endocrine and Metabolic Disorders, Dhaka; and Chittagong Chest Disease Hospital, Chittagong, Bangladesh. To present operational data and discuss the challenges of implementing FAST (Find cases Actively, Separate safely and Treat effectively) as a tuberculosis (TB) transmission control strategy. FAST was implemented sequentially at three hospitals. Using Xpert® MTB/RIF, 733/6028 (12.2%, 95%CI 11.4-13.0) patients were diagnosed with unsuspected TB. Patients with a history of TB who were admitted with other lung diseases had more than twice the odds of being diagnosed with unsuspected TB as those with no history of TB (OR 2.6, 95%CI 2.2-3.0, P stakeholder engagement and laboratory capacity are important for sustainability and scalability.

  16. Review of Fast Reactor Activities, March 1980

    International Nuclear Information System (INIS)

    Balz, W.

    1980-01-01

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

  17. Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.

    1985-01-01

    A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Burch, W.D.; Herndon, J.N.; Stradley, J.G.

    1990-01-01

    Research and development in fast reactor reprocessing has been under way about 20 years in several countries throughout the world. During the past decade in France and the United Kingdom, active development programs have been carried out in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the EBR-II facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. Germany and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in all of these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper will focus principally on the search for improved facility concepts and better maintenance systems in the CFRP and, in turn, on how developments at ORNL have influenced the technology elsewhere

  20. Removing coordinated metal ions from proteins: a fast and mild method in aqueous solution.

    Science.gov (United States)

    Carrer, Charlotte; Stolz, Michael; Lewitzki, Erwin; Rittmeyer, Claudia; Kolbesen, Bernd O; Grell, Ernst

    2006-08-01

    Thermodynamic and kinetic studies of metal binding to proteins require the investigation of metal-free proteins, which are often difficult to obtain. We have developed a very fast and mild method to eliminate metal ions from proteins by column chromatography using a commercially available Ni-NTA-type stationary phase. This material, initially designed for protein purification purposes in biotechnology, acts as a strong cation chelator when Ni2+ ions are removed. We have tested this new method with Ca-ATPase, an integral membrane protein exhibiting a strong affinity for Ca2+. By eluting the protein over the Ni2+-free NTA gel, we could remove 95% of the total Ca2+ and obtain an essentially Ca2+-free protein. This method is efficient with only a small amount of NTA gel, and we suggest that it can be applied in general for removal of metal ions from proteins. Moreover, as this procedure can be carried out under mild conditions, the chosen protein kept its enzymatic activity.

  1. Electrolyte additive enabled fast charging and stable cycling lithium metal batteries

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Jianming; Engelhard, Mark H.; Mei, Donghai; Jiao, Shuhong; Polzin, Bryant J.; Zhang, Ji-Guang; Xu, Wu

    2017-03-01

    Batteries using lithium (Li) metal as anodes are considered promising energy storage systems because of their high energy densities. However, safety concerns associated with dendrite growth along with limited cycle life, especially at high charge current densities, hinder their practical uses. Here we report that an optimal amount (0.05 M) of LiPF6 as an additive in LiTFSI-LiBOB dual-salt/carbonate-solvent-based electrolytes significantly enhances the charging capability and cycling stability of Li metal batteries. In a Li metal battery using a 4-V Li-ion cathode at a moderately high loading of 1.75mAh cm(-2), a cyclability of 97.1% capacity retention after 500 cycles along with very limited increase in electrode overpotential is accomplished at a charge/discharge current density up to 1.75 mA cm(-2). The fast charging and stable cycling performances are ascribed to the generation of a robust and conductive solid electrolyte interphase at the Li metal surface and stabilization of the Al cathode current collector.

  2. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-01

    The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

  3. Development of metallic fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Ho; Lee, Chong Yak; Lee, Myung Ho and others

    1999-03-01

    With the vacuum melting and casting of the U-10wt%Zr alloy which is metallic fuel for liquid metal fast breeder reactor, we studied the microstructure of the alloy and the parameters of the melting and casting for the fuel rods. Internal defects of the U-10wt%Zr fuel by gravity casting, were inspected by non-destructive test. U-10wt%Zr alloy has been prepared for the thermal stability test in order to estimate the decomposition of the lamellar structure with relation to swelling under irradiation condition. (author)

  4. Quantitative fuel motion determination with the CABRI fast neutron hodoscope

    International Nuclear Information System (INIS)

    Baumung, K.; Augier, G.

    1991-01-01

    The fast neutron hodoscope installed at the CABRI reactor in Cadarache, France, is employed to provide quantitative fuel motion data during experiments in which single liquid-metal fast breeder reactor test pins are subjected to simulated accident conditions. Instrument design and performance are reviewed, the methods for the quantitative evaluation are presented, and error sources are discussed. The most important findings are the axial expansion as a function of time, phenomena related to pin failure (such as time, location, pin failure mode, and fuel mass ejected after failure), and linear fuel mass distributions with a 2-cm axial resolution. In this paper the hodoscope results of the CABRI-1 program are summarized

  5. Fracture of metals samples under conditions of fast heating by intensive X-ray radiation

    Directory of Open Access Journals (Sweden)

    Golubev V.K.

    2012-08-01

    Full Text Available Results on studying the fracture of metals samples in the form of thin disks under fast heating by the X-ray pulse with the complete spectrum are presented in the paper. The samples of such metals as iron, copper, AMg6 aluminum, VT14 titanium, molybdenum, tungsten, cadmium, lead and zinc were tested. The samples were fixed in the special cartridges equipped with the gauges of a mechanical recoil momentum. The cartridges with samples were placed at such distances from the X-ray irradiator where the energy fluxes were 1.38, 0.90 and 0.29kJ/cm2. The irradiating X-ray pulse was about 2 ns in duration. After testing, the depth of material ablation from a sample front surface and the degree and character of its spall damage were determined. The method of metallographic analysis was used for these purposes. Numerical calculations of loading conditions were made with the use of an equation of state taking into account the process of evaporation. The calculated value of maximum negative pressure in the sample at the coordinate corresponding to the formation of spallation zones or spall cracks was conventionally accepted as the material resistance to spall fracture. The comparison of obtained results with the data on the fracture of examined materials in the conditions of fast heating by the X-ray pulse with the hard spectrum and a high-current electron beam was conducted.

  6. Consequence analysis of core meltdown accidents in liquid metal fast reactor

    International Nuclear Information System (INIS)

    Suk, S.D.; Hahn, D.

    2001-01-01

    Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)

  7. Delineation of First-Order Elastic Property Closures for Hexagonal Metals Using Fast Fourier Transforms

    Directory of Open Access Journals (Sweden)

    Nicholas W. Landry

    2015-09-01

    Full Text Available Property closures are envelopes representing the complete set of theoretically feasible macroscopic property combinations for a given material system. In this paper, we present a computational procedure based on fast Fourier transforms (FFTs to delineation of elastic property closures for hexagonal close packed (HCP metals. The procedure consists of building a database of non-zero Fourier transforms for each component of the elastic stiffness tensor, calculating the Fourier transforms of orientation distribution functions (ODFs, and calculating the ODF-to-elastic property bounds in the Fourier space. In earlier studies, HCP closures were computed using the generalized spherical harmonics (GSH representation and an assumption of orthotropic sample symmetry; here, the FFT approach allowed us to successfully calculate the closures for a range of HCP metals without invoking any sample symmetry assumption. The methodology presented here facilitates for the first time computation of property closures involving normal-shear coupling stiffness coefficients. We found that the representation of these property linkages using FFTs need more terms compared to GSH representations. However, the use of FFT representations reduces the computational time involved in producing the property closures due to the use of fast FFT algorithms. Moreover, FFT algorithms are readily available as opposed to GSH codes.

