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

A worldwide survey of fast breeder reactors

International Nuclear Information System (INIS)

Hennies, H.H.

1986-01-01

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

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

International Nuclear Information System (INIS)

Schikorr, W.M.

1977-01-01

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

Fast breeder fuel cycle, worldwide and French prospects

International Nuclear Information System (INIS)

Rapin, M.

1982-01-01

A review is given of fast breeder fuel cycle development from both the technological and the economical points of view. LMFBR fuel fabrication, reactor operation, spent fuel storage and transportation, reprocessing and fuel cycle economics are topics considered. (U.K.)

Status of fast breeder reactors and associated fuel cycle in India

International Nuclear Information System (INIS)

Chellapandi, P.

2009-01-01

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

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

Status of the breeder fuel cycle in the United States

International Nuclear Information System (INIS)

Burch, W.D.

1985-01-01

This paper reviews the status and plans for the fast reactor fuel cycle in the United States. The United States is undertaking a complete reexamination of its entire breeder program strategy, and the direction of the new program is not yet clear. Studies in progress to examine the associated fuel cycle strategies as they relate to the overall emerging breeder strategy are described. The present status of and recent developments in the fuel cycle R and D programs are summarized

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

The fast breeder reactor

International Nuclear Information System (INIS)

Patterson, W.

1990-01-01

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

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)

Alternative cycles and fast breeders, a look to the future

International Nuclear Information System (INIS)

Dahlberg, R.C.

1979-01-01

The various nuclear strategies that have been proposed to eke out available uranium are briefly summarised. A recent approach is to develop the concept of a 'transitional symbiosis strategy' which accepts the principle of the fast breeder/advanced converter for the long-range future, and to suggest that advanced converters be designed to be near-breeders. Some of the issues affecting strategic planning in the transitional period are reviewed further. The need for the thorium cycle in both advanced converters (or near breeders) and fast reactors is emphasised. This type of high temperature gas-cooled reactor appears to be the technology most suited for symbiosis with FBRs. (UK)

On fusion and fission breeder reactors

International Nuclear Information System (INIS)

Brandt, B.; Schuurman, W.; Klippel, H.Th.

1981-02-01

Fast breeder reactors and fusion reactors are suitable candidates for centralized, long-term energy production, their fuel reserves being practically unlimited. The technology of a durable and economical fusion reactor is still to be developed. Such a development parallel with the fast breeder is valuable by reasons of safety, proliferation, new fuel reserves, and by the very broad potential of the development of the fusion reactor. In order to facilitate a discussion of these aspects, the fusion reactor and the fast breeder reactor were compared in the IIASA-report. Aspects of both reactor systems are compared

Astrid (fast breeder nuclear reactor)

International Nuclear Information System (INIS)

2014-01-01

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

Comparative economics of the breeder and light water reactor

International Nuclear Information System (INIS)

Chow, B.G.

1980-01-01

The issue of breeder timing is studied in this article via a breakeven analysis in which the key driving variables are conveniently segregated into two groups, with uranium price providing the linkage. In one group, the technical and cost characteristics of reactors and fuel cycles determine the uranium breakeven price. In the other group, nuclear demand projections and the uranium supply schedule determine the time paths of uranium price for a given composition of reactor types. The author finds that, even if proliferation risk is ignored, the breeder is not economically competitive with a 30%-improved once-through light water reactor before the year 2030 in the USA and in the world outside communist areas as a whole in 90% of the cases examined. In the exceptional cases, the penalty of delaying commercial breeder introduction to 2030 is small and well within the noise level of long-term energy planning. (author)

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)

Power generation costs for alternate reactor fuel cycles

International Nuclear Information System (INIS)

Smolen, G.R.; Delene, J.G.

1980-09-01

The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

Fast breeder reactor fuel reprocessing in France

International Nuclear Information System (INIS)

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

1984-08-01

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

Parametric studies on the fuel salt composition in thermal molten salt breeder reactors

International Nuclear Information System (INIS)

Nagy, K.; Kloosterman, J.L.; Lathouwers, D.; Van der Hagen, T.H.J.J.

2008-01-01

In this paper the salt composition and the fuel cycle of a graphite moderated molten salt self-breeder reactor operating on the thorium cycle is investigated. A breeder molten salt reactor is always coupled to a fuel processing plant which removes the fission products and actinides from the core. The efficiency of the removal process(es) has a large influence on the breeding capacity of the reactor. The aim is to investigate the effect on the breeding ratio of several parameters such as the composition of the molten salt, moderation ratio, power density and chemical processing. Several fuel processing strategies are studied. (authors)

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

The fast breeder reactor

International Nuclear Information System (INIS)

Keck, O.

1984-01-01

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

Fissile fuel dynamics of breeder/converter reactors

International Nuclear Information System (INIS)

Harms, A.A.

1978-01-01

The long-term fissile fuel dynamics for a hierarchy of fission reactors covering the range from pure-burners to super-breeders is examined. It is found that the breeding gains of the core and blanket can be used to identify several distinct fissile fuel histories and elucidate the importance of fuel cycle characteristics such as the time dependence of the fissile fuel doubling time. On this basis, a self-sufficient fission reactor is introduced and its determining characteristics are identified. (author)

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

Alternative breeder and near-breeder systems

International Nuclear Information System (INIS)

Critoph, E.

1983-01-01

Nuclear power reactor systems have been developed over the last three decades to the point where they are economically competitive, safe and reliable sources of electrical energy. However, with our present knowledge of fissile resources, there is no assurance that the commercially proven reactor systems, using their current fuel cycles, could play a major role in supplying the total world energy needs of the next, and subsequent, centuries. There is a wide consensus that such assurance requires development of reactor systems with very significantly improved fuel resource utilization. The best known of these, and the one currently receiving the lion's share of attention and effort, is the fast breeder reactor (FBR). This paper reviews the characteristics, development status and planned programmes for alternative concepts to the FBR that meet the requirement for large improvement in fuel resource utilization, i.e. alternative breeder and near-breeder systems. These include: heavy-water reactors operating on thorium fuel cycles, light-water high-conversion and breeder reactors, high-temperature gas-cooled reactors operating on thorium fuel cycles, molten salt reactors and heavy-water suspension reactors. Any attempt to make a logical choice for exploitation among these various alternatives involves a consideration of the interplay between reactor system characteristics on the one hand and a forecast of political and economic environments on the other. The reactor breeding (or conversion) ratio has received a great deal of emphasis, but an optimum choice depends also on a consideration of several other factors, including out- and in-reactor specific fuel inventories, fuel fabrication and reprocessing costs, reactor capital cost and load factor, fuel resources and demand growth rate of capacity. Possible variations in this optimum choice with time and regional location are discussed

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.)

Fission-suppressed hybrid reactor: the fusion breeder

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Coops, M.S.

1982-12-01

Results of a conceptual design study of a 233 U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed

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

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

International Nuclear Information System (INIS)

2012-12-01

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

Interim Report on Fluid-Fuel Thermal Breeder Reactors (Revised)

International Nuclear Information System (INIS)

MacPherson, H. G.; Alexander, L. G.; Carter, W. L.; Chapman, R. H.; Kinyon, B. W.; Miller, J. W.

1960-01-01

The merits of aqueous-homogeneous ), graphite-moderated molten salt (MSBR) , and graphite-moderated liquid-bismuth (LBBR) breeder reactors operated at nearly comparable fuel-cycle costs (~1.5 mills/kwhr) were evaluated. The net electrical plant capability was assumed to be 1000 MwE, and the fuel and fertile streams were processed continuously on-site. The specific powers based on fuel were 1.2, 1.2, and 0.5 MwE/kg respectively, and 5.9, 3.7, and 5.3 MwE/tonne based on thorium. Net breeding ratios were 1.10, 1.07, and 1.07, giving doubling times of 5-1/2, 11, and 25 full power years . The fuel-cycle costs at the design points selected were 1.4, 1.3, and 1.6 mills/kwhr . The AHBR has an advantage in breeding ratio and doubling time because D 2 O is superior to graphite as a moderator in breeder reactors. MSBR has an advantage in fuel-cycle costs and in inventory of uranium in the fertile stream as a result of using a solution blanket.

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,

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

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

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

An overview of the potential of the CANDU reactor as a thermal breeder

International Nuclear Information System (INIS)

Slater, J.B.

1977-02-01

This paper is concerned with the use of thorium as a fuel in the existing CANDU concept. The neutron balance of the reactor core is analyzed and an assessment is made of the potential for development of a thermal 'breeder' reactor system. It is concluded that while the SSET cycle (i.e. self-sufficient equilibrium thorium cycle) appears feasible, there is little potential for developing a significant 'breeding' fuel cycle if current reactor operating capability and capital costs are to be maintained. (author)

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

International breeder reactor development

International Nuclear Information System (INIS)

Traube, K.

1976-01-01

For more than a decade, sodium cooled breeder reactors have now been in the focus of advanced nuclear power development in the major industrialized countries. In the sixties, a total of seven small experimental nuclear power stations were commissioned. Two of these have been shut down in the meantime, the others continue to work satisfactorily, their main purpose being the development of fuel elements. The years 1972-1974 saw the commissioning of the prototype power stations in the 300 MWe power category in France, the United Kingdom and the Soviet Union. Presently, other experimental reactors are under construction in the Federal Republic of Germany, Italy, Japan, the United States, plus another Soviet 600 MWe prototype reactor and the SNR 300 DeBeNeLux prototype at Kalkar. A comparison of the technological features either implemented or planned in the prototype and experimental power plants and of their fuel elements reveals a remarkable similarity in the basic concepts pursued in different countries. The two types of breeder reactors, viz. the loop and the pool types, show a closer resemblance to each other than do pressurized and boilling water reactors. The growing awareness of administrative problems emerging in the approaching phase of the introduction of large breeder power stations in a number of European countries has recently led to a streamlining effort in the structure of industries and to tentative steps towards international cooperation on a broad basis. (orig.) [de

Calculation of the evolution of molten salt breeder reactor

International Nuclear Information System (INIS)

Esteves, Fernando de Avelar

1999-01-01

A forecast for the future electrical consumption in Brazil and forecast of the nuclear electrical generation demand are discussed in this paper, which includes also an analysis on advanced nuclear reactors concept to supply that demand. This paper presents a concise description of the Molten Salt Breeder Reactor, considered the most appropriated to meet that demand. This paper also presents the burnup calculation modeling, including the operation modeling of this type of reactor from an initial load o 233 U up to the equilibrium cycle, the results of these calculations and its analysis. (author)

Programme and current status of fast breeder reactor development in Japan

International Nuclear Information System (INIS)

Suita, T.; Oyama, A.

1977-01-01

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

Molten salt breeder reactor

International Nuclear Information System (INIS)

1977-01-01

MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

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

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

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

The need for high performance breeder reactors

International Nuclear Information System (INIS)

Vaughan, R.D.; Chermanne, J.

1977-01-01

It can be easily demonstrated, on the basis of realistic estimates of continued high oil costs, that an increasing portion of the growth in energy demand must be supplied by nuclear power and that this one might account for 20% of all the energy production by the end of the century. Such assumptions lead very quickly to the conclusion that the discovery, extraction and processing of the uranium will not be able to follow the demand; the bottleneck will essentially be related to the rate at which the ore can be discovered and extracted, and not to the existing quantities nor their grade. Figures as high as 150.000 T/annum and more would be quickly reached, and it is necessary to wonder already now if enough capital can be attracted to meet these requirements. There is only one solution to this problem: improve the conversion ratio of the nuclear system and quickly reach the breeding; this would lead to the reduction of the natural uranium consumption by a factor of about 50. However, this condition is not sufficient; the commercial breeder must have a breeding gain as high as possible because the Pu out-of-pile time and the Pu losses in the cycle could lead to an unacceptable doubling time for the system, if the breeding gain is too low. That is the reason why it is vital to develop high performance breeder reactors. The present paper indicates how the Gas-cooled Breeder Reactor [GBR] can meet the problems mentioned above, on the basis of recent and realistic studies. It briefly describes the present status of GBR development, from the predecessors in the gas cooled reactor line, particularly the AGR. It shows how the GBR fuel takes mostly profit from the LMFBR fuel irradiation experience. It compares the GBR performance on a consistent basis with that of the LMFBR. The GBR capital and fuel cycle costs are compared with those of thermal and fast reactors respectively. The conclusion is, based on a cost-benefit study, that the GBR must be quickly developed in order

Breeder reactor fuel fabrication system development

International Nuclear Information System (INIS)

Bennett, D.W.; Fritz, R.L.; McLemore, D.R.; Yatabe, J.M.

1981-01-01

Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

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

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

The GBR reactor an economically competitive Breeder

International Nuclear Information System (INIS)

Chermanne, J.

1974-01-01

In this article the design is described of a 1200 MWe fast breeder, gas-cooled reactor (GBR-4), prepared by a group of experts of the Gas Breeder Reactor Association and used as a reference system for economical and safety evaluations, as well as for defining the research and development program focussed on such concept and the specifications of the prospective demonstrative plant

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Assessment of the thorium fuel cycle in power reactors

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.; Allen, E.J.

1977-01-01

A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

Gas core reactors for actinide transmutation and breeder applications. Annual report

International Nuclear Information System (INIS)

Clement, J.D.; Rust, J.H.

1978-01-01

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions

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

Quantities of actinides in nuclear reactor fuel cycles

International Nuclear Information System (INIS)

Ang, K.P.

1975-01-01

The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000 MW reactors of the following types: water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breeder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium, and recycled uranium. The radioactivity levels of plutonium, americium, and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the United States nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium processed in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing and fuel fabrication to eliminate the off-site transport of separated plutonium. (U.S.)

The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

International Nuclear Information System (INIS)

Dautray, R.

