WorldWideScience

Sample records for burner fast reactors

  1. Fast Burner Reactor Devoted to Minor Actinide Incineration

    International Nuclear Information System (INIS)

    This study proposes a new fast reactor core concept dedicated to plutonium and minor actinide burning by transmutation. This core has a large power level of ∼1500 MW(electric) favoring the economic aspect. To promote plutonium and minor actinide burning as much as possible, total suppression of 238U, which produces 239Pu by conversion, and large quantities of minor actinides in the core are desirable. Therefore, the 238U-free fuel is homogeneously mixed with a considerable quantity of minor actinides.From the safety point of view, both the Doppler effect and the coolant (sodium) void reactivity become less favorable in a 238U-free core. To preserve these two important safety parameters on an acceptable level, a hydrogenated moderator separated from the fuel and nuclides, such as W or 99Tc, is added to the core in the place of 238U. Tungsten and 99Tc have strong capture resonances at appropriate energies, and 99Tc itself is a long-lived fission product to be transmuted with profit.This core allows the achievement of a consumption rate of ∼100 kg/TW(electric).h of transuranic elements, ∼70 kg/TW(electric).h for plutonium (due to 238U suppression), and 30 to 35 kg/TW(electric).h for minor actinides. In addition, ∼14 kg/TW(electric).h of 99Tc is destroyed when this element is present in the core (the initial loading of 99Tc is >4000 kg in the core).The activity of newly designed subassemblies has also been investigated in comparison to standard fast reactor subassemblies (neutron sources, decay heat, and gamma dose rate). Finally, a transmutation scenario involving pressurized water reactors and minor actinide-burning fast reactors has been studied to estimate the necessary proportion of burner reactors and the achievable radiotoxicity reduction with respect to a reference open cycle

  2. Neutron economy and transmutation performance of coupling system of fast reactor and a-burner

    International Nuclear Information System (INIS)

    Neutron economy and transmutation performance are examined for a fast reactor (FR), a PWR, and an A-Burner which transmutes Minor-Actinide (MA) in a well-thermalized neutron field and a slightly hard neutron field optimized for the burn-up of 246Cm. The neutron economies of the FR and the A-Burner are, respectively, favorable and acceptable to transmute MA. The coupling system of both the reactors can reduce 8.1 ton of MA to almost zero within 60 years by using one FR and one A-Burner. This coupling system is expected to achieve the final goal of transmutation, i.e., to make us free from the geological disposal. (authors)

  3. ZZ WPPR-FR-MOX/BNCMK, Benchmark on Pu Burner Fast Reactor

    International Nuclear Information System (INIS)

    Description of program or function: In order to intercompare the characteristics of the different reactors considered for Pu recycling, in terms of neutron economy, minor actinide production, uranium content versus Pu burning, the NSC Working Party on Physics of Plutonium Recycling (WPPR) is setting up several benchmark studies. They cover in particular the case of the evolution of the Pu quality and Pu fissile content for Pu recycling in PWRs; the void coefficient in PWRs partly fuelled with MOX versus Pu content; the physics characteristics of non-standard fast reactors with breeding ratios around 0.5. The following benchmarks are considered here: - Fast reactors: Pu Burner MOX fuel, Pu Burner metal fuel; - PWRs: MOX recycling (bad quality Pu), Multiple MOX recycling

  4. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  5. Actinide transmutation using inert matrix fuels versus recycle in a low conversion fast burner reactor

    International Nuclear Information System (INIS)

    infinite fuel residence time. In previous work we have shown that the amount of fluence required to achieve a unit of burnup in yttrium stabilized ZrO2 based IMF with 85 w/o zirconium oxide and 15 w/o minor actinides (MA) and plutonium increases dramatically beyond 750 MWd/kgIHM (75% burnup). In this paper we discuss the repository implications for recycle of actinides in LWR's using this type of IMF and compare this to actinide recycle in a low conversion fast burner reactor. We perform the analysis over a finite horizon of 100 years, in which reprocessing of spent LWR fuel begins in 2020. Reference [1] C. Lombardi and A. Mazzola, Exploiting the plutonium stockpiles in PWRs by using inert matrix fuel, Annals of Nuclear Energy. 23 (1996) 1117-1126. [2] U. Kasemeyer, J.M. Paratte, P. Grimm and R. Chawla, Comparison of pressurized water reactor core characteristics for 100% plutonium-containing loadings, Nuclear Technology. 122 (1998) 52-63. [3] G. Ledergerber, C. Degueldre, P. Heimgartner, M.A. Pouchon and U. Kasemeyer, Inert matrix fuel for the utilisation of plutonium, Progress in Nuclear Energy. 38 (2001) 301-308. [4] U. Kasemeyer, C. Hellwig, J. Lebenhaft and R. Chawla, Comparison of various partial light water reactor core loadings with inert matrix and mixed oxide fuel, Journal of Nuclear Materials. 319 (2003) 142-153. [5] E.A. Schneider, M.R. Deinert and K.B. Cady, Burnup simulations of an inert matrix fuel using a two region, multi-group reactor physics model, in Proceedings of the physics of advanced fuel cycles, PHYSOR 2006, Vancouver, BC, 2006. [6] E.A. Schneider, M.R. Deinert and K.B. Cady, Burnup simulations and spent fuel characteristics of ZRO2 based inert matrix fuels, Journal of Nuclear Materials. 361 (2007) 41-51. (authors)