  8. Fusion breeder: its potential role and prospects

    International Nuclear Information System (INIS)

    Lee, J.D.

    1981-01-01

    The fusion breeder is a concept that utilizes 14 MeV neutrons from D + T → n(14.1 MeV) + α(3.5 MeV) fusion reactions to produce more fuel than the tritium (T) needed to sustain the fusion process. This excess fuel production capacity is used to produce fissile material (Pu-239 or U-233) for subsequent use in fission reactors. We are concentrating on a class of blankets we call fission suppressed. The blanket is the region surrounding the fusion plasma in which fusion neutrons interact to produce fuel and heat. The fission-suppressed blanket uses non-fission reactions (mainly (n,2n) or (n,n't)) to generate excess neutrons for the production of net fuel. This is in contrast to the fast fission class of blankets which use (n,fiss) reactions to generate excess neutrons. Fusion reactors with fast fission blankets are commony known as fusion-fission hybrids because they combine fusion and fission in the same device

  9. Fast reactors with axial arrangement of oxide and metal fuels in the core

    International Nuclear Information System (INIS)

    Troyanov, M.F.; Ilyunin, V.G.; Matveev, V.I.; Murogov, V.M.; Proshkin, A.A.; Rudneva, V.Ya.; Shmelev, A.N.

    1980-01-01

    Problems of using metal fuel in fast reactor (FR) core are discussed Results are given of the calculation of two-dimentional (R-Z) FR version having a composed core with the combined usage of oxide and metal fuels having parameters close to optimal from the point of view of fuel breeding rate, an oxide subzone having increased enrichment and a decreased proper conversion ratio. A reactor is considered where metallic fuel elements are placed from the side of ''cold'' coolant inlet (400-480 deg C), and oxide fuel elements - in the region where the coolant has a higher temperature (500-560 deg C). It is shown that the new fuel breeding rate in such a reactor can be increased by 20-30% as compared with an oxide fuel reactor. Growth of the total conversion ratio is mainly stipulated with the increase of the inner conversion ratio of the core (CRC) which is important not only from the point of view of nuclear fuel breeding rate but also the optimization of the mode of powerful fast reactor operation with provision for the change in reactivity in the process of its continuous operation. The fact, that the core version under investigation has a CRC value slightly exceeding unit, stipulates considerably less reactivity change as compared with the oxide version in the process of the reactor operation and permits at a constant reactor control system power to significantly increase the time between reloadings and, therefore, to increase the NPP load factor which is of great importance both from the point of view of economy and the improvement of operation conditions as well as of reactor operation reliability. It is concluded on the base of the analysis of the results obtained that FRs with the combined usage of oxide and metal fuels having an increased specific load and increased conversion ratio as compared with the oxide fuel FRs provide a higher rate of development of the whole nuclear power balanced with respect to the fuel [ru

  10. Test reactor: basic to U.S. breeder reactor development

    International Nuclear Information System (INIS)

    Miller, B.J.; Harness, A.J.

    1975-01-01

    Long-range energy planning in the U. S. includes development of a national commercial breeder reactor program. U. S. development of the LMFBR is following a conservative sequence of extensive technology development through use of test reactors and demonstration plants prior to construction of commercial plants. Because materials and fuel technology development is considered the first vital step in this sequence, initial U. S. efforts have been directed to the design and construction of a unique test reactor. The Fast Flux Test Facility, FFTF, is a 400 MW(t) reactor with driver fuel locations, open test locations, and closed loops for higher risk experiments. The FFTF will provide a prototypic LMFBR core environment with sufficient instrumentation for detailed core environmental characterization and a testing capability substituted for breeder capability. The unique comprehensive fuel and materials testing capability of the FFTF will be key to achieving long-range objectives of increased power density, improved breeding gain and shorter doubling times. (auth)

  11. A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2017-06-26

    A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.

  12. Status of Liquid Metal Fast Reactor Development in the United States of America, March 1987

    International Nuclear Information System (INIS)

    Horton, K.E.

    1987-01-01

    In order to meet the objective to develop and demonstrate economically competitive reactor designs and associated fuel cycles early in the next century, the U.S. program has become more focused. Two innovative reactor designs supported by the metal-fueled Integral Fast Reactor program are being directed at fulfilling a series of advanced reactor goals. The supporting technology programs and facilities are being refocused to support the overall goals. International collaboration is being broadened to provide the two-way support across the spectrum of plant projects and the fuel cycle. This program is intended to maintain the technology base into the time period (mid-1990s) when a private sector demonstration could be initiated. (author)

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

    International Nuclear Information System (INIS)

    1996-04-01

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

  14. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Chan Bock Lee

    2016-10-01

    Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  15. Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  16. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    2012-01-01

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

  17. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    2013-01-01

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

  18. Heaters to simulate fuel pins for heat transfer tests in single-phase liquid-metal-flow

    International Nuclear Information System (INIS)

    Casal, V.; Graf, E.; Hartmann, W.

    1976-09-01

    The development of heaters for thermal simulation of the fuel elements of liquid metal cooled fast breeder reactors (SNR) is reported. Beginning with the experimental demands various heating methods are discussed for thermodynamic investigations of the heat transfer in liquid metals. Then a preferred heater rod is derived to simulate the fuel pins of a SNR. Finally it is reported on the fabrication and the operation practice. (orig.) [de

  19. in meat production III. Feeder - breeder dimorphism

    African Journals Online (AJOL)

    Feeder- breeder dimorphism is advantageous when large offspring for slaughter is obtained from small breeding animals. The effect of feeder- breeder dimorphism on herd efficiency is evaluated for terminal crossbreeding and growth modification by biotechnological or dietary means. Selection criteria for breeds or lines in ...

  20. Semi-metallic, strong conductive polymer microfiber, method and fast response rate actuators and heating textiles

    KAUST Repository

    Zhou, Jian

    2016-06-09

    A method comprising: providing at least one first composition comprising at least one conjugated polymer and at least one solvent, wet spinning the at least one first composition to form at least one first fiber material, hot-drawing the at least one fiber to form at least one second fiber material. In lead embodiments, high-performance poly(3,4-ethylenedioxy- thiophene)/poly(styrenesulfonate) (PEDOT/PSS) conjugated polymer microfibers were fabricated via wet- spinning followed by hot-drawing. In these lead embodiments, due to the combined effects of the vertical hot-drawing process and doping/de-doping the microfibers with ethylene glycol (EG), a record electrical conductivity of 2804 S · cm-1 was achieved. This is believed to be a six-fold improvement over the best previously reported value for PEDOT/PSS fibers (467 S · cm-1) and a twofold improvement over the best values for conductive polymer films treated by EG de-doping (1418 S · cm-1). Moreover, these lead, highly conductive fibers experience a semiconductor-metal transition at 313 K. They also have superior mechanical properties with a Young\\'s modulus up to 8.3 GPa, a tensile strength reaching 409.8 MPa and a large elongation before failure (21%). The most conductive fiber also demonstrates an extraordinary electrical performance during stretching/unstretching: the conductivity increased by 25% before the fiber rupture point with a maximum strain up to 21%. Simple fabrication of the semi-metallic, strong and stretchable wet-spun PEDOT/PSS microfibers can make them available for conductive smart electronics. A dramatic improvement in electrical conductivity is needed to make conductive polymer fibers viable candidates in applications such as flexible electrodes, conductive textiles, and fast-response sensors and actuators.

  1. Thermal breeder fuel enrichment zoning

    Science.gov (United States)

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  2. Tritium transport and release from lithium ceramic breeder materials

    International Nuclear Information System (INIS)

    Johnson, C.E.; Kopasz, J.P.; Tam, S.W.