2011-01-01

The author gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the fifties. Neutron transport theory, thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, heat exchanges...) have now attained maturity, sufficient to implement sodium cooling circuits. However, the use of metallic sodium still raises certain severe questions in terms of safe handling and security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchangers) are undergoing in-depth research so as to last longer. The fuel cycle, notably the re-fabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts. (author)

Dynamics and control of molten-salt breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Sing, Vikram; Lish, Matthew R.; Chvala, Ondrej; Upadhyaya, Belle R. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

2017-08-15

Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR) system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

Dynamics and control of molten-salt breeder reactor

Directory of Open Access Journals (Sweden)

Vikram Singh

2017-08-01

Full Text Available Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

Gas-cooled breeder reactor safety

Energy Technology Data Exchange (ETDEWEB)

Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

1981-01-15

The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

The nuclear question at the start of the '80s: the breeder reactor

International Nuclear Information System (INIS)

Owen, R.; Svensson, B.

1980-01-01

The four newspaper articles and the letter cover the following matters: general introduction about breeder reactors and the situation in Swedish politics; visit to Dounreay to discuss breeder reactors (breeding, safety, plutonium production, radiation protection); PuO 2 -UO 2 mixed fuel; description of breeder reactors; efficiency in use of U-235; DFR and PFR; breeder reactors in Swedish politics (arguments for and against nuclear power in general, breeder reactors in particular); discussion of the future of nuclear power in Sweden. (U.K.)

Coatings for fast breeder reactor components

International Nuclear Information System (INIS)

Johnson, R.N.

1984-04-01

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

Breeder control fusion reactor. Topical interview

Energy Technology Data Exchange (ETDEWEB)

Schlueter, A [Max-Planck-Institut fuer Plasmaphysik, Garching/Muenchen (Germany, F.R.)

1977-09-01

The energy sources of the future are extremely controversial. The consumption of fossil fuel shall decrease during the next decades, because exhaustion of the resources, pollution, increase of CO/sub 2/ in the atmosphere and other reasons. But at present the question it not yet settled which alternative energy system should replace the fossil fuel. First of all nuclear energy in the form of fission reactions seems to come into operation to a larger extent. The next step may be the controlled thermonuclear fusion reaction. Furthermore, a comparison between fusion and fission is given which shows that fusion would bring about less risks than the breeders. An advantage of the fusion reactor would be the fact that the fuel cycle is closed. Unfortunately, the physical questions are not as yet satisfactorily clarified so that one cannot be sure whether a fusion reactor can really be built.

Liquid metal cooled fast breeder nuclear reactors

International Nuclear Information System (INIS)

Gatley, J.A.

1979-01-01

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

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)

Alternative nuclear fuel cycles and gas-cooled breeders

International Nuclear Information System (INIS)

Pellaud, B.; Dahlberg, R.C.

1979-01-01

The authors assert that it is prudent to concurrently develop more than one breeder system and to develop advanced converter reactors along with breeders. They provide arguments to support these assertions. (Auth.)

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)

Optimization of binary breeder reactor VI - An acceptable project of binary breeder reactor

International Nuclear Information System (INIS)

Ishiguro, Y.; Dias, A.F.

1986-05-01

A binary breeder reactor that achieves desired characteristics reasonably well has been developed. Its design and characteristics are reported. Previous models showed several complications that result from introduction of 233 U/Th fuel in the core of a LMFBR, compared to purely Pu/U fueled ones. In this new model, the core is made larger to achieve higher breeding ratios and longer refueling intervals, the number of fuel assemblies is increased to accomodate a larger number of control rod assemblies required to compensate for reactivity losses and to control oscillations of the power densities, and, consequently, the fuel inventories are higher. High fuel burnups are achieved without too much complications in the refueling schedule and the power densities can be maintained reasonably constant over an operational cycle. Low sodium void reactivity reduce the potential for severe accidents and a reasonably efficient utilization of thorium can be realized. (Author) [pt

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.)

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)

Consolidated fuel reprocessing programme: Analysis of various options for the breeder fuel cycle in the USA

International Nuclear Information System (INIS)

Stradley, J.G.; Burch, W.D.; Yook, H.R.

1986-01-01

The United States Department of Energy (DOE) has established a programme to develop innovative liquid metal reactor (LMR) designs to assist in developing future U.S. reactor strategy. The paper describes studies in progress to examine various fuel cycle strategies that relate to the reactor strategy. Three potential fuel cycle options that focus on supporting an initial 1300 MW(e) reactor station have been defined: (1) Completion and utilization of the Breeder Reprocessing Engineering Test/Secure Automated Fabrication (BRET/SAF) in the Fuels and Materials Examination Facility (FMEF) at Hanford, Washington; (2) a co-located fuel cycle facility; and (3) delayed closure of the fuel cycle for five to ten years. The BRET, designed as a development facility, has sufficient capacity to service the needs of an initial module at an LMR station. It appears feasible to increase this capacity and to utilize SAF in the FMEF to accommodate the projected output (up to 35 MtHM/year) from the 1300 MW(e) liquid-metal concepts under study. Plans developed within the United States Consolidated Management Office for an initial reactor project have envisioned that cost savings could be realized by delaying the closure of the fuel cycle as long as supplies of plutonium could be obtained relatively inexpensively. This might prove to be only five to ten years, but even that period might be long enough for the fuel cycle costs to be spread over more than one reactor rather than loaded on the initial project. This concept is being explored as is the question of the future coupling of a light water reactor reprocessing industry for plutonium supply to breeder recycle

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

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

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.

UF6 breeder reactor power plants for electric power generation

International Nuclear Information System (INIS)

Rust, J.H.; Clement, J.D.; Hohl, F.

1976-01-01

The reactor concept analyzed is a 233 UF 6 core surrounded by a molten salt (Li 7 F, BeF 2 , ThF 4 ) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

Conceptual design of Indian molten salt breeder reactor

International Nuclear Information System (INIS)

Vijayan, P.K.; Basak, A.; Dulera, I.V.; Vaze, K.K.; Basu, S.; Sinha, R.K.

2014-01-01

The fuel in a molten salt breeder reactor is in the form of a continuously circulating molten salt. Fluoride based salts have been almost universally proposed. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. This constitutes a major technological challenge for this type of reactors. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian Molten Salt Breeder Reactor (IMSBR). Presently various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel fundamental studies as regards various molten salts have also been initiated. This paper would discuss conceptual design of these reactors, as well as associated issues and technologies

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

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

International Nuclear Information System (INIS)

Swaminathan, P.

1999-01-01

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

Fuel Summary Report: Shippingport Light Water Breeder Reactor

International Nuclear Information System (INIS)

Illum, D.B.; Olson, G.L.; McCardell, R.K.

1999-01-01

The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749

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

Gas cooled fast breeder reactors using mixed carbide fuel

International Nuclear Information System (INIS)

Kypreos, S.

1976-09-01

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

Growth rates of breeder reactor fuel. Final report

International Nuclear Information System (INIS)

Ott, K.O.

1979-01-01

During the contract period, a consistent formalism for the definition of the growth rates (and thus the doubling time) of breeder reactor fuel has been developed. This formalism was then extended to symbiotic operation of breeder and converter reactors. Further, an estimation prescription for the growth rate has been developed which is based upon the breeding worth factors. The characteristics of this definition have been investigated, which led to an additional integral concept, the breeding bonus

The fast breeder reactor

International Nuclear Information System (INIS)

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

1990-01-01

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

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

Refueling system for the gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Hawke, B.C.

1980-05-01

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

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Feasibility study on commercialization of fast breeder reactor cycle systems interim report of phase II. Technical study report for nuclear fuel cycle systems

International Nuclear Information System (INIS)

Sato, Koji; Amamoto, Ippei; Inoue, Akira

2004-06-01

As a part of the feasibility study on commercialization of fast breeder reactor cycle systems, the plant concept concerning the fuel cycle systems (combination of the reprocessing and the fuel fabrication) has been constructed to reduce their total cost by the introduction of various innovative techniques and to apply their utmost superior efficiency from such standpoints of a decrease in the environmental burden, better resource utilization and proliferation resistance improvement by the low decontamination transuranium element (TRU) recycle. This interim report of Phase II describes the results of an on-going study which will cover a five-year period. For oxide fuels, the system which combines the use of the advanced aqueous reprocessing using three main methods such as the crystallization method, the simplified solvent extraction method, and the extraction chromatography method for minor actinide (MA) recovery, as well as the simplified pelletizing fuel fabrication which rationalized a powder mixing process etc., has abundant current results and a high technical feasibility for the basic process. Though this system faces difficulties in the technical development of control technology of the extraction chromatography and the fabrication technology of low decontamination TRU fuel etc., its expected practical use is possible at an early stage. As for the super-critical direct extraction reprocessing, it is necessary to fulfill more basic data although further economical improvement of an advanced aqueous reprocessing is expected. The system which combines the advanced aqueous reprocessing and the gelation sphere packing fuel fabrication has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the fuel fabrication process. However, this system will shoulder additional cost for the reagent recovery process and the waste liquid treatment process due to need to dispose of a large bulk of process waste liquid. The system which

Breeders. Les surregenerateurs

Energy Technology Data Exchange (ETDEWEB)

1978-01-01

The interest lying in the nuclear energy as a source of electric power is recalled with a view to the consumption of primary energy and electric power; the position of France is particularly discussed. The fast breeder reactor is presented and the state of development of this type of reactor in France is discussed with the planning of the Creys-Malville power plant. The problems successively examined are concerned with: fuel cycle, radioactivity and effluents, breeder safety, licensing procedures, sodium coolant, plutonium, fuel reprocessing, environmental impact and waste management.

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

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

Experimental Breeder Reactor I Preservation Plan

Energy Technology Data Exchange (ETDEWEB)

Julie Braun

2006-10-01

Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

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

Uranium alloys for using in fast breeder reactors

International Nuclear Information System (INIS)

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

1988-08-01

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

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

Assessment of residual life of fast breeder test reactor

International Nuclear Information System (INIS)

Srinivasan, G.

2016-01-01

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

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.)

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.

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

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.

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.

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

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

Preliminary design of a Binary Breeder Reactor

International Nuclear Information System (INIS)

Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C.

2014-10-01

A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

Clinch River breeder project gets boost

International Nuclear Information System (INIS)

Hill, W.H.

1982-01-01

Progress on the Clinch River Breeder Reactor Plant project, the United States' next step in developing liquid metal fast breeder technology is examined including consideration of Plant design, component fabrication and testing, construction schedule, funding, fuel cycle development and licensing. (U.K.)

Materials requirements for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

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

1978-01-01

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

Optimization of binary breeder reactor. 1. Sodium void reactivity and Doppler effect in a new model

International Nuclear Information System (INIS)

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

1985-01-01

A model for the Binary Breeder Reactor (BBR) is examined for the inherent safety characteristics, sodium void reactivity and Doppler effect in the beginning of cycle and a hypothetical end of cycle. In addition to the standard fueling mode of the BBR, two others are considered: U 238 /U 233 -alternate fueling, and U 238 /PU-normal fueling of LMFBRs. (Author) [pt

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

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2000-01-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

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

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

International Nuclear Information System (INIS)

Kapoor, R.P.

2001-01-01

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

Status of fast breeder reactor development in the United States

Energy Technology Data Exchange (ETDEWEB)

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

1981-05-01

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

Status of fast breeder reactor development in the United States

International Nuclear Information System (INIS)

Horton, K.

1981-01-01

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

The thorium fuel cycle in water-moderated reactor systems

International Nuclear Information System (INIS)

Critoph, E.

1977-05-01

Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An 'inventory' of uranium of between 1 and 2 Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium) is some two decades

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Pool type liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Guthrie, B.M.

1978-08-01

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

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)

Fuel cycle cost analysis on molten-salt reactors

International Nuclear Information System (INIS)

Shimazu, Yoichiro

1976-01-01

An evaluation is made of the fuel cycle costs for molten-salt reactors (MSR's), developed at Oak Ridge National Laboratory. Eight combinations of conditions affecting fuel cycle costs are compared, covering 233 U-Th, 235 U-Th and 239 Pu-Th fuels, with and without on-site continuous fuel reprocessing. The resulting fuel cycle costs range from 0.61 to 1.18 mill/kWh. A discussion is also given on the practicability of these fuel cycles. The calculations indicate that somewhat lower fuel cycle costs can be expected from reactor operation in converter mode on 235 U make-up with fuel reprocessed in batches every 10 years to avoid fission product precipitation, than from operation as 233 U-Th breeder with continuous reprocessing. (auth.)

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)

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

Mechanical design of a light water breeder reactor

International Nuclear Information System (INIS)

Fauth, W.L. Jr.; Jones, D.S.; Kolsun, G.J.; Erbes, J.G.; Brennan, J.J.; Weissburg, J.A.; Sharbaugh, J.E.

1976-01-01

In a light water reactor system using the thorium-232--uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements. 4 claims, 24 drawing figures

Reasons for opposition to the breeder reactor

International Nuclear Information System (INIS)

Timbal-Duclaux, Louis

1979-01-01

The author gives a sociological analysis of the opposition to breeder reactors in France, stressing that the antinuclear groups main thrust of protest against the Super-Phenix has dimished since its apex two years ago [fr

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)

Design features of the Light Water Breeder Reactor (LWBR) which improve fuel utilization in light water reactors (LWBR development program)

International Nuclear Information System (INIS)

Hecker, H.C.; Freeman, L.B.

1981-08-01

This report surveys reactor core design features of the Light Water Breeder Reactor which make possible improved fuel utilization in light water reactor systems and breeding with the uranium-thorium fuel cycle. The impact of developing the uranium-thorium fuel cycle on utilization of nuclear fuel resources is discussed. The specific core design features related to improved fuel utilization and breeding which have been implemented in the Shippingport LWBR core are presented. These design features include a seed-blanket module with movable fuel for reactivity control, radial and axial reflcetor regions, low hafnium Zircaloy for fuel element cladding and structurals, and a closely spaced fuel rod lattice. Also included is a discussion of several design modifications which could further improve fuel utilization in future light water reactor systems. These include further development of movable fuel control, use of Zircaloy fuel rod support grids, and fuel element design modifications

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)

Radioactive waste management at a Liquid Metal Fast Breeder Reactor

International Nuclear Information System (INIS)

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

1979-01-01

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

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

International Nuclear Information System (INIS)

Meacham, S.A.