    1994-01-01

    In an operating fusion reactor,, the tritium breeding blanket will reach a condition in which the tritium release rate equals the production rate. The tritium release rate must be fast enough that the tritium inventory in the blanket does not become excessive. Slow tritium release will result in a large tritium inventory, which is unacceptable from both economic and safety viewpoints As a consequence, considerable effort has been devoted to understanding the tritium release mechanism from ceramic breeders and beryllium neutron multipliers through theoretical, laboratory, and in-reactor studies. This information is being applied to the development of models for predicting tritium release for various blanket operating conditions

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

    International Nuclear Information System (INIS)

    King, T.L.; Landry, R.R.; Throm, E.D.; Wilson, J.N.

    1991-12-01

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

  4. Transient bowing of core assemblies in advanced liquid metal fast reactors

    International Nuclear Information System (INIS)

    Kamal, S.A.; Orechwa, Y.

    1986-01-01

    Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety

  5. Soft x-ray continuum radiation transmitted through metallic filters: An analytical approach to fast electron temperature measurements

    International Nuclear Information System (INIS)

    Delgado-Aparicio, L.; Hill, K.; Bitter, M.; Tritz, K.; Kramer, T.; Stutman, D.; Finkenthal, M.

    2010-01-01

    A new set of analytic formulas describes the transmission of soft x-ray continuum radiation through a metallic foil for its application to fast electron temperature measurements in fusion plasmas. This novel approach shows good agreement with numerical calculations over a wide range of plasma temperatures in contrast with the solutions obtained when using a transmission approximated by a single-Heaviside function [S. von Goeler et al., Rev. Sci. Instrum. 70, 599 (1999)]. The new analytic formulas can improve the interpretation of the experimental results and thus contribute in obtaining fast temperature measurements in between intermittent Thomson scattering data.

  6. SACO-1: a fast-running LMFBR accident-analysis code

    International Nuclear Information System (INIS)

    Mueller, C.J.; Cahalan, J.E.; Vaurio, J.K.

    1980-01-01

    SACO is a fast-running computer code that simulates hypothetical accidents in liquid-metal fast breeder reactors to the point of permanent subcriticality or to the initiation of a prompt-critical excursion. In the tradition of the SAS codes, each subassembly is modeled by a representative fuel pin with three distinct axial regions to simulate the blanket and core regions. However, analytic and integral models are used wherever possible to cut down the computing time and storage requirements. The physical models and basic equations are described in detail. Comparisons of SACO results to analogous SAS3D results comprise the qualifications of SACO and are illustrated and discussed

  7. Instrumentation and control improvements at Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I ampersand C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I ampersand C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I ampersand C systems of the next generation of liquid metal reactor (LMR) plants

  8. Experience with EBR-II [Experimental Breeder Reactor] driver fuel

    International Nuclear Information System (INIS)

    Seidel, B.R.; Porter, D.L.; Walters, L.C.; Hofman, G.L.

    1986-01-01

    The exceptional performance of Experimental Breeder Reactor-II (EBR-II) metallic driver fuel has been demonstrated by the irradiation of a large number of elements under steady-state, transient overpower, and loss-of-flow conditions. High burnup with high reliability has been achieved by a close coupling of element design and materials selection. Quantification of reliability has allowed full utilization of element lifetime. Improved design and duct materials currently under test are expected to increase the burnup from 8 to 14 at.%

  9. Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

    International Nuclear Information System (INIS)

    2003-01-01

    In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

  10. Development of an innovative plate-type SG for fast breeder reactor. Proposal of the concept and the evaluation of the fabricating method by the test fabrication of the partial model

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Kinoshita, Izumi

    2006-01-01

    The concept of an innovative plate type SG for the fast reactor fabricated by using the HIP (Hot Isostatic Pressing) method was proposed. The heat transfer plate, which is assembled with rectangular tubes and is fabricated by HIP method, is surrounded by leakage detection spaces. It is possible to apply it to both the pool-type and the loop-type LMFR. In this report, the fabrication technique was studied about the concept for the loop-type LMFR, and the following results were obtained. Hip tests, tensile tests, and structure observation were performed to clarify the suitable HIP condition for the modified 9Cr-1Mo steel. As a result, the optimum condition of 1150 deg C x 1200 kgf/cm 2 x 3 hr was found. Nickel-type solder (BNi-5) and gold-type solder (BAu-4) were selected as a joining material to laminate the heat transfer tube plates. Through the comparison of tensile tests, BAu-4 that showed a more excellent joining performance was selected on the assumption of the margin of 5 mm from the welding line. After buckling load had been clarified, the BAu-4 brazing of the heat transfer tube plates was performed using a hot pressing method. Problems were not observed in the welding of simulated header, and in the fabricating of the partial model of SG. (author)

  11. The integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1990-01-01

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

  12. Major welfare issues in broiler breeders

    NARCIS (Netherlands)

    Jong, de I.C.; Guemene, D.

    2011-01-01

    Under current practices, broiler parent stock (broiler breeders) encounter several welfare problems, such as feed restriction and injury during mating. Intensive selection for production traits, especially growth rate, is associated with increased nutritious requirement and thus feed consumption,

  13. A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

    Directory of Open Access Journals (Sweden)

    Tanju Sofu

    2015-04-01

    Full Text Available The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel–coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

  14. Status and prospects of thermal breeding. Preliminary report by Thermal Breeder Consultants Group

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Thermal breeders like the liquid-fuel reactors, the seed-and-blanket reactor or the CANDU thorium reactor have been reviewed by an IAEA Consultants Group to evaluate in how far large-scale nuclear energy production for mankind is possible in the future. 'Large scale' is defined as 60% of the total energy consumption in 2050. A future average consumption rate for mankind of 5kW/capita and a population growth according to UN estimates results in an energy growth rate for mankind of still a factor of 8. The paper shows that only with liquid-fuel thermal breeders - either aqueous suspensions or molten salts - and under certain conditions also with fast breeder reactors (very fast reprocessing) is it possible to supply 60% of this energy requirement by nuclear means in 2050. Other breeder systems cannot meet the requirements because they demand too large a fissile inventory. Finally, it is shown that a combination of both systems, thermal liquid-fuel and fast reactors, is always better than one system separately. (author)

  15. Fast-release orodispersible tramadol as analgesia in hysterosalpingography with a metal cannula or a balloon catheter.

    Science.gov (United States)

    Stoop, D; De Brucker, M; Haentjens, P; Talebian, A; de Mey, J; Devroey, P

    2010-06-01

    This study aimed to evaluate the potential benefit, in terms of pain relief, of the new oral fast-release orodispersible galvanic form of tramadol in women undergoing hysterosalpingography (HSG) with either a metal cannula or a balloon catheter. In a randomized, double-blind, placebo-controlled, 2 x 2 factorial-design trial, conducted at a single academic centre, 128 women were assigned into groups: (I) tramadol and a metal cannula, (II) tramadol and a balloon catheter, (III) placebo and a metal cannula or (IV) placebo and a balloon catheter. The primary end-point was pain registered by the patients on 10-cm visual analogue scales (VASs) at various times during and after the procedure. Secondary end-points included side effects and pain as assessed by the same physician during HSG. The main effect of tramadol versus placebo medication (i.e. I and II versus III and IV) was a statistically significant difference (P effect of the balloon catheter versus metal cannula (i.e. II and IV versus I and III) was a non-significant (P = 0.82) difference in patient-reported VAS of -0.05 (-0.49 to +0.39) and -2% (-21% to +17%). There were no medication-HSG device interactions and no differences in side effects. During and after HSG, fast-release orodispersible tramadol significantly reduces pain without increasing side effects.