1982-01-01

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

Future fuel cycle development for CANDU reactors

International Nuclear Information System (INIS)

Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

1987-01-01

The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

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.)

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

International Nuclear Information System (INIS)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

Baumier, J.; Duchatelle, L.

1984-01-01

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

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

International Nuclear Information System (INIS)

Schroeder, R.; Wagner, J.

1975-06-01

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

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

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

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

Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

International Nuclear Information System (INIS)

Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

2015-01-01

Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

Past and present role of fast breeder reactors in Italy

International Nuclear Information System (INIS)

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

1978-01-01

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

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.)

Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

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

1993-01-01

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

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

International Nuclear Information System (INIS)

Massimino, R.J.; Williams, D.A.

1983-05-01

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

Energy Technology Data Exchange (ETDEWEB)

Massimino, R.J.; Williams, D.A.

1983-05-01

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

High-temperature and breeder reactors - economic nuclear reactors of the future

International Nuclear Information System (INIS)

Djalilzadeh, A.M.

1977-01-01

The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

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)

The development of the gas cooled fast breeder reactor

International Nuclear Information System (INIS)

Dalle Donne, M.; Kernforschungszentrum Karlsruhe

1975-01-01

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

Pebble Bed Reactor: core physics and fuel cycle analysis

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Worley, B.A.

1979-10-01

The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes.

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

International Nuclear Information System (INIS)

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

1979-07-01

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

Conceptual design of reactor assembly of prototype fast breeder reactor

International Nuclear Information System (INIS)

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

1996-01-01

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

Advance reactor and fuel-cycle systems--potentials and limitations for United States utilities

International Nuclear Information System (INIS)

Zebroski, E.L.; Williams, R.F.

1979-01-01

This paper reviews the potential benefits and limitations of advance reactor and fuel-cycle systems for United States utilities. The results of the review of advanced technologies show that for the near and midterm, the only advance reactor and fuel-cycle system with significant potential for United States utilities is the current LWR, and evolutionary, not revolutionary, enhancements. For the long term, the liquid-metal breeder reactor continues to be the most promising advance nuclear option. The major factors leading to this conclusion are summarized

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.)

Practical introduction of thorium fuel cycles

International Nuclear Information System (INIS)

Kasten, P.R.

1982-01-01

The pracitcal introduction of throrium fuel cycles implies that thorium fuel cycles compete economically with uranium fuel cycles in economic nuclear power plants. In this study the reactor types under consideration are light water reactors (LWRs), heavy water reactors (HWRs), high-temperature gas-cooled reactors (HTGRs), and fast breeder reactors (FBRs). On the basis that once-through fuel cycles will be used almost exclusively for the next 20 or 25 years, introduction of economic thorium fuel cycles appears best accomplished by commercial introduction of HTGRs. As the price of natural uranium increases, along with commercialization of fuel recycle, there will be increasing incentive to utilize thorium fuel cycles in heavy water reactors and light water reactors as well as in HTGRs. After FBRs and fuel recycle are commercialized, use of thorium fuel cycles in the blanket of FBRs appears advantageous when fast breeder reactors and thermal reactors operate in a symbiosis mode (i.e., where 233 U bred in the blanket of a fast breeder reactor is utilized as fissile fuel in thermal converter reactors)

Real time simulation method for fast breeder reactors dynamics

International Nuclear Information System (INIS)

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

1985-01-01

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

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

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

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

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

Breeder reactor program in the USA

International Nuclear Information System (INIS)

Brewer, S.

1978-01-01

In the United States, commercial fuel reprocessing and demonstration test of plutonium breeder reactors were now postponed. LMFBR project and schedule of FFTF and afterwards await the results of INFCE. However, this is not discarding the development of LMFBRs. With the existing energy resources, the United States can have the opportunity to make careful and thorough study. Emphasis is placed on the research and development on new alternative types of fuel. FFTF going to be operated soon should provide effective means for the developments of FBR fuel and materials. High priority is to be retained for the test and development of sodium system hardwares. The nuclear proliferation problem is not related to heat transfer and secondary systems; it is associated with the selection of fuel and fuel cycle. The whole program is centered around LMFBR design and development. The target output will be 600 x 10 3 -- 700 x 10 3 MW. In the United States, now is the time to develop excellent products and to study the nuclear proliferation problem more carefully. (Mori, K.)

Swiss breeder research programme

International Nuclear Information System (INIS)

1992-01-01

to find a sound base for the financial support. For the reactor physics research two almost parallel activities were considered. During the first period mainly existing know-how will be applied and a step by step familiarisation with the significance of fast breeder reactor physics is foreseen. New pointwise and group-wise cross section libraries based on ENDFIB-VI and JEF 1.1 have been prepared. A large (1250 MWe) sodium-cooled fast breeder reactor benchmark problem was calculated and the eigenvalues, isothermal core fuel Doppler-reactivities, effective delayed neutron fraction and reactivity worths were compared with a great number of solutions obtained in the past. During the following period new methods and models to calculate burnup-cycles of large breeder reactors should be developed and tested. Data libraries for shielding problems to be used in the ECCO code will be prepared and shielding problems calculated. The thermal hydraulics research is conducted to investigate the flow structures produced by two parallel layers of liquid at different velocities and temperatures. This problem arises particularly on occasions where natural circulation is prevailing and hot and cold streams of liquid come together. At present, tests are carried out with water in an horizontal glass channel (WAMIX). Two flow visualization techniques are being used: laser-sheet induced luminescence and image-analysis of video pictures taken with ink injection marking. Based on the image analysis a determination of the frequency of appearance of vortices (time-dependence) could be made. In the analytical area the computational thermal hydraulics code ASTEC was further validated by participation in an international benchmark calculation exercise. This code is also used to calculate the velocity profiles in the boundary layer of the inlet segment of the WAMIX test section. It is intended to directly participate in the European and the French R and D programmes for sodium-cooled fast breeder

Combined system of accelerator molten-salt breeder (AMSB) apd molten-salt converter reactor (MSCR)

International Nuclear Information System (INIS)

Furukawa, K.; Kato, Y.; Ohmichi, T.; Ohno, H.

1983-01-01

A design and research program is discUssed of the development of accelerator molten-salt breeder (AMSB) consisting of a proton accelerator and a molten fluoride target. The target simultaneously serves as a blanket for fissionable material prodUction. An addition of some amoUnt of fissile nuclides to a melt expands the AMSB potentialities as the fissionable material production increases and the energy generation also grows up to the level of self-provision. Besides the blanket salts may be used as nuclear fuel for molten-salt converter reactor (MSCR). The combined AM SB+MSCR system has better parameters as compared to other breeder reactors, molten-salt breeder reactors (MSBR) included

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

International Nuclear Information System (INIS)

Horton, K.E.

1983-01-01

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

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

Nuclear-fuel-cycle costs. Consolidated Fuel-Reprocessing Program

International Nuclear Information System (INIS)

Burch, W.D.; Haire, M.J.; Rainey, R.H.

1981-01-01

The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel-cycle costs are given for the pressurized-water reactor once-through and fuel-recycle systems, and for the liquid-metal fast-breeder-reactor system. These calculations show that fuel-cycle costs are a small part of the total power costs. For breeder reactors, fuel-cycle costs are about half that of the present once-through system. The total power cost of the breeder-reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment

Operability design review of prototype large breeder reactor (PLBR) designs. Final report, September 1981

International Nuclear Information System (INIS)

Beakes, J.H.; Ehman, J.R.; Jones, H.M.; Kinne, B.V.T.; Price, C.M.; Shores, S.P.; Welch, J.K.

1981-09-01

Prototype Large Breeder Reactor (PLBR) designs were reviewed by personnel with extensive power plant operations experience. Fourteen normal and off-normal events, such as startup, shutdown, refueling, reactor scram and loss of feedwater, were evaluated using an operational evaluation methodology which is designed to facilitate talk-through sessions on operational events. Human factors engineers participated in the review and assisted in developing and refining the review methodologies. Operating experience at breeder reactor facilities such as Experimental Breeder Reactor-II (EBR-II), Enrico Fermi Atomic Power Plant - Unit 1, and the Fast Flux Test Facility (FFTF) was gathered, analyzed, and used to determine whether lessons learned from operational experience had been incorporated into the PLBR designs. This eighteen month effort resulted in approximately one hundred specific recommendations for improving the operability of PLBR designs

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

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

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

2013-09-01

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

Fast breeder reactor at Kalkar. Pt. 2

International Nuclear Information System (INIS)

Degen, G.

1979-02-01

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

Feasibility study on the thorium fueled boiling water breeder reactor

International Nuclear Information System (INIS)

PetrusTakaki, N.

2012-01-01

The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

Assessment of the accident response of a light-water-moderated breeder-reactor system: AWBA development program

International Nuclear Information System (INIS)

High, H.M.

1983-05-01

The predicted accident response for a light water moderated, thorium/U-233 fueled, seed-blanket reactor concept was assessed. The first part of the assessment compared breeder accident response with that of a current commercial pressurized water reactor design for several different types of transients. Based on these comparisons and a review of the various parameter differences between the breeder and a U-235 fueled plant, the second part of the assessment studied the breeder accident behavior in more detail, particularly in areas of potential concern. Based on the two parts of the assessment, it was concluded that the breeder accident response would be very similar to that of present commercial pressurized water reactor plants. The large Doppler and moderator reactivity coefficients of the breeder would significantly reduce the severity of many of the accidents that must be considered. It is expected that the accident response of the breeder can be shown to meet regulatory criteria

Thorium fuel cycle analysis

Energy Technology Data Exchange (ETDEWEB)

Yamaji, K [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

1980-07-01

Systems analysis of the thorium cycle, a nuclear fuel cycle accomplished by using thorium, is reported in this paper. Following a brief review on the history of the thorium cycle development, analysis is made on the three functions of the thorium cycle; (1) auxiliary system of U-Pu cycle to save uranium consumption, (2) thermal breeder system to exert full capacity of the thorium resource, (3) symbiotic system to utilize special features of /sup 233/U and neutron sources. The effects of the thorium loading in LWR (Light Water Reactor), HWR (Heavy Water Reactor) and HTGR (High Temperature Gas-cooled Reactor) are considered for the function of auxiliary system of U-Pu cycle. Analysis is made to find how much uranium is saved by /sup 233/U recycling and how the decrease in Pu production influences the introduction of FBR (Fast Breeder Reactor). Study on thermal breeder system is carried out in the case of MSBR (Molten Salt Breeder Reactor). Under a certain amount of fissile material supply, the potential system expansion rate of MSBR, which is determined by fissile material balance, is superior to that of FBR because of the smaller specific fissile inventory of MSBR. For symbiotic system, three cases are treated; i) nuclear heat supply system using HTGR, ii) denatured fuel supply system for nonproliferation purpose, and iii) hybrid system utilizing neutron sources other than fission reactor.

Steam explosions in sodium cooled breeder reactors

International Nuclear Information System (INIS)

Lundell, B.

1982-01-01

Steam explosion is considered a physical process which transport heat from molten fuel to liquid coolant so fast that the coolant starts boiling in an explosion-like manner. The arising pressure waves transform part of the thermal energy to mechanical energy. This can stress the reactor tank and threaten its hightness. The course of the explosion has not been theoretical explained. Experimental results indicate that the probability of steam explosions in a breeder reactor is small. The efficiency of the transformation of the heat of fusion into mechanical energy in substantially lower than the theoretical maximum value. The mechanical stress from the steam explosion on the reactor tank does not seem to jeopardize its tightness. (G.B.)

Some basic concepts of fast breeder reactor safeguards

International Nuclear Information System (INIS)

Tkharev, E.; Walford, F.J.

1987-04-01

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

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

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

Tritium breeders and tritium permeation barrier coatings for fusion reactor

International Nuclear Information System (INIS)

Yamawaki, Michio; Kawamura, Hiroshi; Tsuchiya, Kunihiko

2004-01-01

A state of R and D of tritium breeders and tritium permeation barrier coatings for fusion reactor is explained. A list of candidate for tritium breeders consists of ceramics containing lithium, for examples, Li 2 O, Li 2 TiO 3 , Li 2 ZrO 3 , Li 4 SiO 4 and LiAlO 2 . The characteristics and form are described. The optimum particle size is from 1 to 10 μm. The production technologies of tritium breeders in the world are stated. Characteristics of ceramics with lithium as tritium breeders are compared. TiC, TiN/TiC, Al 2 O 3 and Cr 2 O 3 -SiO 2 -P 2 O 5 are tritium permeation barrier coating materials. These production methods and evaluation of characteristics are explained. (S.Y.)

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

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

Alternative fuels, fuel cycles, and reactors: are they useful. are they necessary

International Nuclear Information System (INIS)

Spinrad, B.I.

1985-01-01

This chapter discusses reactors, fuel cycles, and fuel production concepts other than those considered conventional in the nuclear community. An attempt is made to look for improvements with the aim of providing cheaper and more durable energy systems, and to contribute toward a solution of the threat of weapons material diversion and weapons proliferation problems. Topics considered include breeding, alternate breeder cycles, alternative reprocessing schemes, symbiotic reactor systems, an interim strategy, and other sources of nuclear fuel. It is determined that the reprocessing of spent fuel is an important safeguard measure in itself

The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

Science.gov (United States)

Dautray, Robert

2011-06-01

The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation

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

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

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)

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

Liquid metal cooled fast breeder nuclear reactor

International Nuclear Information System (INIS)

Scott, D.

1979-01-01

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

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

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)

Safety design study of fast breeder reactors in Japan

International Nuclear Information System (INIS)

Miura, M.; Inagaki, T.

1992-01-01

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

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

Project of a binary breeder reactor and its inherent safety

International Nuclear Information System (INIS)

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

1983-01-01

A core layout for the binary breeder reactor (BBR) is developed based on the results of preliminary burnup calculations. The apparent breeding ratio, in the U 233 /Th fueled inner core, is low due to the accumulation of Pa-233 in the first few months of operation. The loss of reactivity during this time is around 3%. The BBR requires more reactivity control than Pu/U-fueled LMFBRs and the core layout developed has 19 control rod assemblies in the inner core. Three aspects related to the inherent safety of the Binary Breeder Reactor have been studied: the radial distribution of the sodium-void reactivity zone-wise Doppler reactivity and the fractions of delayed neutrons. The results show excellent characteristics for the BRB safety. (Author) [pt

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)

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

International Nuclear Information System (INIS)

1975-01-01

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

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

International Nuclear Information System (INIS)

Yun, Sunghwan; Kim, Sang Ji

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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.

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

Construction and operation of Clinch River Breeder Reactor Plant, docket no. 50-537, Oak Ridge, Roane County, Tennessee

International Nuclear Information System (INIS)

Anon.

1982-01-01

Construction and operation of the Clinch River Breeder Reactor Plant (CRBRP) in Oak Ridge, Tennessee are proposed. The CRBRP would use a liquid-sodium-cooled fast-breeder reactor to produce 975 megawatts of thermal energy (MWt) with the initial core loading of uranium- and plutonium-mixed oxide fuel. This heat would be transferred by heat exchangers to nonradioactive sodium in an intermediate loop and then to a steam cycle. A steam turbine generator would use the steam to produce 380 megawatts of electrical capacity (MWe). Future core design might result in gross power ratings of 1,121 MWt and 439 MWe. Exhaust steam from the turbine generator would be cooled in condensers using two mechanical draft cooling towers. The principal benefit would be the demonstration of the LMFBR concept for commercial use. Electricity generated would be a secondary benefit. Other impacts and effects are discussed

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

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

International Nuclear Information System (INIS)

Obara, Toru; Sekimoto, Hiroshi

1997-01-01

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

Challenges and achievements - Prototype Fast Breeder Reactor construction

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Instrumentation and control of the fast breeder reactors

International Nuclear Information System (INIS)

Juengst, U.

1982-01-01

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

Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

Energy Technology Data Exchange (ETDEWEB)

1979-12-01

The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

International Nuclear Information System (INIS)

1979-12-01

The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials

A solid-breeder blanket and power conversion system for the Mirror Advanced Reactor Study (MARS)

International Nuclear Information System (INIS)

Bullis, R.; Clarkson, I.

1983-01-01

A solid-breeder blanket has been designed for a commercial fusion power reactor based on the tandem mirror concept (MARS). The design utilizes lithium oxide, cooled by helium which powers a conventional steam electric generating cycle. Maintenance and fabricability considerations led to a modular configuration 6 meters long which incorporates two magnets, shield, blanket and first wall. The modules are arranged to form the 150 meter long reactor central cell. Ferritic steel is used for the module primary structure. The lithium oxide is contained in thin-walled vanadium alloy tubes. A tritium breeding ratio of 1.25 and energy multiplication of 1.1 is predicted. The blanket design appears feasible with only a modest advance in current technology

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Preliminary design of a Binary Breeder Reactor; Diseno preliminar de un reactor esferico de quema/cria

Energy Technology Data Exchange (ETDEWEB)

Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, 62550 Jiutepec, Morelos (Mexico)

2014-10-15

A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

Apparatus for power and breeding distribution measurements in breeder reactors

International Nuclear Information System (INIS)

Goldstein, N.P.; Sun, K.H.

1975-01-01

A detection system is disclosed herein for the measurement of power and breeding distribution inside a breeder reactor. Small diameter BeO balls comprising oxides of 235 U and 238 U are inserted into the reactor for activation and withdrawn to be counted in a Ge(Li) counter. Measurements of the activated fission and 239 Np gamma rays yield the desired distributions. (Official Gazette)

Accident analysis of heavy water cooled thorium breeder reactor

International Nuclear Information System (INIS)

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-01-01

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

Accident analysis of heavy water cooled thorium breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

2015-04-16

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

The early realization of fast breeder reactor cycle

International Nuclear Information System (INIS)

Mukai, Kazuo

2007-01-01

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

Clinch River Breeder Reactor Plant Project: construction schedule

International Nuclear Information System (INIS)

Purcell, W.J.; Martin, E.M.; Shivley, J.M.

1982-01-01

The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule

Rough passage in USA for first breeder reactor

International Nuclear Information System (INIS)

Norman, C.

1975-01-01

The present status of the liquid metal fast breeder reactor (LMFBR) programme in the USA is reviewed, including administrative and technical problems. This programme is stated to be the highest priority energy research and development project supported by public funds, estimated at present to cost more than 10,000 million dollars, but it is beset by difficulties and criticism. (U.K.)

Fuel cycle costs for molten-salt reactors

International Nuclear Information System (INIS)

Nagashima, Kikusaburo

1983-01-01

This report describes FCC (fuel cycle cost) estimates for MSCR (molten-salt converter reactor) and MSBR (molten-salt breeder reactor) compared with those for LWRs (PWR and BWR). The calculation is based on the present worth technique with a given discount rate for each cost item, which enables us to make comparison between FCC's for MSCR, MSBR and LWRs. As far as the computational results obtained here are concerned, shown that the FCC's for MSCR and MSBR are 70 -- 60 % lower than the values for LWRs. And it could be said that the FCC for MSCR (Pu-converter) is about 10 % lower than that for MSBR, because of the smaller amount of fissile inventory of MSCR than the inventory of MSBR. (author)

The thorium fuel cycle in water-moderated reactor systems

International Nuclear Information System (INIS)

Critoph, E.

1977-01-01

Current interest in the thorium cycle, as an alternative to the uranium cycle, for water-moderated reactors is based on two attractive aspects of its use - the extension of uranium resources, and the related lower sensitivity of energy costs to uranium price. While most of the scientific basis required is already available, some engineering demonstrations are needed to provide better economic data for rational decisions. Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. There appear to be no major feasibility problems associated with the use of thorium, although development is required in the areas of fuel testing and fuel management. The use of thorium cycles implies recycling the fuel, and the major uncertainties are in the associated costs. Experience in the design and operation of fuel reprocessing and active-fabrication facilities is required to estimate costs to the accuracy needed for adequately defining the range of conditions economically favourable to thorium cycles. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An ''inventory'' of uranium of between 1 and 2Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium), is some two decades

Superalloy applications in the fast breeder reactor

International Nuclear Information System (INIS)

Powell, R.W.

1976-01-01

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

A catalogue of advanced fuel cycles in CANDU-PHW reactors

International Nuclear Information System (INIS)

Veeder, J.; Didsbury, R.

1985-06-01

A catalogue raisonne is presented of various advanced fuel cycle options which have the potential of substantially improving the uranium utilization for CANDU-PHW reactors. Three categories of cycles are: once-through cycles without recovery of fissile materials, cycles that depend on the recovery and recycle of fissile materials in thorium or uranium, cycles that depend primarily on the production of fissile material in a fertile blanket by means of an intense neutron source other than fission, such as an accelerator breeder. Detailed tables are given of the isotopic compositions of the feed and discharge fuels, the logistics of materials and processes required to sustain each of the cycles, and tables of fuel cycle costs based on a method of continuous discounting of cash flow

The economics of fast breeder reactors

International Nuclear Information System (INIS)

Rapin, M.

1990-01-01

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

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

Integral measurement of fission products capture in fast breeder reactors

International Nuclear Information System (INIS)

Martin Deidier, Loick.

1979-12-01

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

Quality assurance in technology development for The Clinch River Breeder Reactor Plant Project

International Nuclear Information System (INIS)

Anderson, J.W.

1980-01-01

The Clinch River Breeder Reactor Plant Project is the nation's first large-scale demonstration of the Liquid Metal Fast Breeder Reactor (LMFBR) concept. The Project has established an overall program of plans and actions to assure that the plant will perform as required. The program has been established and is being implemented in accordance with Department of Energy Standard RDT F 2-2. It is being applied to all parts of the plant, including the development of technology supporting its design and licensing activity. A discussion of the program as it is applied to development is presented

Survey of thorium utilization in power reactor systems

International Nuclear Information System (INIS)

Schwartz, M.H.; Schleifer, P.; Dahlberg, R.C.

1976-01-01

It is clear that thorium-fueled thermal power reactor systems based on current technology can play a vital role in serving present and long-term energy needs. Advanced thorium converters and thermal breeders can provide an expanded resource base from which the world's growing energy demands can be met. Utilization of a symbiotic system of fast breeders and thorium-fueled thermal reactors can be particularly effective in providing low cost power while conserving uranium resources. Breeder reactors are characterized by high capital costs and very low fuel costs since they produce more fuel than they consume. This excess fuel can be used to fuel thermal converter reactors whose capital costs are low. This symbiosis is optimized when 233 U is bred in the fast breeders and then used to fuel high-conversion-ratio thermal converter reactors operating on the thorium-uranium fuel cycle. The thorium-cycle HTGR, after undergoing more than fifteen years of development in both the United States and Europe, provides for the optimum utilization of our limited uranium resources. Other thermal reactor systems, previously operating on the uranium cycle, also show potential in their capability to utilize the thorium cycle effectively

Preliminary safety analysis of molten salt breeder reactor

International Nuclear Information System (INIS)

Cheng Maosong; Dai Zhimin

2013-01-01

Background: The molten salt reactor is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development, which is characterized by remarkable advantages in inherent safety, fuel cycle, miniaturization, effective utilization of nuclear resources and proliferation resistance. ORNL finished the conceptual design of Molten Salt Breeder Reactor (MSBR) based on the design, building and operation of Molten Salt Reactor Experiment (MSRE). Purpose: We attempt to implement the preliminary safety analysis of MSBR in order to provide a reference for the design and optimization of MSBR in the future. Methods: According to the conceptual design of MSBR, a model of safety analysis using point kinetics coupled with the simplified heat transfer mechanism is presented. The model is applied to simulate the transient phenomena of MSBR initiated by an abnormal step reactivity addition and an abnormal ramp reactivity addition at full-power equilibrium condition. Results: The thermal power in the core increases rapidly at the beginning and is accompanied by a rise of the fuel and graphite temperatures after 100, 300, 500 and 600 pcm reactivity addition. The maximum outlet temperature of the fuel in the core is at 1250℃ in 500 pcm reactivity addition, but up to 1350℃ in 600 pcm reactivity addition. The maximum of the power and the temperature are delayed and lower in the ramp reactivity addition rather than in the step reactivity addition. Conclusions: Based on the results, when the reactivity inserted is less than 500 pcm in maximum at full power equilibrium condition, the structural material in Hastelloy-N is not melted and can keep integrity without external control action. And it is necessary to try to avoid inserting a reactivity at short time. (authors)

Fast breeder fuel cycle

International Nuclear Information System (INIS)

1978-09-01

Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

Investigation of base isolation for fast breeder reactor building

International Nuclear Information System (INIS)

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

1989-01-01

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

Status of the DEBENE fast breeder reactor development, March 1979

International Nuclear Information System (INIS)

Daeunert, U.; Kessler, G.

1979-01-01

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

Status of the DEBENE fast breeder reactor development, March 1979

Energy Technology Data Exchange (ETDEWEB)

Daeunert, U; Kessler, G

1979-07-01

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

Evolution of actinides in ThO2 blanket of prototype fast breeder reactor

International Nuclear Information System (INIS)

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

2015-01-01

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

Aiming at super long term application of nuclear energy. Scope and subjects on the water cooled breeder reactor, the 'reduced moderation water reactor'

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2001-01-01

In order to make possible on nuclear energy application for super long term, development of sodium cooling type fast breeder reactor (FBR) has been carried out before today. However, as it was found that its commercialization was technically and economically difficult beyond expectation, a number of nations withdrew from its development. And, as Japan has continued its development, scope of its actual application is not found yet. Now, a research and development on a water cooling type breeder reactor, the reduced moderation water reactor (RMWR)' using LWR technology has now been progressed under a center of JAERI. This RMWR is a reactor intending a jumping upgrade of conversion ratio by densely arranging fuel bars to shift neutron spectrum to faster region. The RMWR has a potential realizable on full-dress plutonium application at earlier timing through its high conversion ratio, high combustion degree, plutonium multi-recycling, and so on. And, it has also feasibility to solve uranium resource problem by realization of conversion ratio with more than 1.0, to contribute to super long term application of nuclear energy. Here was investigated on an effect of reactor core on RMWR, especially of its conversion ratio and plutonium loading on introduction effect as well as on how RMWR could be contributed to reduction of uranium resource consumption, by drawing some scenario on development of power generation reactor and fuel cycle in Japan under scope of super long term with more than 100 years in future. And, trial calculation on power generation cost of the RMWR was carried out to investigate some subjects at a viewpoint of upgrading on economy. (G.K.)

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)

Flexible fuel cycle initiative for the transition period from current reactors to next generation reactors

International Nuclear Information System (INIS)

Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi; Kawamura, Fumio; Shiina, Kouji; Sasahira, Akira

2005-01-01

A sustainable electricity supply by fast breeder reactors (FBRs) is essential to ensure energy security and prevent global warming. Transition from light water reactors (LWRs) to FBRs and establishment of an FBR cycle are indispensable, which requires plutonium (Pu) for the introduction of FBRs. The authors propose advanced system called 'Flexible Fuel Cycle Initiative (FFCI)' which can respond flexibly the future expected technical and social uncertainties, can hold no surplus Pu, and can achieve an economical FBR cycle. In the new concept of FFCI, 2nd LWR reprocessing which would succeed Rokkasho Reprocessing Plant is a simple facility to carry out only uranium (U) removal and residual 'recycle material' is stored or utilized. According to FBRs introduction status, recycle material is immediately treated in an FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of recycle material as a buffer material between LWR and FBR cycles. (author)

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

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

Tritium solubility and permeation in high retention fusion reactor breeder elements

International Nuclear Information System (INIS)

Jakeman, D.

1979-11-01

As an alternative to the current philosophy of reducing the tritium inventory to a minimum by continuously extracting tritium from the breeder of a fusion reactor, an alternative design philosophy is examined in which tritium is contained within high retention breeder elements which can remain in the reactor for a substantial time. To prevent tritium diffusion through the clad of the element it is necessary to maintain a low tritium pressure within the element. Pressures of between 10 4 Pa and 1 Pa appear possible with an element containing a high solubility material provided it is kept below about 400 0 C. This should lead to a leakage into the coolant of between 10 Ci day -1 and 10 4 Ci day -1 which is considerably less than the 10 7 Ci day -1 in present designs. (author)

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

Look at potential options for the fast reactor fuel cycle in the United States

International Nuclear Information System (INIS)

Burch, W.D.