  16. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2000-03-01

    Feasibility study of the liquid-metal MHD power generation system combined with the high-density two-phase natural circulation has been performed for the applicability to the simple, autonomic energy conversion system of the liquid-metal cooled fast reactor. The present system has many promising aspects not only in the energy conversion process, but also in safety and economical improvements of the liquid-metal cooled fast reactor. For example, the high cycle efficiency can be expected because of the similarity of the present cycle to the Ericsson cycle. Sodium-Water Interaction problem can be excluded by proper combination of the working fluids. As the economical feature, the present system is so simple that the liquid-metal main circular pump, the steam turbine generator, and even the steam generator can be excluded if the thermodynamic working fluid is injected directly into the high temperature liquid metal MHD working fluid. In addition, the present system has the potential to be applied to various heat sources including solar energy because of the high flexibility of the operation temperature. In the present paper, as the first step of the feasibility study, the cycle analyses were performed to examine the effects of the main system parameters on the fundamental characteristics of the system. It is found that the cycle efficiency of the present system is enough competitive with that of the conventional steam turbine system. It is, however, found that the cycle efficiency depends strongly on the gas-liquid slip ratio in the two-phase flow channel. As the conclusions, it is recommended to perform experimental study to obtain the fundamental data, such as the gas-liquid slip ratio in the high-density liquid-metal two-phase natural circulation. (author)

  17. Decay heat removal analyses in heavy-liquid-metal-cooled fast breeding reactors. Development of the thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors

    International Nuclear Information System (INIS)

    Sakai, Takaaki; Enuma, Yasuhiro; Ohyama, Kazuhiro

    2001-05-01

    The feasibility study on future commercial fast breeder reactors in Japan has been conducted at JNC, in which various plant design options with all the possible coolant and fuel types are investigated to determine the conditions for the future detailed study. Lead-bismuth eutectic coolant has been selected as one of the possible coolant options. During the phase-I activity of the feasibility study in FY1999 and FY2000, several plant concepts, which were cooled by the heavy liquid metal coolant, were examined to evaluate the feasibility mainly with respect to economical competitiveness with other coolant reactors. A medium-scale (300 - 550 MWe) plant, cooled by a lead-bismuth natural circulation flow in a pool type vessel, was selected as the most possible plant concept for the heavy liquid metal coolant. Thus, a conceptual design study for a lead-bismuth-cooled, natural-circulation reactor of 400 MWe has been performed at JNC to identify remaining difficulties in technological aspect and its construction cost evaluation. In this report, a thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors is described. A Multi-dimensional Steam Generator analysis code (MSG) was applied to evaluate the natural circulation plant by combination with a flow-network-type, plant dynamics code (Super-COPD). By using this combined multi-dimensional plant dynamics code, decay heat removals, ULOHS and UTOP accidents were evaluated for the 100 MWe STAR-LM concept designed by ANL. In addition, decay heat removal by the Primary Reactor Auxiliary Cooling System (PRACS) in the 400 MWe lead-bismuth-cooled, natural-circulation reactor, being studied at JNC, was analyzed. In conclusion, it becomes clear that the combined multi-dimensional plant dynamics code is suitably applicable to analyses of lead-bismuth-cooled, natural-circulation reactors to evaluate thermal-hydraulic phenomena during steady-state and transient conditions. (author)

  18. Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

    International Nuclear Information System (INIS)

    You, WuSeung; Hong, Ser Gi

    2014-01-01

    The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

  19. An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

    Directory of Open Access Journals (Sweden)

    Wuseong You

    2017-12-01

    Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

  20. DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

  1. Prospects of ceramic tritium breeder materials

    International Nuclear Information System (INIS)

    Roth, E.; Roux, N.; Conservatoire National des Arts et Metiers; CEA Centre d'Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette

    1989-01-01

    In this paper the authors examine the prospects of the main ceramics proposed as breeder materials for fusion reactors, i.e. Li-2O, Li-2ZrO-3, LiAlO-2, Li-4SiO-4. To do so they review terms of reference of contemplated blankets for NET, ITER and DEMO, and the proposed blanket concepts and materials. Issues respective to the use of each breeder material are examined, and from this review it is concluded that ceramics are the most favorable breeder materials whose use can be contemplated as well for a driver blanket for NET or ITER and for a DEMO blanket. Ceramics are then compared between themselves and it is seen that, subject to the confirmation of recent experimental results, lithium zirconate could be used with advantage in any of the present blanket concepts, except in those employing lithium at its natural isotopic abundance, in which case only Li-2O can be used. However in specific cases, or in parts of a blanket, other ceramics may be profitably employed. As a general conclusion suggestions are made to further improve ceramic breeder performances, and it is recommended to intensify also work on problems that have to be solved in order to operate ceramic breeder blankets e.g. tritium extraction and recovery systems and conditions of beryllium use. (author). 37 refs.; 12 tabs

  2. Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

    International Nuclear Information System (INIS)

    1994-01-01

    The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned

  3. Experimental program for the Fast Breeder Blanket Facility, FBBF

    International Nuclear Information System (INIS)

    Ott, K.O.; Clikeman, F.M.; Johnson, R.H.; Borg, R.C.

    1976-01-01

    The work performed in the reporting period was primarily concerned with the development of the experimental program (Task A) and with the pre-analysis of future loadings and the impact upon the permanent loading of the two converter regions, which contain 4.8 percent enriched UO 2 rods. It appears necessary that a neutron poison (B 4 C) be placed in the converter (transformer) regions in order to hold, also for future loadings, the k/sub eff/ of a hypothetically flooded FBBF well below 1. Since it is planned to use the same welded converter regions for all experiments, the required B 4 C loading needs to be determined prior to the first blanket loading. Further the equipment needs have been identified (Task D), the 252 Cf-source has been requested on a loan basis (Task E). First discussions with ANL on blanket experiments have been initiated

  4. 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

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

    Indian Academy of Sciences (India)

    gas transportation through heat exchangers (IHX) and pump to grid plate, agglomeration of gas inside grid plate ... During plant transients associated with one secondary loop, temperature in the cold pool associated with ... Flow coast down characteristics of secondary sodium pump (SSP in figure 1) and cold pool capacity ...

  6. 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.)

  7. Thermodynamic analysis of advanced fuels for fast breeder reactors

    International Nuclear Information System (INIS)

    Srivastava, D.; Garg, S.P.; Goswami, G.L.

    1990-01-01

    Six phase fields of interest in the M-C-N system (M= mixed U/Pu) with oxygen as impurity are i) U 1-x3 Pu x3 (=M)+ U 1-x1 Pu x1 C 1-y-z N y O z (= MCN O), ii)C+ U 1 x2 Pu x2 Csub(1.5) (=MCsub(1.5)), iii) MCsub(1.5) + MCNO, iv) C+MCNO, v) UN (1.5) + MCNO and vi) C + UNsub(1.5) + MCNO. In the present work a detailed thermodynamic analysis has been carried out for all the six phase fields existing in the system with x 1 , 1-y-z and y are varying from 0.0 to 1.0 and z as impurity from 0.0 to 0.15 at temperature between 1500K to 2000K. In the first part, composition of the phases in the different phase fields have been calculated as a function of overall composition of the fuel and temperature. In the second part, thermodynamic properties such as partial pressures of N 2 (g), O 2 (g), CO(g), Pu(g), U(g), PuO(g), UO(g), UC 2 (g) and PuC 2 (g) species and carbon potential of the fuel have been calculated as a function of compositions x 1 , y and z at different temperatures. Results obtained are discus sed in detail and compared with the reported measured data. Hitherto, thermodynamic properties for all the phase fields of M-C-N-O system have not been reported. (a uthor). 54 tabs., 13 figs., 24 refs

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

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    The latter occurs especially during transient over-power (TOP) incidents. The wrapper tubes of the ... generator a critical component in determining the efficient running of the plant and demands high integrity of steam ... diffusivity of the fuel, which was measured at BARC as well as at IGCAR, by a technique employing the ...