1984-01-01

This paper reviews the status and plans for the fast reactor fuel cycle in the United States, presents some options that are under consideration, and describes how these options are being evaluated at the present time. The United States will undertake some far-reaching examinations of the entire breeder program strategy in the coming year, and the outcome of these reviews cannot be predicted today. In other papers at this conference you have heard various perspectives from both government and industry representatives. The proposed studies to examine the associated fuel cycle strategies as they relate to the overall emerging breeder strategy are described. The present status of and recent developments in the fuel cycle R and D programs will also be summarized and updated in order to present an overall picture of the United States situation

End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

International Nuclear Information System (INIS)

Richardson, K.D.

1987-10-01

Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580 0 F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs

Inherent safe fast breeder reactors and actinide burners, metallic fuel

International Nuclear Information System (INIS)

Dorner, S.; Schumacher, G.

1991-04-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Design Feasible Area on Water Cooled Thorium Breeder Reactor in Equilibrium States

International Nuclear Information System (INIS)

Sidik Permana; Naoyuki Takaki; Hiroshi Sekimoto

2006-01-01

Thorium as supplied fuel has good candidate for fuel material if it is converted into fissile material 233 U which shows superior characteristics in the thermal region. The Shippingport reactor used 233 U-Th fuel system, and the molten salt breeder reactor (MSBR) project showed that breeding is possible in a thermal spectrum. In the present study, feasibility of water cooled thorium breeder reactor is investigated. The key properties such as flux, η value, criticality and breeding performances are evaluated for different moderator to fuel ratios (MFR) and burn-ups. The results show the feasibility of breeding for different MFR and burn-ups. The required 233 U enrichment is about 2% - 9% as charge fuel. The lower MFR and the higher enrichment of 233 U are preferable to improve the average burn-up; however the design feasible window is shrunk. This core shows the design feasible window especially in relation to MFR with negative void reactivity coefficient. (authors)

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

International Nuclear Information System (INIS)

Roberts, R.W.

1977-01-01

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

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)

Methodology for estimating sodium aerosol concentrations during breeder reactor fires

International Nuclear Information System (INIS)

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

1985-01-01

We have devised and applied a methodology for estimating the concentration of aerosols released at building surfaces and monitored at other building surface points. We have used this methodology to make calculations that suggest, for one air-cooled breeder reactor design, cooling will not be compromised by severe liquid-metal fires

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

Investigation of zero-release cycle using fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-11-01

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

Compatibility of 316L stainless steel with tritium breeders for fusion reactors

International Nuclear Information System (INIS)

Broc, M.; Fauvet, P.; Flament, T.; Sannier, J.

1986-06-01

Compatibility problems with structural materials are a concern for the choice of the tritium breeder for fusion reactors. In the frame of the European Programme on Fusion Technology, two types of blankets are considered: liquid (eutectic lithium-lead alloy at 0.68 wt % Li: 17Li83Pb) and solid (lithium aluminate or silicate) breeders. This paper is devoted to compatibility studies of 316L stainless steel with 17Li83Pb alloy and γ-LiA10 2 ceramic

Cold trap dismantling and sodium removal at a fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-15

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

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

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

A manufacturer's view of the US breeder program

International Nuclear Information System (INIS)

Arnold, W.H.

1982-01-01

A liquid metal fast breeder reactor (LMFBR) was selected for development in a program to develop breeder reactors in general. The LMFBR is a sodium-cooled fast reactor which operates at a high conversion ratio of fertile-to-fissile material while generating electricity at a high thermal efficiency. The breeder has the added capacity to operate on the plutonium in Light Water Reactor spent fuel, and on U-238. A governmental/industrial infrastructure must be developed. Criteria for breeder deployment are listed. Construction of the Clinch River Breeder reactor is a necessary step in the progression to a mature breeder. Then the large prototype LMFBR should be built. Foreign collaboration is considered. Finally, a capital cost analysis indicates LMFBR cost-effectiveness

Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

International Nuclear Information System (INIS)

Rubini, L.A.

1978-02-01

An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

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

Protected air-cooled condenser for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Louison, R.; Boardman, C.E.

1981-01-01

The long term residual heat removal for the Clinch River Breeder Reactor Plant (CRBRP) is accomplished through the use of three protected air-cooled condensers (PACC's) each rated at 15M/sub t/ following a normal or emergency shutdown of the reactor. Steam is condensed by forcing air over the finned and coiled condenser tubes located above the steam drums. The steam flow is by natural convection. It is drawn to the PACC tube bundle for the steam drum by the lower pressure region in the tube bundle created from the condensing action. The concept of the tube bundle employs a unique patented configuration which has been commercially available through CONSECO Inc. of Medfore, Wisconsin. The concept provides semi-parallel flow that minimizes subcooling and reduces steam/condensate flow instabilities that have been observed on other similar heat transfer equipment such as moisture separator reheaters (MSRS). The improved flow stability will reduce temperature cycling and associated mechanical fatigue. The PACC is being designed to operate during and following the design basis earthquake, depressurization from the design basis tornado and is housed in protective building enclosure which is also designed to withstand the above mentioned events

Feasible region of design parameters for water cooled thorium breeder reactor

International Nuclear Information System (INIS)

Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

2007-01-01

The performances of a light water cooled thorium breeder reactor have been investigated. A feasible region of fresh fuel enrichment and moderator to fuel ratio (MFR) is found to satisfy the constrains of criticality, breeding, and negative void coefficient for several burnups of discharged fuel. The equilibrium fuel cycle burnup calculation has been performed which is coupled with the cell calculation. The MFR is changed to investigate its effect to the breeding capability and void reactivity coefficient profile for different average discharged burnups. For moderated cases, the conversion ratio (CR) decreases with increasing burnup and MFR. The ratio of fissile inventory in equilibrium core to the initial fissile loading (FIR) has the maximum value at certain burnups depending on the MFR and its value increases with the decreasing MFR. Considering to the breeding capability of the reactor, for burnups of equal to 30 GWd/t or higher, the MFR≤0.3 is needed. For the larger MFR and lower burnups, the void reactivity coefficient becomes more negative with an increasing void fraction. The most negative value of the void reactivity coefficient is obtained at MFR=0.3. (author)

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

International Nuclear Information System (INIS)

1978-06-01

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

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)

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

International Nuclear Information System (INIS)

1983-01-01

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

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

Status of EC solid breeder blanket designs and R and D for demo fusion reactors

International Nuclear Information System (INIS)

Proust, E.; Anzidei, L.; Moons, F.

1994-01-01

Within the European Community Fusion Technology Program two solid breeder blankets for a DEMO reactor are being developed. The two blankets have various features in common: helium as coolant and as tritium purge gas, the martensitic steel MANET as structural material and beryllium as neutron multiplier. The configurations of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder materials are LiAlO 2 or Li 2 ZrO 3 in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li 4 SiO 4 and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium. The main critical issues for both blankets are the behavior of the breeder ceramics and of beryllium under irradiation and the tritium control. Other issues are the low temperature irradiation induced embrittlement of MANET, the mechanical effects caused by major plasma disruptions, and safety and reliability. The R and D work concentrate on these issues. The development of martensitic steels including MANET is part of a separate program. Breeder ceramics and beryllium irradiations have been so far performed for conditions which do not cover the peak values injected in the DEMO blankets. Further irradiations in thermal reactors and in fast reactors, especially for beryllium, are required. An effective tritium control requires the development of permeation barriers and/or of methods of oxidation of the tritium in the main helium cooling systems. First promising results have been obtained also in field of mechanical effects from plasma disruptions and safety and reliability, however further work is required in the reliability field and to validate the codes for the calculations of the plasma disruption effects. (authors). 8 figs., 2 tabs., 53 refs

Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Indian Academy of Sciences (India)

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

Fast breeder reactor research

International Nuclear Information System (INIS)

1975-01-01

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

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Radiological control aspects of the fabrication of the Light Water Breeder Reactor core (LWBR Development Program)

International Nuclear Information System (INIS)

Schultz, B.G.

1979-05-01

A description is presented of the radiological control aspects of the fabrication of the Light Water Breeder Reactor (LWBR) core. Included are the radiological control criteria applied for the design and use of fabrication facilities, the controls and limits imposed to minimize radiaion exposure to personnel, and an evaluation of the applied radiological program in meeting the program objectives. The goal of the LWBR program is to develop the technology to breed in light water reactors so that nuclear fuel may be used significantly more efficiently in these reactors. This technology is being developed by designing and fabricating a breeder reactor core, utilizing thoria (ThO 2 ) and binary thoria--urania (ThO 2 - 233 UO 2 ) fuel, to be operated in the existing pressurized water reactor plant owned by the Department of Energy at Shippingport, Pennsylvania

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

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

Potential use of thorium through fusion breeders in the Indian context

International Nuclear Information System (INIS)

Srinivasan, M.; Basu, T.K.; Subba Rao, K.

1991-01-01

The Indian Nuclear Programme is based on a three stage strategy: the first stage of about 10 GWe comprises of natural uranium fuelled Pressurised Heavy Water Reactors (PHWRs); the second stage would consist of Liquid Metal Cooled Fast Breeder Reactors (LMFBRs) to be fuelled with plutonium generated in the first stage PHWRs and the third stage is envisaged to be based on advanced converters/breeders operating on the Th/U-233 cycle. It has generally been assumed that the initial inventory of U-233 for the third stage reactors would be generated in the blankets of LMFBRs containing thorium. But the success of this strategy depends crucially on the attainment of LMFBR doubling times as short as 14 years. The progress registered in recent years in the magnetic confinement of fusion plasmas has opened up the prospects of developing Fusion Breeders for the direct conversion of fertile 232 Th into fissile 233 U using the 14 MeV neutron released in the (D-T) fusion reaction. A detailed study of the dependence of the 233 U production characteristics as well as energy cost of fissile fuel production of such systems on parameters such as plasma energy gain Q, blanket neutron multiplication has been carried out. The growth rate dynamics of the symbiotic combination of 233 U generating fusion breeders with PHWRs operating on the Th/U-233 cycle in the so called near-breeder regime has been examined. 95% of the energy generated by PHWRs operating with Th/ 233 U fuel would arise from thorium consumption rather than fission of the initially loaded 233 U. A few sub-engineering breakeven fusion breeders producing U-233 at an energy cost well under 200 MeV per atom are adequate to give the requisite nuclear capacity growth rates in conjunction with such near breeder PHWRs. This corresponds to only a 5% diversion of the grid electrical power for the operation of such fusion breeders. In summary the symbiotic combination of a few fusion breeders with a number of PHWRs gives fresh hopes

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

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

Utilization of large electromagnetic pumps in the fast breeder reactors

International Nuclear Information System (INIS)

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

1984-04-01

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

Towards commercial fast breeder reactors the first 1200 MWe unit

International Nuclear Information System (INIS)

Banal, M.; Carle, R.

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

Examination of the use of the breeder reactor to enhance security of energy supply. Final report

International Nuclear Information System (INIS)

Jones, G.S.; Chow, B.G.

1980-01-01

Early commercialization of the breeder reactor has been justified in terms of its promise for lessening a nation's dependence on imported uranium. The relationship between breeder timing (assuming a vigorous but not instantaneous breeder penetration rate) and the reduction in uranium requirements is examined using Japan as an example. Comparing early and delayed breeder introductions (2006 and 2026 respectively in the high growth projections), the difference in cumulative uranium requirements is slight through the year 2045; for both scenarious, annual requirements remain high. The early breeder introduction strategy is more attractive when it incorporates a 30 percent improved LWR. However, given the unlikelihood of sufficient funding for the development of both designs, the optimal strategy for Japan is stockpiling uranium for an improved LWR and delaying introduction of the breeder

Accuracy and computational time of a hierarchy of growth rate definitions for breeder reactor fuel

International Nuclear Information System (INIS)

Maudlin, P.J.; Borg, R.C.; Ott, K.O.

1979-01-01

For a hierarchy of four logically different definitions for calculating the asymptotic growth of fast breeder reactor fuel, an investigation is performed concerning the comparative accuracy and computational effort associated with each definition. The definition based on detailed calculation of the accumulating fuel in an expanding park of reactors asymptotically yields the most accurate value of the infinite time growth rate, γ/sup infinity/, which is used as a reference value. The computational effort involved with the park definition is very large. The definition based on the single reactor calculation of the equilibrium surplus production rate and fuel inventory gives a value for γ/sup infinity of comparable accuracy to the park definition and uses significantly less central processor unit (CPU) time. The third definition is based on a continuous treatment of the reactor fuel cycle for a single reactor and gives a value for γ/sup infinity/ that accurately approximates the second definition. The continuous definition requires very little CPU time. The fourth definition employs the isotopic breeding worths, w/sub i//sup */, for a projection of the asymptotic growth rate. The CPU time involved in this definition is practically nil if its calculation is based on the few-cycle depletion calculation normally performed for core design and critical enrichment evaluations. The small inaccuracy (approx. = 1%) of the breeding-worth-based definition is well within the inaccuracy range that results unavoidably from other sources such as nuclear cross sections, group constants, and flux calculations. This fully justifies the use of this approach in routine calculations

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

International Nuclear Information System (INIS)

Fischer, J.J.

1978-01-01

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

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

International Nuclear Information System (INIS)

1977-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Scaller, K; Vrillon, B

1980-02-01

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

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

International Nuclear Information System (INIS)

Scaller, K.; Vrillon, B.

1980-01-01

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

Reliability modeling of Clinch River breeder reactor electrical shutdown systems

International Nuclear Information System (INIS)

Schatz, R.A.; Duetsch, K.L.

1974-01-01

The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

Breeder design for enhanced performance and safety characteristics

International Nuclear Information System (INIS)

Fischer, G.J.; Atefi, B.; Yang, J.W.; Galperin, A.; Segev, M.

1980-01-01

A fast breeder reactor design has been created which offers a considerably extended fuel cycle and excellent performance characteristics. An example of a core designed to operate on a ten-year fuel cycle is described in some detail. Use of metal fuel along with a moderator such as beryllium oxide dispersed throughout the core provides both design flexibility and safety advantages such as a strong Doppler feedback and limited sodium void reactivity gain. Local power variations are small for the entire cycle; control requirements are also modest, and fuel cycle costs are low

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

Scram reliability under seismic conditions at the Experimental Breeder Reactor II

International Nuclear Information System (INIS)

Roglans, J.; Wang, C.Y.; Hill, D.J.

1993-01-01

A Probabilistic Risk Assessment of the Experimental Breeder Reactor II has recently been completed. Seismic events are among the external initiating events included in the assessment. As part of the seismic PRA a detailed study has been performed of the ability to shutdown the reactor under seismic conditions. A comprehensive finite element model of the EBR-II control rod drive system has been used to analyze the control rod system response when subjected to input seismic accelerators. The results indicate the control rod drive system has a high seismic capacity. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure

Simplified inelastic analysis methods applied to fast breeder reactor core design

International Nuclear Information System (INIS)

Abo-El-Ata, M.M.

1978-01-01

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1978-07-01

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

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 ...

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

A capital cost reduction study on the fast breeder reactor plant

International Nuclear Information System (INIS)

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

1991-01-01

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

Estimation, comparison, and evaluation of advanced fission power reactor generation costs

International Nuclear Information System (INIS)

Waddell, J.D.

1977-01-01

The study compares the high-temperature gas-cooled reactor (HTGR), the gas-cooled fast reactor (GCFR), the molten-salt breeder reactor (MSBR), the light water breeder reactor (LWBR), and the heavy water reactor (HWR) with proposed light water reactors (LWR) and liquid-metal fast breeder reactors (LMFBR). The relative electrical generation costs, including the effects of the introduction of advanced reactor fuel cycles into the U.S. nuclear power economy, were projected through the year 2030. The study utilized the NEEDS computer code which is a simulation of the U.S. nuclear power economy. The future potential electrical generation costs and cumulative consumption of uranium ore were developed using characterizations of the advanced systems. The reactor-fuel cycle characterizations were developed from literature reviews and personal discussions with the proponents of the various systems. The study developed a ranking of the concepts based on generation costs and uranium consumption

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Ageing management practice in Fast Breeder Test Reactor

International Nuclear Information System (INIS)

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

2006-01-01

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

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.

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

U.S. reference paper on national decisions on breeder development and deployment

International Nuclear Information System (INIS)

1978-10-01

Factors involved in making national decisions on the deployment of breeder reactor systems are identified in terms of a nation's potential for electrification, capital resources, the available industrial and manpower infrastructure and importance attached to energy independence and the degree to which a breeder program can help realize this objective in the time scale of interest. The specific factors analysed are: the high capital cost of the breeder and the one-time transition costs to bring the breeder to maturity the high breeder research, development and demonstration costs, the impact of discount rate, and the fuel cycle costs, e.g. indigeneous facilities or purchase of services. A principal conclusion of this paper is that nations may find it more economical to continue to deploy LWRs for a number of years rather than to consider the breeder option because of the initial high breeder capital cost and high breeder R and D costs

Accelerator molten-salt breeder reactor

International Nuclear Information System (INIS)

Furukawa, Kazuo; Kuroi, Hideo; Kato, Yoshio; Oomichi, Toshihiko.

1979-01-01

Purpose: To obtain fission products and to transmute transuranium elements and other radioactive wastes by the use of Accelerator Molten-Salt Breeder Reactor. Constitution: Beams from an accelerator pipe at one end of a target vessel is injected through a window into target molten salts filled inside of the target vessel. The target molten salts are subjected to pump recycling or spontaneous convection while forcively cooled by blanket molten salts in an outer vessel. Then, energy is recovered from the blanket molten salts or the target molten salts at high temperatures through electric power generation or the like. Those salts containing such as thorium 232 and uranium 238 are used as the blanket molten salts so that fission products may be produced by neutrons generated in the target molten salts. PbCl 2 -PbF 2 and LiF-BeF 2 -ThF 4 can be used as the target molten salts and as the blanket molten salts respectively. (Seki, T.)

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)

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

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.%

Compatibility of steels for fast breeder reactor in high temperature sodium

International Nuclear Information System (INIS)

Yuhara, Shunichi

1981-01-01

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

Transition period fuel cycle from current to next generation reactors for Japan

International Nuclear Information System (INIS)

Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi; Kawamura, Fumio; Shiina, Kouji; Sasahira, Akira

2007-01-01

Long-term energy security and global warming prevention can be achieved by a sustainable electricity supply with next generation fast breeder reactors (FBRs). Current light water reactors (LWRs) will be replaced by FBRs and FBR cycle will be established in the future considering the limited amount of uranium (U) resource. The introduction of FBRs requires plutonium (Pu) recovered from LWR spent fuel. The authors propose advanced system named Flexible Fuel Cycle Initiative (FFCI)' which can supply enough Pu and hold no surplus Pu, can respond flexibly the future technical and social uncertainties, and can achieve an economical FBR cycle. FFCI can simplify the 2nd LWR reprocessing facility for Japan (after Rokkasho Reprocessing Plant) which only carries out U removal from LWR spent fuel. Residual 'Recycle Material' is, according to FBRs introduction status, immediately treated in the FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of Recycle Material as a buffer material between LWR and FBR cycles. (author)

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

International Nuclear Information System (INIS)

Tomabechi, K.

1975-01-01

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

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

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.)

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

International Nuclear Information System (INIS)

Sackett, J.I.

1983-01-01

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

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

International Nuclear Information System (INIS)

1975-01-01

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

A computer program for calculation of the fuel cycle in pressurized water reactors

International Nuclear Information System (INIS)

Solanilla, R.

1976-01-01

The purpose of the FUCEFURE program is two-fold: first, it is designed to solve the problem of nuclear fuel cycle cost in one pressurized light water reactor calculation. The code was developed primarily for comparative and sensitivity studies. The program contains simple correlations between exposure and available depletion data used to predict the uranium and plutonium content of the fuel as a function of the fuel initial enrichment. Second, it has been devised to evaluate the nuclear fuel demand associated with an expanding nuclear power system. Evaluation can be carried out at any time and stage in the fuel cycle. The program can calculate the natural uranium and separate work requirements of any final and tails enrichment. It also can determine the nuclear power share of each reactor in the system when a decision has been made about the long-term nuclear power installations to be used and the types of PWR and fast breeder reactor characteristics to be involved in them. (author)

Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

International Nuclear Information System (INIS)

Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

2000-01-01

Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

U.S. leans toward denatured thorium cycle

International Nuclear Information System (INIS)

Smock, R.

1977-01-01

Denatured thorium appears to be the most promising among the nonproliferating alternatives to the plutonium cycle, which the Carter Administration is trying to cancel. Criteria for a better system include uranium utilization comparable to current light water reactors and minimal separation of fissile material into the waste stream. Comparisons with other systems conclude that thorium is preferable because it can lead to an acceptable fast breeder. The thorium cycle can be placed in energy centers for sensitive facilities and can also be introduced into ongoing light water systems. Reprocessing can be handled in the centers, where thorium can be mixed with plutonium for use in reactors within the center, while light water reactors operate on the outside. Any fuel leaving the center would be unsuitable for weapons. Later adaptation to in-center fast breeders will extend energy supplies, although a thorium breeder will be less efficient than a plutonium fast breeder. Denatured thorium is a technical answer to a complex political problem, but those in the nuclear industry see the U.S. goal of a nonproliferating fuel as futile in the light of world politics and breeder efforts in other countries

Evaluation of the influence of seismic restraint characteristics on breeder reactor piping systems

International Nuclear Information System (INIS)

Mello, R.M.; Pollono, L.P.

1979-01-01

For the Clinch River Breeder Reactor Plant (CRBRP) heat transport system piping within the reactor containment building, dynamic analyses of the piping loops have been performed to study the effect of restraint stiffness on the dynamic behavior of the piping. In addition, analysis and testing of typical CRBRP restraint system components have been performed for the purpose of quantifying and verifying the basic characteristics of the restraints used in the piping system dynamic analysis

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

Fast breeder reactors secondary piping potential sodium leakage rate assessment

International Nuclear Information System (INIS)

Alicino, F.; Cardini, S.

1989-01-01

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

Diversion analysis and safeguards measures for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Persiani, P.J.

1981-10-01

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

Immediate relation of ING to fast breeder reactor programs

Energy Technology Data Exchange (ETDEWEB)

Lewis, W B

1969-07-01

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

Immediate relation of ING to fast breeder reactor programs

International Nuclear Information System (INIS)

Lewis, W.B.

1969-01-01

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

Advanced converters and reactors

International Nuclear Information System (INIS)

Haefele, W.; Kessler, G.

1984-01-01

As Western Europe and most countries of the Asia-Pacific region (except Australia) have only small natural uranium resources, they must import nuclear fuel from the major uranium supplier countries. The introduction of advanced converter and breeder reactor technology allows a fuel utilization of a factor of 4 to 100 higher than with present low converters (LWRs) and will make uranium-importing countries less vulnerable to price jumps and supply stops in the uranium market. In addition, breeder-reactor technology will open up a potential that can cover world energy requirements for several thousand years. The enormous development costs of advanced converter and breeder technologies can probably be raised only by highly industrialized countries. Those highly industrialized countries that have little or no uranium resources (Western Europe, Japan) will probably be the first to introduce this advanced reactor technology on a commercial scale. A number of small countries and islands will need only small power reactors with inherent safety capabilities, especially in the beginning of their nuclear energy programs. For economic reasons, the fuel cycle services should come from large reprocessing centers of countries having sufficiently large nuclear power programs or from international fuel cycle centers. (author)

Application of mass-predictions to isotope-abundances in breeder-reactor cores

CERN Document Server

Kirchner, G

1981-01-01

The decay-heat and isotope composition of breeder reactor-cores is calculated at normal shut-down, and a core disintegration event. Using the ORIGEN-code, the influence of the most neutron-rich fission-yield nuclei is studied. Their abundances depend on the assumption about the nuclear data (mass and half-lives). The total decay-heat is not changed from any technical viewpoint. (15 refs).

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

International Nuclear Information System (INIS)

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

1975-01-01

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

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

Clinch River Breeder Reactor secondary control rod system

International Nuclear Information System (INIS)

McKeehan, E.R.; Sim, R.G.

1977-01-01

The shutdown system for the Clinch River Breeder Reactor (CRBR) includes two independent systems--a primary and a secondary system. The Secondary Control Rod System (SCRS) is a new design which is being developed by General Electric to be independent from the primary system in order to improve overall shutdown reliability by eliminating potential common-mode failures. The paper describes the status of the SCRS design and fabrication and testing activities. Design verification testing on the component level is largely complete. These component tests are covered with emphasis on design impact results. A prototype unit has been manufactured and system level tests in sodium have been initiated

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

Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

Directory of Open Access Journals (Sweden)

Stanisz Przemysław

2015-09-01

Full Text Available The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System and LEADER (Lead-cooled European Advanced Demonstration Reactor projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs, and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

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.

EBR-II [Experimental Breeder Reactor-II] system surveillance using pattern recognition software

International Nuclear Information System (INIS)

Mott, J.E.; Radtke, W.H.; King, R.W.

1986-02-01

The problem of most accurately determining the Experimental Breeder Reactor-II (EBR-II) reactor outlet temperature from currently available plant signals is investigated. Historically, the reactor outlet pipe was originally instrumented with 8 temperature sensors but, during 22 years of operation, all these instruments have failed except for one remaining thermocouple, and its output had recently become suspect. Using pattern recognition methods to compare values of 129 plant signals for similarities over a 7 month period spanning reconfiguration of the core and recalibration of many plant signals, it was determined that the remaining reactor outlet pipe thermocouple is still useful as an indicator of true mixed mean reactor outlet temperature. Application of this methodology to investigate one specific signal has automatically validated the vast majority of the 129 signals used for pattern recognition and also highlighted a few inconsistent signals for further investigation

International strategies for breeder development

International Nuclear Information System (INIS)

Zaleski, C.P.; Zebroski, E.L.

1992-01-01

This paper studies the perspectives of breeder reactors development. The near term context has led some experts to the conclusion that breeder reactor technology is too far ahead of its time. Some have compared breeders to the supersonic airplane, Concorde: good technical performance but failure in its economic dimensions. In this paper, the author points out the major shortcomings of such an assessment which may be valid in the short time. However, with a short-term market-dominated perspective that uses an 8% discount rate, one can neglect every thing that is going to happen in 50 years. 6 refs., 11 figs

Recommendations concerning models and parameters best suited to breeder reactor environmental radiological assessments

International Nuclear Information System (INIS)

Miller, C.W.; Baes, C.F. III; Dunning, D.E. Jr.

1980-05-01

Recommendations are presented concerning the models and parameters best suited for assessing the impact of radionuclide releases to the environment by breeder reactor facilities. These recommendations are based on the model and parameter evaluations performed during this project to date. Seven different areas are covered in separate sections

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

International Nuclear Information System (INIS)

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

1980-01-01

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

It is now time to proceed with a gas-cooled breeder reactor (GBR) demonstration plant

International Nuclear Information System (INIS)

Chermanne, J.