  9. Overview of pool hydraulic design of Indian prototype fast breeder ...

    Indian Academy of Sciences (India)

    plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in ... (thermal fatigue) due to various incidents taking place in the plant. High thermal expansion coefficient and low thermal .... prediction of temperature fluctuations by special modelling techniques.

  10. 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)

  11. 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

  12. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    Abstract. In a thermal neutron reactor, multiple recycle of U–Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving ...

  13. The fast breeder reactor and the electricity provision system

    International Nuclear Information System (INIS)

    1975-01-01

    Due to inflexible planning, the electricity provision system is bound to generate demands for big centralized power plants, that is, nuclear power plants. A lack of feed-back from producer to consumer results in a technical distribution net-work that cannot react to big changes in supply and demand. This makes it particularly difficult to include alternative energy sources with fluctuating energy production

  14. The streaming effect in gas-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Collussi, I.

    The importance of neutron streaming in the GCFR is evaluated by taking into consideration the anisotropy due to coolant and control rod channels. Calculation is done using a numerical-analytical method developed in this paper and compared with results obtained using the methods of Benoist and Ligou. Comparison of the results obtained by these three methods shows that streaming effect is strongly dependent on the axial buckling 'B 2 2 '. The influence of neutron streaming on the reactivity is shown to be negligilbe and, in consequence, the GCFRs may be considered homogeneous to a good approximation. For accurate calculation the neutron streaming should be considered, mainly for radiation damage and shielding calculation [pt

  15. Design safety limits in prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Puthiyavinayagam, P.; Roychowdhury, D.G.; Govindarajan, S.; Chellapandi, P.; Singh, Om Pal; Chetal, S.C.

    2002-01-01

    Full text: PFBR is designed to operate at 450 W/cm peak linear heat rating to a peak burn up of 100,000 MWd/t which corresponds to a damage dose of 85 dpa. The targetted reliability is to restrict pin failure to 1 in 10,000. All the design basis events are classified into four categories. Design safety limits imposed for DBE are in terms of temperatures, radiation doses and structural design parameters. Radiation limits are imposed in relation to RCB from the plant personnel and public point of view. Fuel pin integrity is assured with a detailed damage analysis by adopting cumulative damage concept for fixing clad temperature limits. Fuel temperatures are limited to melting point to preclude fuel slumping for events up to category 3. Partial melting is allowed for events in category 4 and the results obtained from transients experiments show that partial melting up to 50% of pellet area does not result in clad failure. Coolant temperatures are limited to boiling point to avoid burnout and reactivity effects

  16. 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

  17. 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)

  18. Fast mixed spectrum reactor concept

    International Nuclear Information System (INIS)

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

    1979-04-01

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

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

    International Nuclear Information System (INIS)

    2015-01-01

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

  20. Evolution of fast reactor core spectra in changing a heavy liquid metal coolant by molten PB-208

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, D. A.; Mitenkova, E. F. [Nuclear Safety Inst., Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Khorasanov, G. L.; Zemskov, E. A.; Blokhin, A. I. [State Scientific Center, Russian Federation, Inst. of Physics and Power Engineering, Bondarenko Square 1, Obninsk, 249033 (Russian Federation)

    2012-07-01

    In the paper neutron spectra of fast reactor cooled with lead-bismuth or lead-208 are given. It is shown that in changing the coolant from lead-bismuth to lead-208 the core neutron spectra of the fast reactor FR RBEC-M are hardening in whole by several percents when a little share of low energy neutrons (5 eV - 50 keV) is slightly increasing. The shift of spectra to higher energies permits to enhance the fuel fission while the increased share of low energy neutrons provides more effective conversion of uranium-238 into plutonium due to peculiarity of {sup 238}U neutron capture cross section. Good neutron and physical features of molten {sup 208}Pb permit to assume it as perspective coolant for fast reactors and accelerator driven systems. The one-group cross sections of neutron radiation capture, {sigma}(n,g), by {sup 208}Pb, {sup 238}U, {sup 99}Tc, mix of lead and bismuth, {sup nat}Pb-Bi, averaged over neutron spectra of the fast reactor RBEC-M are given. It is shown that one-group cross sections of neutron capture by material of the liquid metal coolant consisted from lead enriched with the stable lead isotope, {sup 208}Pb, are by 4-7 times smaller {sigma}(n,g) for the coolant {sup nat}Pb-Bi. The economy of neutrons in the core cooled with {sup 208}Pb can be used for reducing reactor's initial fuel load, increasing fuel breeding and transmutation of long lived fission products, for example {sup 99}Tc. Good neutron and physical features of lead enriched with {sup 208}Pb permit to consider it as a perspective low neutron absorbing coolant for fast reactors and accelerator driven systems. (authors)

  1. Development of a fast thermal response microfluidic system using liquid metal

    International Nuclear Information System (INIS)

    Gao, Meng; Gui, Lin

    2016-01-01

    Room temperature liquid metal gallium alloy has been widely used in many micro-electromechanical systems applications, such as on-chip electrical microheaters, micro temperature sensors, micro pumps and so on. Injecting liquid metal into microchannels can provide a simple, rapid, low-cost but efficient way to integrate these elements in microfluidic chips with high accuracy. The liquid metal-filled microstructures can be designed in any shape and easily integrated into microfluidic chips. In this paper, an on-chip liquid metal-based thermal microfluidic system is proposed for quick temperature control at the microscale. The micro system utilizes just one microfluidic chip as a basic working platform, which has liquid metal-based on-chip heaters, temperature sensors and electroosmotic flow pumps. Under the comprehensive control of these elements, the micro system can quickly change the temperature of a target fluid in the microfluidic chip. These liquid metal-based on-chip elements are very helpful for the fabrication and miniaturization of the microfluidic chip. In this paper, deionized water is used to test the temperature control performance of the thermal microfluidic system. According to the experimental results, the micro system can efficiently control the temperature of water ranging from 28 °C to 90 °C. The thermal microfluidic system has great potential for use in many microfluidic applications, such as on-chip polymerase chain reaction, temperature gradient focusing, protein crystallization and chemical synthesis. (paper)

  2. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. Modelling energy utilisation in broiler breeder hens.

    Science.gov (United States)

    Rabello, C B V; Sakomura, N K; Longo, F A; Couto, H P; Pacheco, C R; Fernandes, J B K

    2006-10-01

    1. The objective of this study was to determine a metabolisable energy (ME) requirement model for broiler breeder hens. The influence of temperature on ME requirements for maintenance was determined in experiments conducted in three environmental rooms with temperatures kept constant at 13, 21 and 30 degrees C using a comparative slaughter technique. The energy requirements for weight gain were determined based upon body energy content and efficiency of energy utilisation for weight gain. The energy requirements for egg production were determined on the basis of egg energy content and efficiency of energy deposition in the eggs. 2. The following model was developed using these results: ME = kgW0.75(806.53-26.45T + 0.50T2) + 31.90G + 10.04EM, where kgW0.75 is body weight (kg) raised to the power 0.75, T is temperature ( degrees C), G is weight gain (g) and EM is egg mass (g). 3. A feeding trial was conducted using 400 Hubbard Hi-Yield broiler breeder hens and 40 Peterson males from 31 to 46 weeks of age in order to compare use of the model with a recommended feeding programme for this strain of bird. The application of the model in breeder hens provided good productive and reproductive performance and better results in feed and energy conversion than in hens fed according to strain recommendation. In conclusion, the model evaluated predicted an ME intake which matched breeder hens' requirements.

  4. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

    2012-01-01

    CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

  5. Fast programming metal-gate Si quantum dot nonvolatile memory using green nanosecond laser spike annealing

    Science.gov (United States)

    Lien, Yu-Chung; Shieh, Jia-Min; Huang, Wen-Hsien; Tu, Cheng-Hui; Wang, Chieh; Shen, Chang-Hong; Dai, Bau-Tong; Pan, Ci-Ling; Hu, Chenming; Yang, Fu-Liang

    2012-04-01

    The ultrafast metal-gate silicon quantum-dot (Si-QD) nonvolatile memory (NVM) with program/erase speed of 1 μs under low operating voltages of ± 7 V is achieved by thin tunneling oxide, in situ Si-QD-embedded dielectrics, and metal gate. Selective source/drain activation by green nanosecond laser spike annealing, due to metal-gate as light-blocking layer, responds to low thermal damage on gate structures and, therefore, suppresses re-crystallization/deformation/diffusion of embedded Si-QDs. Accordingly, it greatly sustains efficient charge trapping/de-trapping in numerous deep charge-trapping sites in discrete Si-QDs. Such a gate nanostructure also ensures excellent endurance and retention in the microsecond-operation Si-QD NVM.

  6. Analytical modeling of core hydraulics and flow management in breeder reactors

    International Nuclear Information System (INIS)

    Carelli, M.D.; Willis, J.M.

    1979-01-01

    An analytical model representing the hydraulic behavior of the primary system of fast breeder nuclear reactors is discussed. A computer code capable of detailing the core flow distribution and characterizing the flow and pressure drop in each reactor component is presented. Application of this method to the reactor core thermal-hydraulic design has allowed optimization of the flow management with consequent upgrading in performance, reduction of unnecessary conservatism and very substantial cost savings. Typical quantitative examples are presented

  7. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    International Nuclear Information System (INIS)

    Kim, Y. C.; Na, B. C.; Hahn, D. H.

    1997-04-01

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  8. Techniques developed to evaluate the fracture toughness offast breeder reactor duct. [Use of J-integral tests

    Energy Technology Data Exchange (ETDEWEB)

    Huang, F. H.; Wire, G. L.

    1979-01-01

    Large changes in strength and ductility of metals after irradiation are known to occur. The fracture toughness of irradiated metals, which is related to the combined strength and ductility of a material, may be significantly reduced and the potential for unstable crack extension increased. Therefore, the resistance of cladding and duct materials to fracture after exposure to fast neutron environments is of concern. Existing Type 316 stainless steel irradiated ducts are relatively thin and since this material retains substantial ductility, even after irradiation, the fracture behavior of the duct material cannot be analyzed by linear elastic fracture mechanics techniques. Instead, the multispecimen R-curve method and J-integral analysis were used to develop an experimental approach to evaluate the fracture toughness of thin breeder reactor duct materials irradiated at elevated temperatures. Alloy A-286 was chosen for these experiments because the alloy exhibits elastic/plastic behavior and the fracture toughness data of thicker (12 mm) specimens were available for comparison. Technical problems associated with specimen buckling and remote handling were treated in this work. The results are discussed in terms of thickness criterion for plane strain.

  9. Simple and fast fabrication of superhydrophobic metal wire mesh for efficiently gravity-driven oil/water separation.

    Science.gov (United States)

    Song, Botao

    2016-12-15

    Superhydrophobic metal wire mesh (SMWM) has frequently been applied for the selective and efficient separation of oil/water mixture due to its porous structure and special wettability. However, current methods for the modification of metal wire mesh to be superhydrophobic suffered from problems with respect to complex experimental procedures or time-consuming process. In this study, a very simple, time-saving and single-step electrospray method was proposed to fabricate SMWM and the whole procedure required about only 2min. The morphology, surface composition and wettability of the SMWM were all evaluated, and the oil/water separation ability was further investigated. In addition, a commercial available sponge covered with SMWM was fabricated as an oil adsorbent for the purpose of oil recovery. This study demonstrated a convenient and fast method to modify the metal wire mesh to be superhydrophobic and such simple method might find practical applications in the large-scale removal of oils. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Weaver, K.D.; Herring, J.S.; Macdonald, P.E.

    2001-01-01

    Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

  11. Economic Viability of Metallic Sodium-Cooled Fast Reactor Fuel in Korea

    Directory of Open Access Journals (Sweden)

    S. K. Kim

    2013-01-01

    Full Text Available This paper evaluates whether SFR metallic nuclear fuel can be economical. To make this determination, the cost of SFCF (SFR fuel cycle facilities was estimated, and the break-even point of the manufacturing cost of SFR metallic nuclear fuel for direct disposal option was then calculated. As a result of the cost estimation, the levelized unit cost (LUC for SFCF was calculated to be 5,311 $/kgHM, and the break-even point was calculated to be $5,267/kgHM. Therefore, the cost difference between LUC and the break-even point is not only small but is also within the relevant range of the uncertainty level of Class 3 in accordance with a generic cost estimate classification matrix of AACE (the Association for the Advancement of Cost Engineering. This means it is very difficult to judge the economical feasibility of SFR metallic nuclear fuel because as of today there are no commercial facilities in Korea or the world. The economic feasibility of SFR metallic nuclear fuel, however, will be enhanced if the mass production of SFCF becomes possible in the future.

  12. A standardized evaluation of artifacts from metallic compounds during fast MR imaging

    DEFF Research Database (Denmark)

    Murakami, Shumei; Verdonschot, Rinus G; Kataoka, Miyoshi

    2016-01-01

    OBJECTIVE: Metallic compounds present in the oral and maxillofacial regions (OMR) cause large artifacts during MR scanning. We quantitatively assessed these artifacts embedded within a phantom according to standards set by the American Society for Testing and Materials (ASTM). MATERIALS AND METHO...

  13. Recent progress in the development of metallic fuel

    International Nuclear Information System (INIS)

    Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Lahm, C.E.; Pahl, R.G.; Tsai, H.C.

    1990-01-01

    Tests to date demonstrate that metallic fuel for advanced liquid metal reactors performs well, is easily reprocessed and refabricated and provides inherent reactor safety within an economic design. The behavior and performance of metallic fuel is key to the demonstration of the Integral Fast Reactor (IFR) concept at Argonne National Laboratory. Since 1985, more than 40 assemblies of experimental fuel in addition to the standard metallic driver fuel for Experimental Breeder Reactor 2 (EBR-2)have been irradiated; several more continue to be designed and fabricated. Results have characterized the influence of a wide range of fabrication, design and material variables upon irradiation behavior throughout the fuel lifetime under normal and upset conditions including operation with breached cladding. Results of test, both in- and out-of-reactor, indicate that metallic fuel is readily and economically fabricated, capable of achieving high exposure and long reactor residence times, and possesses unique and promising safety features. 9 refs., 6 figs

  14. FOWL CHOLERA IN A BREEDER FLOCK

    Directory of Open Access Journals (Sweden)

    Z. Parveen, A. A. Nasir, K.Tasneem and A. Shah

    2003-12-01

    Full Text Available During January, 2003 Pasteurella multocida the causative agent of fowl cholera was isolated from a breeder flock in Lahore District. The age of the flock was 245 days. Increased mortality, swollen wattles and lameness were the clinical findings present in almost all the affected birds, while gross lesions were typical of fowl cholera. To prove the virulence of the organism, mice and six-week old cockerals were infected and P. multocida was reisolated.