1976-01-01

Since 1969, the GBRA has been making studies to provide evidence on questions which were not clear regarding the Gas-cooled Breeder Reactor: design feasibility and performance, safety, strategy and economics, and R and D necessary for a demonstration plant. Studies were carried out on a 1200-MW(e) commercial reference design with pin fuel, which was also used as a basis for a definition of the GBR demonstration plant. During the six years, a great deal of information has been generated at GBRA and it confirms the forecasts of the promoters of the Gas-cooled Breeder Reactor that the GBR is an excellent reactor from all points of view: design - the reactor can be engineered without major difficulty, using present techniques. As far as fuel is concerned, LMFBR fuel technology is applicable plus limited specific development effort. Performance - the GBR is the best breeder with oxide fuel and using conventional techniques. The strategy studies carried out at GBRA have clearly shown that a high performance breeder such as the GBR is absolutely required in large quantities by the turn of the century in order to avoid dependence on natural uranium resources. Regarding safety, a major step forward has been made when an ad hoc group on GBR safety, sponsored by the EEC, concluded that no major difficulties were anticipated which would prevent the GBR reaching adequate safety standards. Detailed economic assessments performed on an item-to-item basis have shown that the cost of a GBR with its high safety standard is about the same as that of an HTR. One can therefore conclude that, with the present cost of natural uranium, the GBR is competitive with the LWRs. Owing to the very limited R and D effort necessary and the obvious safety, economic and strategic advantages of the concept, it is concluded that the development and construction of a GBR demonstration plant must be started now if one wants to secure an adequate energy supply past the turn of the century. (author)

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

International Nuclear Information System (INIS)

Delporte, E.; Vanderborck, Y.

1988-01-01

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Alternative fuel cycles and non-proliferation aspects

International Nuclear Information System (INIS)

Kessler, G.

1980-10-01

The most important physical characteristics of the U/Pu and the Th/U fuel cycles and the technical data of the most significant converter reactors operating with Th/U fuel are outlined in the report. Near breeders as well as breeders with a thermal neutron spectrum are briefly discussed, and the potential of breeders with fast neutron spectra in the Th/U fuel is outlined. The essential criteria for the comparison of the alternative fuel cycles with the reference Pu/U cycle are the consumption of natural uranium, the numbers of U-233 producing and U-233 consuming converter reactors and the amounts of fission material transported and handled within the fuel cycle (reprocessing, refabrication). Although the alternative U/Th fuel cycles are feasible with some advantages and some disadvantages as compared to the reference U/Pu cycle, not much experience has so far been gathered with pilot plants of the fuel cycle. The respective status in reprocessing, refabrication and waste disposal is briefly discussed. Finally, a comparison of the risk potential inherent in secular storage is presented and questions of resistance to proliferation and of safeguards of the U/Th cycle are discussed

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Raw materials problems in connection with fast breeder type reactors

International Nuclear Information System (INIS)

Hirsch, H.; Kreusch, J.

1981-01-01

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

Molten-salt reactor strategies viewed from fuel conservation effect, (1)

International Nuclear Information System (INIS)

Furuhashi, Akira

1976-01-01

Saving of material requirements in the long-term fuel cycle is studied by introducing molten-salt reactors with good neutron economy into a projection of nuclear generating capacity in Japan. In this first report an examination is made on the effects brought by the introduction of molten-salt converter reactors starting with Pu which are followed by 233 U breeders of the same type. It is shown that the sharing of some Pu in the light water- and fast breeder-reactor system with molten-salt reactors provides a more rapid transition to the self-supporting, breeding cycle than the simple fast breeding system, thus leading to an appreciable fuel conservation. Considerations are presented on the strategic repartition of generating capacity among reactor types and it is shown that all of the converted 233 U should be promptly invested to molten-salt breeders to quickly establish the dual breeding system, instead of recycling to converters themselves. (auth.)

Control-rod, pressure and flow-induced accident and transient analysis of a direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor

International Nuclear Information System (INIS)

Kitoh, Kazuaki; Koshizuka, Seiichi; Oka, Yoshiaki

1996-01-01

The features of the direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor (SCFBR) are high thermal efficiency and simple reactor system. The safety principle is basically the same as that of an LWR since it is a water-cooled reactor. Maintaining the core flow is the basic safety requirement of the reactor, since its coolant system is the one through type. The transient behaviors at control rod, pressure and flow-induced abnormalities are analyzed and presented in this paper. The results of flow-induced transients of SCFBR were reported at ICONE-3, though pressure change was neglected. The change of fuel temperature distribution is also considered for the analysis of the rapid reactivity-induced transients such as control rod withdrawal. Total loss of flow and pump seizure are analyzed as the accidents. Loss of load, control rod withdrawal from the normal operation, loss of feedwater heating, inadvertent start of an auxiliary feedwater pump, partial loss of coolant flow and loss of external power are analyzed as the transients. The behavior of the flow-induced transients is not so much different from the analyses assuming constant pressure. Fly wheels should be equipped with the feedwater pumps to prolong the coast-down time more than 10s and to cope with the total loss of flow accident. The coolant density coefficient of the SCFBR is less than one tenth of a BWR in which the recirculation flow is used for the power control. The over pressurization transients at the loss of load is not so severe as that of a BWR. The power reaches 120%. The minimum deterioration heat flux ratio (MDHFR) and the maximum pressure are sufficiently lower than the criteria; MDHFR above 1.0 and pressure ratio below 1.10 of 27.5 MPa, maximum pressure for operation. Among the reactivity abnormalities, the control rod withdrawal transient from the normal operation is analyzed

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

Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Yanagi, Yoshihiko; Sato, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Kikuchi, Shigeto; Kuroda, Toshimasa; Kosaku, Yasuo; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

2001-11-01

Within a design study of a fusion DEMO reactor aiming at demonstrating technologies of fusion power plant, supercritical water is applied as a coolant of solid breeder blanket to attain high thermal efficiency. The blanket has multi-layer composed of solid breeder pebbles (Li{sub 2}O) and neutron multiplier pebbles (Be) which are radially separated by cooling panels. The first wall and the breeding region are cooled by supercritical water below and above the pseudo-critical temperature, respectively. Temperature distribution and tritium breeding ratio (TBR) have been estimated by one-dimensional nuclear and thermal calculations. The local TBR as high as 1.47 has been obtained after optimization of temperature distribution in the breeder region under the following conditions: neutron wall loading of 5 MW/m{sup 2}, {sup 6}Li enrichment of 30% and coolant temperature at inlet of breeder region of 380degC. In the case of the higher coolant temperature 430degC of the breeder region the local TBR was reduced to be 1.40. This means that the net TBR higher than 1.0 could be expected with the supercritical-water-cooled blanket, whose temperature distribution in the breeder region would be optimized by following the coolant temperature, and where a coverage of the breeder region is assumed to be 70%. (author)

A review of fast reactor programme in Japan

International Nuclear Information System (INIS)

Matsuno, Y.; Bando, S.

1981-03-01

The fast breeder reactor development project in Japan made progress in the past year, and will be continued in the next fiscal 1981. The scale of efforts both in budget and personnel will be similar to those in fiscal 1980. The budget for R and D works and for the construction of the fast breeder prototype reactor ''Monju'' will be approximately 20 billion yen and 27 billion yen, respectively, excluding the wage of the personnel concerned. The number of the technical personnel currently engaging in fast breeder reactor development in the Power Reactor and Nuclear Fuel Development Corp. is about 530. As for the experimental fast reactor ''Joyo'', three operational cycles at 75 MWt have been completed in August, 1980, and the fourth cycle has started in March, 1981. As for the prototype reactor ''Monju'', progress was made toward the construction, and the environmental impact statement on the reactor was approved by the authorities concerned. The studies on the preliminary design of large LMFBRs have been made by the PNC and also by power companies. The design study carried out by the PNC is concerned with a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of the commissioning of ''Monju''. The highlights and topics in the development activities for fast breeder reactors in the past twelve months are summarized in this report. (Kako, I.)

Gas-cooled reactors: the importance of their development

International Nuclear Information System (INIS)

Kasten, P.R.

1979-06-01

The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U 3 O 8 before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production

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)

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

International Nuclear Information System (INIS)

Raj, B.

2007-01-01

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

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

International Nuclear Information System (INIS)

Kani, Yoshio

2003-01-01

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

A thorium breeder reactor concept for optimal energy extraction from uranium and thorium

International Nuclear Information System (INIS)

Jagannnathan, V.; Lawande, S.V.

1999-01-01

An attractive thorium breeder reactor concept has been evolved from simple physics based guidelines for induction of thorium in a major way in an otherwise enriched uranium reactor. D 2 O moderator helps to maximise reactivity for a given enrichment. A relatively higher flux level compared to LWRs offers the advantage of higher rate of 233 U production in thoria rods. Thus fresh thoria clusters consider no feed enrichment. In an equilibrium core, a full batch of pure thoria clusters are loaded during each fuel cycle. They undergo irradiation for about one year duration. By this time they accumulate nearly 70% of the asymptotic stable concentration of 233 U, if they face a flux level of the order of 10 14 n/cm 2 /sec. In the next fuel cycle, these thoria rods in ring cluster form are juxtaposed with the fresh enriched fuel rods, also in ring cluster form. Such integrated fuel assemblies are then irradiated for four or five fuel cycles, at the end of which U as well as Th rods attain a reasonably high burnup of about 30-32 MWD/kg. The core characteristics are quite attractive. The core excess reactivity remains low due to large thoria inventory which makes the net burnup reactivity load to be below 1%. The core is capable of being operated in an annual batch mode of operation like a LWR. The control requirement during power operation is negligible. Xenon over-ride requirement is low and can be managed by partial withdrawal of a few thoria clusters. Void reactivity is nearly zero or negative by the optimum design of the fuel cluster. Reactivity changes due to temperatures of fuel, coolant and moderator are also small. (author)

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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.)

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

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

International Nuclear Information System (INIS)

Sackett, J.I.

1983-01-01

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

Breeder--now or never

International Nuclear Information System (INIS)

Murphy, P.M.

1978-01-01

The timing of the commercial introduction of the LMFBR (Liquid Metal Fast Breeder Reactor) will be an important factor in its ability to supply a significant fraction of the nation's future electrical needs. The number of breeders we can build initially will be limited by the size of our low-cost urnium resources and by the rate at which LWR's (Light Water Reactors) are placed in service. Since this uranium resource is fixed in size while electrical demand will grow geometrically, it is clear that the sooner the breeder is introduced commercially the larger will be the fraction of electrical demand that it can supply. An early commercial introduction on an adequate scale requires full-scale resumption of LWR construction and redirection of LMFBR development programs toward a near-term commercial prototype

Breeder: now or never

International Nuclear Information System (INIS)

Murphy, P.M.

1978-01-01

The timing of the commercial introduction of the liquid metal fast breeder reactor (LMFBR) will be an important factor in its ability to supply a significant fraction of the nation's future electrical needs. The number of breeders we can build initially will be limited by the size of our low-cost uranium resources and by the rate at which light water reactors (LWRs) are placed in service. Since this uranium resource is fixed in size while electrical demand will grow geometrically, it is clear that the sooner the breeder is introduced commercially the larger will be the fraction of electrical demand it can supply. An early commercial introduction on an adequate scale requires full-scale resumption of LWR construction and redirection of LMFBR development programs toward a near-term commercial prototype

The thorium fuel cycle

International Nuclear Information System (INIS)

Merz, E.R.

1977-01-01

The utilization of the thorium fuel cycle has long since been considered attractive owing to the excellent neutronic characteristics of 233 U, and the widespread and cheap thorium resources. Rapidly increasing uranium prices, public reluctance for widespread Pu recycling and expected delays for the market penetration of fast breeders have led to a reconsideration of the thorium fuel cycle merits. In addition, problems associated with reprocessing and waste handling, particularly with re-fabrication by remote handling of 233 U, are certainly not appreciably more difficult than for Pu recycling. To divert from uranium as a nuclear energy source it seems worth while intensifying future efforts for closing the Th/ 233 U fuel cycle. HTGRs are particularly promising for economic application. However, further research and development activities should not concentrate on this reactor type alone. Light- and heavy-water-moderated reactors, and even future fast breeders, may just as well take advantage of a demonstrated thorium fuel cycle. (author)

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

International Nuclear Information System (INIS)

1993-01-01

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

A middle evaluation report on R and D subjects in fiscal year 2000. Evaluation subject: 'investigation on actualization strategy of fast breeder reactor cycle'

International Nuclear Information System (INIS)

2000-08-01

The Japan Nuclear Cycle Development Institute (JNC) consulted the titled middle evaluation to the Subject Evaluation Committee (SEC) according to the Schematic indication on practice procedure of evaluation common to the generalized national R and D' and so on. By receiving the consult, SEC on wastes treatment and disposal carried out evaluation of this subject on basis of documents proposed from JNC and discussions at SEC according to an evaluation procedure determined by SEC. This subject was already investigated at the third group on establishment of long term program on new nuclear energy and concluded its promotion, it can be said that aim and meaning of its R and D is clear. And, at a viewpoint of middle- and long-term business program of JNC, it can also be positioned to be at an important mission. In order to carry out flexible response to versatile needs in future for giving a meaning of development strategy of fast breeder reactor (FBR) cycle, under full understanding of variability of a promise condition for the aim and meaning of this research such as environment around development of FBR cycle development, R nad D of the program must be promoted. In this program picking-out of every subject and research program are adequately promoted to be enough evaluated for the middle results. Here was summarized by the evaluation results with documents proposed by JNC. As a result of the evaluation, it was shown that as general directionality of this program was judged to be valid. (G.K.)

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

International Nuclear Information System (INIS)

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

1981-04-01

Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators

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

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

Method for cooling a breeder reactor and breeder reactor for applying the method

International Nuclear Information System (INIS)

Gast, K.

1977-01-01

The fuel assemblies of the fission zone and the breeder subassemblies in the radial breeding blanket are supported on a double bottom. The coolant gets into the interspace of the double bottom below the blanket. This part of the space is separated from the interspace of the double bottom below the fission zone. Each breeder subassembly consists of a twin tube. The coolant enters the inner tube, flows through it upwards in axial direction, is then deflected on the upper end, and afterwards, in the annulus of the twin tube, flows down again in axial direction into the inlet region, below the double bottom. From there on it flows upwards through the fuel assemblies of the fission zone from below. Thereby a uniformly high coolant outlet temperature is obtained. (DG) [de

Characterizing the tribological behaviour of fast breeder reactor materials

International Nuclear Information System (INIS)

Depierre, J.; Raffailhac, J.

1984-04-01

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

Clinch river breeder reactor plant steam generator water quality

Energy Technology Data Exchange (ETDEWEB)

Van Hoesen, D; Lowe, P A

1975-07-01

The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

Clinch river breeder reactor plant steam generator water quality

International Nuclear Information System (INIS)

Van Hoesen, D.; Lowe, P.A.