  15. Fast Monitoring Soil Environmental Qualities of Heavy Metal by Portable X-Ray Fluorescence Spectrometer.

    Science.gov (United States)

    Wang, Bao; Yu, Jian-xin; Huang, Biao; Hu, Wen-you; Chang, Qing

    2015-06-01

    Portable X-ray fluorescence (PXRF) spectrometer as a new type of equipment for quick test has a prominent prospect, but there are also shortcomings of detection range and limition, therefore this paper studied the suitability of PXRF spectrometer in monitoring soil environmental qualities of heavy metals included Cr, Ni, Cu, Zn, Pb, Cd, As and Hg, the aim of this paper is to screen elements which can be detected by this kind of instrument and evaluate the accuracy of test results. The research method is to test heavy metals contaminated soil samples by PXRF spectrometer, evaluate the accuracy of test results of PXRF compared with inductively coupled plasma mass(ICP-MS), then establish linear regression relationship between analysis results of PXRF and ICP-MS method. The results show that, (1) When measuring the soil environmental quality, PXRF spectrometer is appropriate to measure the content of Pb, Zn, Cr and Cu, except Ni, Cd, As and Hg. (2) Compared with the test value of ICP-MS, the test value of Pb and Zn is lower, the test value of Cu is higher, the test value of Cr is too high, all the results of PXRF spectrometer should be linear corrected according to standard analysis method. In conclusion, PXRF spectrometer is suitable for monitoring environmental quality of soil which is polluted by heavy metal such as Pb, Zn, Cr and Cu, it is an analysis means with characteristics of simple and rapid, accurate and reliable. The innovation of this article is that reasonable avoiding the shortcomings of PXRF spectrometer as using the instrument to monitor soil environmental quality, at last improved the application value of test results.

  16. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    International Nuclear Information System (INIS)

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option

  17. History and evolution of the breeder reactor

    International Nuclear Information System (INIS)

    Carle, R.

    1989-01-01

    The concept of the breeder reactor is almost as old as the idea of the nuclear reactor itself. From the very first years following the discovery of nuclear fission, scientists and technicians tried to turn mankind's eternal dream into reality; that is, enjoy an abundant source of energy without using up our raw material reserves. Nuclear energy offered several solutions to realize this dream. One of them, fusion, seemed out of our grasp in the near future. But fission of 235 U was possible, and the Manhattan Project soon furnished ample proof of this theory. However, everyone working in this field was conscious of the fact that thermal neutron reactors make very inefficient use of the energy potential contained in natural uranium. The solution was to use in a core sufficiently rich in fissile matter, the excess neutrons to convert the 238 U, so poorly used by other types of reactors, into fissile 239 Pu. Regeneration, or 'breeding' of fuel, can multiply the energy drawn from a ton of uranium by a factor of 50 to 100. This would enable us to ward off the specter of an energy shortage and the rapid depletion of uranium mines. As early as 1945 in Los Alamos, Enrico Fermi stated: 'The country which first develops a breeder reactor will have a great competitive edge in atomic energy.' The development of the breeder reactor in the USA and around the world is discussed

  18. The development of the breeder reactor program of the German Federal Republic

    International Nuclear Information System (INIS)

    Haunschild, H.H.

    1984-01-01

    This article recapitulates the stages of the program which is in hand in the German Federal Republic concerning breeder reactors with fast neutrons. In particular, it describes in detail the causes for the delays in the construction of the demonstration reactor SRN 300 (Kalkar) for which the commissioning has now been fixed for 1987. It emphasizes the importance of the collaboration which exists in Europe for the establishment of the projects and the construction of fast neutron power stations, including the SNR 2 reactor; the construction of the latter could begin in the German Federal Republic at the end of the Eighties [fr

  19. United States Department of Energy breeder reactor staff training domestic program

    International Nuclear Information System (INIS)

    1984-01-01

    Two US DOE projects in the Pacific Northwest offer unique on-the-scene training opportunities at sodium-cooled fast-reactor plants: the Fast Flux Test Facility (FFTF) near Richland, Washington, which has operated successfully in a wide range of irradiation test programs since 1980; and the Experimental Breeder Reactor II (EBR-II) near Idaho Falls, Idaho, which has been in operation for approximately 20 years. Training programs have been especially designed to take advantage of this plant experience. Available courses are described

  20. Fast neutron spectrum in the reflector of swimming pool reactor behind metallics slabs

    International Nuclear Information System (INIS)

    Brousse, J.C.

    1970-01-01

    The large perturbations of fast neutron spectrum were measured behind lead, aluminium and iron slabs in the Siloette reflector at the CENG. The neutron slowing down is chiefly depending of the inelastic reaction. The reaction cross section increases with energy; a spectrum softening is deduced. This is verified. We tried to determine the spectrum shape by calculation to fit the measurements. Calculations were firstly made in unidimensional geometry by the NIOBE transport equation resolution code and by the SANE Monte-Carlo code. The results does not agree with the experimental determined values. Finally a semi-empirical method for studying a tridimensional geometry was chosen. We have obtained calculation results in a perfect agreement with measurements. The method is described. (author) [fr

  1. Application of direct numerical analysis by Fast Fourier Transform to isotopic exchange process in a metal hydride particle bed

    International Nuclear Information System (INIS)

    Fukada, Satoshi; Matsuo, Hiroshi; Mitsuishi, Nobuo

    1993-01-01

    Hydrogen isotope separation based on isotopic exchange on metal hydride surfaces has been proposed. In this paper, the material balance equation of deuterium in a particle bed of a metal hydride and the diffusion equation are solved in a Laplace domain, and Fast Fourier Transform (FFT) is adopted to invert Laplace transforms numerically. The system considered includes steps of fluid-film diffusion, isotopic exchange reaction on surfaces and diffusion in solid particles. The effects of each mass-transfer resistance and axial dispersion on the shape of effluent curves are revealed under two different boundary conditions. Experimental effluent curves were obtained using the pulse-change of an influent deuterium concentration in a LaNi 3 Al 2 hydride bed. The exchange capacity and the overall mass-transfer capacity coefficient were determined from fitting in the time domain and were compared with those by Fourier and moment analyses in terms of the time and accuracy of the calculations. The Fourier analysis gave almost the same values of the mass-transfer quantities in shorter time compared with the analysis by FFT and gave the values of up to the third moment more accurately than those by the moment definition. (author)

  2. What determines hatchling weight: breeder age or incubated egg weight?

    Directory of Open Access Journals (Sweden)

    AB Traldi

    2011-12-01

    Full Text Available Two experiments were carried out to determine which factor influences weight at hatch of broiler chicks: breeder age or incubated egg weight. In Experiment 1, 2340 eggs produced by 29- and 55-week-old Ross® broiler breeders were incubated. The eggs selected for incubation weighed one standard deviation below and above average egg weight. In Experiment 2, 2160 eggs weighing 62 g produced by breeders of both ages were incubated. In both experiments, 50 additional eggs within the weight interval determined for each breeder age were weighed, broken, and their components were separated and weighed. At hatch, hatchlings were sexed and weighed, determining the average initial weight of the progeny of each breeder age. Data were analyzed using the Analyst program of SAS® software package. In Experiment 1, the weight difference between eggs produced by young and mature breeders was 10.92 g, and the component that mostly influenced this difference was the yolk (7.51 g heavier in mature breeders, compared with 4.23 g difference in albumen and 0.8 g in eggshell weights. Hatchling weight difference was 9.4 g higher in eggs from mature breeders. In Experiment 2, egg weight difference was only 0.74 g, but yolk weight was 4.59 g higher in the eggs of mature breeders. The results obtained in the present study indicate that hatchling weight is influenced by egg weight, and not by breeder age.

  3. Correlations between breeder age, egg cholesterol content, blood cholesterol level and hatchability of broiler breeders.

    Science.gov (United States)

    Dikmen, B Yilmaz; Sahan, U

    2007-02-01

    1. The research was carried out to investigate correlations between breeder age, egg cholesterol content, blood cholesterol level and hatchability of broiler breeders. 2. Egg cholesterol content increased with increased breeder age. The mean yolk cholesterol contents (mg per g yolk) were 10.47+/-0.28, 15.34+/-0.65 and 15.64+/-0.71 mg/g at 28, 45 and 65 weeks of age, respectively. 3. There were positive correlations between yolk weight and yolk cholesterol content (mg/g yolk) (r=01.941; Pegg cholesterol content (mg/egg) (r=0.980; Pegg yolk cholesterol content and hatchability of fertile eggs (r=-0.345; Peggs (r=-0.574; Pcholesterol levels were 165.1+/-11.04, 166.5+/-11.97 and 179.5+/-11.33 mg/dl at 28, 45 and 65 weeks of age, respectively.

  4. Modification and characterization of cellulose cotton fibers for fast extraction of some precious metal ions.

    Science.gov (United States)

    Monier, M; Akl, M A; Ali, Wael M

    2014-05-01

    In this work, native cellulose cotton fibers were first modified through graft copolymerization of polyacrylonitrile (PAN) and then by insertion of phenyl thiosemicarbazide moieties to finally produce C-PTS chelating fibers, which were fully characterized using various instrumental techniques such as SEM, FTIR, EDX and XRD spectra. The obtained C-PTS were employed in removal and extraction of Au(3+), Pd(2+) and Ag(+) precious metal ions from their aqueous solutions using batch experiments. The kinetic studies showed that the pseudo-second-order model exhibited the best fit for the experimental data. In addition, the adsorption isotherm studies indicated that the adsorption follows the Langmuir model and the maximum adsorption capacities for Au(3+), Pd(2+) and Ag(+) were 198.31, 87.43 and 71.14 mg/g respectively. Copyright © 2014 Elsevier B.V. All rights reserved.

  5. Core concepts for 'zero-sodium-void-worth core' in metal fuelled fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

    1991-01-01

    Core design options to reduce the sodium void worth in metal fuelled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a 'pancaked' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket-zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. (author)

  6. Core concepts for ''zero-sodium-void-worth core'' in metal fuelled fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

    1991-01-01

    Core design options to reduce the sodium void worth in metal fueled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a ''pancaked'' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. 16 refs., 2 figs., 3 tabs

  7. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  8. Modeling of constituent redistribution in U Pu Zr metallic fuel

    Science.gov (United States)

    Kim, Yeon Soo; Hayes, S. L.; Hofman, G. L.; Yacout, A. M.

    2006-12-01

    A computer model was developed to analyze constituent redistribution in U-Pu-Zr metallic nuclear fuels. Diffusion and thermochemical properties were parametrically determined to fit the postirradiation data from a fuel test performed in the Experimental Breeder Reactor II (EBR-II). The computer model was used to estimate redistribution profiles of fuels proposed for the conceptual designs of small modular fast reactors. The model results showed that the level of redistribution of the fuel constituents of the designs was similar to the measured data from EBR-II.

  9. Interim waste storage for the Integral Fast Reactor

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  10. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-01

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology

  11. Fast Fourier-based simulation of off-resonance artifacts in steady-state gradient echo MRI applied to metal object localization

    NARCIS (Netherlands)

    Zijlstra, Frank; Bouwman, Job G; Braškute, I; Viergever, Max A; Seevinck, PR

    2017-01-01

    PURPOSE: To accelerate simulation of off-resonance artifacts in steady-state gradient echo MRI by using fast Fourier transforms and demonstrate its applicability to metal object localization. THEORY AND METHODS: By exploiting the repetitive nature of steady-state pulse sequences it is possible to

  12. Transport of breeder reactor-fire-generated sodium oxide aerosols for building-wake-dominated meteorology

    Energy Technology Data Exchange (ETDEWEB)

    Fields, D.E.; Cooper, A.C.; Miller, C.W.

    1987-02-01

    This report describes the methodology used and results obtained in efforts to estimate the sodium aerosol concentrations at air intake ports of a liquid-metal cooled, fast-breeder nuclear reactor. An earlier version of this methodology has been previously discussed (Fields and Miller, 1985). A range of wind speeds from 2 to 10 m/s is assumed, and an effort is made to include building wake effects which, in many cases, dominate the dispersal of aerosols near buildings. For relatively small release rates, on the order of 1 to 10 kg/s, the plume rise is small and estimates of aerosol concentrations are derived using the methodology of Wilson and Britter (1982), which describes releases from surface vents. For release rates on the order of 100 kg/s much higher release velocities are expected, and plume rise is considered. An effective increase in release height is computed using the Split-H methodology with a parameterization suggested by Ramsdell (1983), and the release source strength is transformed to rooftop level. Evaluation of the acute release aerosol concentration is then based on the methodology for releases from a surface release of this transformed source strength. For a horizontal release, a methodology is developed to chart the plume path as a function of release and site meteorology parameters. Results described herein must be regarded as maximum aerosol concentrations, based on models derived from generic wind tunnel studies. More accurate and site-specific results may be obtained through wind tunnel simulations and through simulating emissions from release points other than those assumed here.

  13. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1982-01-01

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

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

    Science.gov (United States)

    Mosunova, N. A.

    2018-05-01

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

  15. Design activities of a fusion experimental breeder

    International Nuclear Information System (INIS)

    Huang, J.; Feng, K.; Sheng, G.

    2001-01-01

    The fusion reactor design studies in China are under the support of a fusion-fission hybrid reactor research Program. The purpose of this program is to explore the potential near-term application of fusion energy to support the long-term fusion energy on the one hand and the fission energy development on the other. During 1992-1996 a detailed consistent and integral conceptual design of a Fusion Experimental Breeder, FEB was completed. Beginning from 1996, a further design study towards an Engineering Outline Design of the FEB, FEB-E, has started. The design activities are briefly given. (author)

  16. Design activities of a fusion experimental breeder

    International Nuclear Information System (INIS)

    Huang, J.; Feng, K.; Sheng, G.

    1999-01-01

    The fusion reactor design studies in China are under the support of a fusion-fission hybrid reactor research Program. The purpose of this program is to explore the potential near-term application of fusion energy to support the long-term fusion energy on the one hand and the fission energy development on the other. During 1992-1996 a detailed consistent and integral conceptual design of a Fusion Experimental Breeder, FEB was completed. Beginning from 1996, a further design study towards an Engineering Outline Design of the FEB, FEB-E, has started. The design activities are briefly given. (author)

  17. Burnup characteristics of binary breeder reactors

    International Nuclear Information System (INIS)

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

    1983-01-01

    Burnup calculations of a binary breeder reactor have been done for two cases of fueling. In one case the U 233 /TH fueled inner core and the Pu/U-fueled outer core have the same number of fuel assemblies. In the other case two outermost rings in the inner core are Pu/U-fueled. The second case is considered for an initial phase of thorim cycle introduction when the supply of U 233 could be limited. Results show an efficient breeding on the thorium cycle in both cases. (Author) [pt

  18. The development of breeder reactors in Japan

    International Nuclear Information System (INIS)

    Segawa, M.

    1984-01-01

    In the framework of a global analysis of the various available sources of energy, Japan has reserved a prominent place to the nuclear energy and, in the long-term view, to the breeder reactor which will be due for commercial deployment in 20)10. To achieve these objectives, three stages are envisaged, one of the experimental reactor Joyo (in service), one of the demonstration reactor Monju (its construction has been decided), and one of the pre-commercial reactor (due to be taken in hand at the beginning of the Nineties). Efforts will be made in parallel concerning the fuel cycle [fr

  19. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

    The databases for solid breeder ceramics (Li{sub 2},O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAlO{sub 2}) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized.

  20. Neutronics design for a fusion breeder

    International Nuclear Information System (INIS)

    You Chenglun; Xie Zhongyou

    1987-01-01

    As a fusion breeder, one of the most important figure is support ratio which reflects the economic and fuel production performance of the system to a great extent. In this paper, the support ratio is calculated by using one dimension transport program ANISN and optimized by adjusting 6 Li enrichment and blanket arrangement. The radial distribution of producted U-233 is also taken into account. Measures are taken for better blanket design, and satisfactory results are obtained. Tritium breeding ratio T reaches 1.11 and the support ratio is enhanced from 11 to 14. The engineering, safety and environment performance are improved