1975-01-01

The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

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

International Nuclear Information System (INIS)

2014-11-01

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

Steam generator tubing development for commercial fast breeder reactors

International Nuclear Information System (INIS)

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

1981-01-01

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

A new combination of membranes and membrane reactors for improved tritium management in breeder blanket of fusion machines

International Nuclear Information System (INIS)

Demange, D.; Staemmler, S.; Kind, M.

2011-01-01

Tritium used as fuel in future fusion machines will be produced within the breeder blanket. The tritium extraction system recovers the tritium to be routed into the inner-fuel cycle of the machine. Accurate and precise tritium accountancy between both systems is mandatory to ensure a reliable operation. Handling in the blanket huge helium flow rates containing tritium as traces in molecular and oxide forms is challenging both for the process and the accountancy. Alternative tritium processes based on combinations of membranes and membrane reactors are proposed to facilitate the tritium management. The PERMCAT process is based on counter-current isotope swamping in a palladium membrane reactor. It allows recovering tritium efficiently from any chemical species. It produces a pure hydrogen stream enriched in tritium of advantage for integration upstream of the accountancy stage. A pre-separation and pre-concentration stage using new zeolite membranes has been studied to optimize the whole process. Such a combination could improve the tritium processes and facilitate accountancy in DEMO.

Design and R and D activities on ceramic breeder blanket for fusion experimental reactors in JAERI

International Nuclear Information System (INIS)

Kurasawa, T.; Takatsu, H.; Sato, S.; Nakahira, M.; Furuya, K.; Hashimoto, T.; Kawamura, H.; Kuroda, T.; Tsunematsu, T.; Seki, M.

1995-01-01

Design and R and D activities on ceramic breeder blanket of a fusion experimental reactor have been progressed in JAERI. A layered pebble bed type ceramic breeder blanket with water cooling is a prime candidate concept. Design activities have been concentrated on improvement of the design by conducting detailed analyses and also by fabrication procedure consideration based on the current technologies. A wide variety of R and Ds have also been conducted in accordance with the design activities. Development of fabrication technology of the blanket box structure and its mechanical testing, elementary testing on thermal performances of the pebble bed, and engineering-oriented material tests of breeder and beryllium pebbles are the main achievements during the last two years. (orig.)

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

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

International Nuclear Information System (INIS)

Dee, J.B.; Macken, T.

1977-01-01

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

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

Linear accelerator-breeder (LAB): a preliminary analysis and proposal

International Nuclear Information System (INIS)

1976-01-01

The development and demonstration of a Linear Accelerator-Breeder (LAB) is proposed. This would be a machine which would use a powerful linear accelerator to produce an intense beam of protons or deuterons impinging on a target of a heavy element, to produce spallation neutrons. These neutrons would in turn be absorbed in fertile 238 U or 232 Th to produce fissile 239 Pu or 233 U. Though a Linear Accelerator-Breeder is not visualized as competitive to a fast breeder such as the LMFBR, it would offer definite benefits in improved flexibility of options, and it could probably be developed more rapidly than the LMFBR if fuel cycle problems made this desirable. It is estimated that at a beam power of 300 MW a Linear Accelerator-Breeder could produce about 1100 kg/year of fissile 239 Pu or 233 U, which would be adequate to fuel from 2,650 to 15,000 MW(e) of fission reactor capacity depending on the fuel cycle used. A two-year design study is proposed, and various cost estimates are presented. The concept of the Linear Accelerator-Breeder is not new, having been the basis for a major AEC project (MTA) a number of years ago. It has also been pursued in Canada starting from the proposal for an Intense Neutron Generator (ING) several years ago. The technical basis for a reasonable design has only recently been achieved. The concept offers an opportunity to fill an important gap that may develop between the short-term and long-term energy options for energy security of the nation

FRG paper on assessment of fuel cycles

International Nuclear Information System (INIS)

1979-01-01

The paper deals with the assessment of the nuclear fuel cycle under different aspects: Assured energy supply, economy, environmental aspects, and non-proliferation philosophy. The results of an assessment of nuclear fuel variants along these lines for several types of commercial reactors (light-water reactors, heavy-water reactors, high-temperature reactors, and fast breeders) are presented in tables

Reactivity control system of a passively safe thorium breeder pebble bed reactor

International Nuclear Information System (INIS)

Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

2014-01-01

Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF 3 gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MW th passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel, between 100 cm

Reactivity control system of a passively safe thorium breeder pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Wols, F.J., E-mail: f.j.wols@tudelft.nl; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

2014-12-15

Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF{sub 3} gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MW{sub th} passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel

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

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

International Nuclear Information System (INIS)

Tsoulfanidis, N.; Jankhah, M.H.

1979-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

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

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

International Nuclear Information System (INIS)

1992-01-01

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

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

Implementation of multivariable control techniques with application to Experimental Breeder Reactor II

International Nuclear Information System (INIS)

Berkan, R.C.

1990-06-01

After several successful applications to aerospace industry, the modern control theory methods have recently attracted many control engineers from other engineering disciplines. For advanced nuclear reactors, the modern control theory may provide major advantages in safety, availability, and economic aspects. This report is intended to illustrate the feasibility of applying the linear quadratic Gaussian (LQG) compensator in nuclear reactor applications. The LQG design is compared with the existing classical control schemes. Both approaches are tested using the Experimental Breeder Reactor 2 (EBR-2) as the system. The experiments are performed using a mathematical model of the EBR-2 plant. Despite the fact that the controller and plant models do not include all known physical constraints, the results are encouraging. This preliminary study provides an informative, introductory picture for future considerations of using modern control theory methods in nuclear industry. 10 refs., 25 figs

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.

Analysis of thorium and uranium fuel cycles in an iso-breeder lead fast reactor using extended-EQL3D procedure

International Nuclear Information System (INIS)

Fiorina, Carlo; Krepel, Jiri; Cammi, Antonio; Franceschini, Fausto; Mikityuk, Konstantin; Ricotti, Marco Enrico

2013-01-01

Highlights: ► Extension of EQL3D procedure to calculate radio-toxicity and decay heat. ► Characterization of uranium- and thorium-fueled LFR from BOL to equilibrium. ► Safety improvements for a LFR in a closed thorium cycle. ► Advantages of thorium-fueled LFR in terms of decay heat and radio-toxicity generation. ► Safety, decay heat and radio-toxicity concerns for a Th–Pu beginning-of-life core. - Abstract: Use of thorium in fast reactors has typically been considered as a secondary option, mainly thanks to a possible self-sustaining thorium cycle already in thermal reactors and due to the limited breeding capabilities compared to U–Pu in the fast neutron energy range. In recent years nuclear waste management has become more important, and the thorium option has been reconsidered for the claimed potential to burn transuranic waste and the lower build-up of hazardous isotopes in a closed cycle. To ascertain these claims and their limitations, the fuel cycle isotopic inventory, and associated waste radio-toxicity and decay heat, should be quantified and compared to the case of the uranium cycle using realistic core configurations, with complete recycle of all the actinides. Since the transition from uranium to thorium fuel cycles will likely involve a transuranic burning phase, this transition and the challenges that the evolving fuel actinide composition presents, for instance on reactor feedback parameters, should also be analyzed. In the present paper, these issues are investigated based on core physics analysis of the Lead-cooled Fast Reactor ELSY, performed with the fast reactor ERANOS code and the EQL3D procedure allowing full-core characterization of the equilibrium cycle and the transition cycles. In order to compute radio-toxicity and decay heat, EQL3D has been extended by developing a new module, which has been assessed against ORIGEN-S and is presented here. The capability of the EQL3D procedure to treat full-core 3D geometries allowed to

Structural analysis of the Upper Internals Structure for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Houtman, J.L.

1979-01-01

The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides control of core outlet flow to prevent severe thermal transients from occuring at the reactor vessel and primary heat transport outlet piping, provides instrumentation to monitor core performance, provides support for the control rod drivelines, and provides secondary holddown of the core. All of the structural analysis aspects of assuring the UIS is structurally adequate are presented including simplified and rigorous inelastic analysis methods, elevated temperature criteria, environmental effects on material properties, design techniques, and manufacturing constraints

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

Advanced thorium cycles in LWRs and HWRs

International Nuclear Information System (INIS)

Radkowsky, A.

The main aspects of advanced thorium cycles in LWRs and HWRs are reviewed. New concepts include the seed blanket close packed heavy water breeder, the light water seed blanket thorium burner and self-induced thorium cycle in CANDU type reactors. (author)

A comparison of fusion breeder/fission client and fission breeder/fission client systems for electrical energy production

International Nuclear Information System (INIS)

Land, R.J.; Parish, T.A.

1983-01-01

A parametric study that evaluated the economic performance of breeder/client systems is described. The linkage of the breeders to the clients was modelled using the stockpile approach to determine the system doubling time. Since the actual capital costs of the breeders are uncertain, a precise prediction of the cost of a breeder was not attempted. Instead, the breakeven capital cost of a breeder relative to the capital cost of a client reactor was established by equating the cost of electricity from the breeder/client system to the cost of a system consisting of clients alone. Specific results are presented for two breeder/client systems. The first consisted of an LMFBR with LWR clients. The second consisted of a DT fusion reactor (with a 238 U fission suppressed blanket) with LWR clients. The economics of each system was studied as a function of the cost of fissile fuel from a conventional source. Generally, the LMFBR/LWR system achieved relatively small breakeven capital cost ratios; the maximum ratio computed was 2.2 (achieved at approximately triple current conventional fissile material cost). The DTFR/LWR system attained a maximum breakeven capital cost ratio of 4.5 (achieved at the highest plasma quality (ignited device) and triple conventional fissile cost)

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.

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)

01 - The sustainable management of radioactive materials with fourth-generation reactors

International Nuclear Information System (INIS)

2012-12-01

This publication first explains the reasons for the development of fourth-generation nuclear systems: the issue of resources, the issue of used fuels, technologies for fast breeder fourth-generation reactors, the alternative of HFC-cooled light water reactors, the potential market for fast-breeder reactors, strategies in different countries. Then, the authors address the strategy development for these fourth-generation reactors: requirements, safety, economic competitiveness, resistance to proliferation, flexible management of nuclear resources, development scheme (sodium or gas-based heat transfer), and fuel cycle. They finally discuss the possible transition scenarios towards this new generation

Preliminary study of the thermo-hydraulic behaviour of the binary breeder reactor

International Nuclear Information System (INIS)

Silveira Luz, M. da; Ferreira, W.J.

1984-06-01

Continuing the development of the Binary Breeder Reactor, its physical configuration and the advantages of differents types of spacers are analysed. In order to simulate the thermo-hydraulic behaviour and obtain data for a preliminary evaluation of the core geometry, the COBRA III C code was used to study the effects of the lenght and diameter of the fuel element, the coolant inlet temperature, the system pressure, helicoidal pitch and the pitch to diameter ratio. (Author) [pt

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

Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

International Nuclear Information System (INIS)

Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

2012-09-01

selection of chemistry controls is vital for NPPs with liquid metal cooled reactors. This paper highlights principles and approaches to chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors. The recommendations on how to arrange chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors are based taking into account: - the experience with operation of fossil power industry; - secondary side water chemistry of lead-bismuth eutectics cooled nuclear reactors at submarines; - steam/water cycles of NPPs with sodium cooled fast breeders BN-350 and BN-600; - secondary water chemistry at conventional NPPs with WER, RBMK and some other reactors. (authors)

Frequency response testing at Experimental Breeder Reactor II using discrete-level periodic signals

International Nuclear Information System (INIS)

Rhodes, W.D.; Larson, H.A.

1990-01-01

The Experimental Breeder Reactor 2 (EBR-2) reactivity-to-power frequency-response function was measured with pseudo-random, discrete-level, periodic signals. The reactor power deviation was small with insignificant perturbation of normal operation and in-place irradiation experiments. Comparison of results with measured rod oscillator data and with theoretical predictions show good agreement. Moreover, measures of input signal quality (autocorrelation function and energy spectra) confirm the ability to enable this type of frequency response determination at EBR-2. Measurements were made with the pseudo-random binary sequence, quadratic residue binary sequence, pseudo-random ternary sequence, and the multifrequency binary sequence. 10 refs., 7 figs., 3 tabs

Preliminary study of molten-salt breeder reactor using the WIMS-D/4 code

International Nuclear Information System (INIS)

Oliveira, J.T. de.

1994-01-01

The features and operation of the Molten-Salt Breeder Reactors - MSBR -are presented. Information about the conversion, breeding and Thorium burn-up chain with the differential equations for the isotopes is given. A few group constants for the different cells of the Single Fluid MSBR 1000 MWe are also presented. The WIMS methods, resonant treatment, leakage corrections, burn up chains, input and output data are commented. (author). 55 refs

Upgrading program of the experimental fast reactor Joyo

International Nuclear Information System (INIS)

Yoshida, A.; Yogo, S.

2001-01-01

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

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-15

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

Comparison of the radiological impacts of thorium and uranium nuclear fuel cycles

International Nuclear Information System (INIS)

Meyer, H.R.; Witherspoon, J.P.; McBride, J.P.; Frederick, E.J.

1982-03-01

This report compares the radiological impacts of a fuel cycle in which only uranium is recycled, as presented in the Final Generic Environmental Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel in Light Water Cooled Reactors (GESMO), with those of the light-water breeder reactor (LWBR) thorium/uranium fuel cycle in the Final Environmental Statement, Light Water Breeder Reactor Program. The significant offsite radiological impacts from routine operation of the fuel cycles result from the mining and milling of thorium and uranium ores, reprocessing spent fuel, and reactor operations. The major difference between the impacts from the two fuel cycles is the larger dose commitments associated with current uranium mining and milling operations as compared to thorium mining and milling. Estimated dose commitments from the reprocessing of either fuel type are small and show only moderate variations for specific doses. No significant differences in environmental radiological impact are anticipated for reactors using either of the fuel cycles. Radiological impacts associated with routine releases from the operation of either the thorium or uranium fuel cycles can be held to acceptably low levels by existing regulations