  6. Lead coolant for the fast reactor-burner with a hard neutron spectrum

    International Nuclear Information System (INIS)

    The possibility of the efficiency increasing of the junior actinides burn-up on the account of their fission in the fast reactor by using the Pb-208 lead isotope as a coolant is considered. The calculation relative values of the fission reduction: rates and capture the neutrons by the Np-237, Am-241 and Am-243 nuclei in the fast reactor blanket for different coolants: sodium, lead, lead of natural composition and lead with 100%-Pb-208 content are presented. The radioactivity, induced during one year of the reactor operation and its decay during the time up to 1000 years is calculated for the lead coolants of various isotopic composition

  7. Concept of fast reactors-plutonium burners and their fuel cycle

    International Nuclear Information System (INIS)

    In this report the concept is considered of a closed fuel cycle of nuclear power, consisting of thermal and fast reactors, providing utilization of practically all actinides produced in the nuclear power. The major calculation results and the ways to form fast reactors cores for effective actinides burning are presented. The existing limitations on the fresh fuel composition by heat release and radiation characteristics are given. The calculation studies results on the fuel cycle characteristics at repeated fuel recycling in a system of VVER-1000 and BN-800 types reactors are presented. The calculations were carried out for different type cores of the BN-800 reactor -with oxide fuel of increased enrichment and with fuel without uranium-238, varying the burn-out level and the decay time of spent fuel. (author)

  8. Fast burner reactor benchmark results from the NEA working party on physics of plutonium recycle

    International Nuclear Information System (INIS)

    As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed; fuel cycle scenarios using either PUREX/TRUEX (oxide fuel) or pyrometallurgical (metal fuel) separation technologies were specified. These benchmarks were designed to evaluate the nuclear performance and radiotoxicity impact of a transuranic-burning fast reactor system. International benchmark results are summarized in this paper; and key conclusions are highlighted

  9. Status of the EC-FP7 Project ARCAS: Comparing the economics of accelerator-driven systems and fast reactors as minor actinide burners

    International Nuclear Information System (INIS)

    The ARCAS Project aims to compare, on a technological and economical basis, accelerator-driven systems and fast reactors as minor actinide burners. It is split into five work packages: the reference scenario definition, the fast reactor system definition, the accelerator-driven system definition, the fuel reprocessing and fabrication facilities definition and the economical comparison. This paper summarises the status of the project and its five work packages. (authors)

  10. Some aspects of risk reduction strategy by multiple recycling in fast burner reactors of the plutonium and minor actinide inventories

    International Nuclear Information System (INIS)

    This paper shows the impact of recycling light water reactor (LWR) mixed oxide (MOX) fuel in a fast burner reactor on the plutonium (Pu) and minor actinide (MA) inventories and on the related radioactivities. Reprocessing of the targets for multiple recycling will become increasingly difficult as the burnup increases. Multiple recycling of Pu + MA in fast reactors is a feasible option which has to be studied very carefully: the Pu (except the isotopes Pu-238 and Pu-240), Am and Np levels decrease as a function of the recycle number, while the Cm-244 level accumulates and gradually transforms into Cm-245. Long cooling times (10 + 2 years) are necessary with aqueous processing. The paper discusses the problems associated with multiple reprocessing of highly active fuel types and particularly the impact of Pu-238, Am-241 and Cm-244 on the fuel cycle operations. The calculations were performed with the zero-dimensional ORIGEN-2 code. The validity of the results depends on that of the code and its cross-section library. The time span to reduce the initial inventory of Pu + MA by a factor of 10 amounts to 255 years when average burnups are limited to 150 GW.d t-1 (tonne). (orig.)

  11. Some aspects of risk reduction strategy by multiple recycling in fast burner reactors of the plutonium and minor actinide inventories

    International Nuclear Information System (INIS)

    The paper shows the impact of recycling LWR-MOX fuel in a fast burner reactor on the plutonium (Pu) and minor actinide (MA) inventories and on the related radio activities. Reprocessing of the targets for multiple recycling will become increasingly difficult as the burn up increases. Multiple recycling of Pu + MA in fast reactors is a feasible option which has to be studied very carefully: the Pu (except the isotopes Pu-238 and Pu-240), Am and Np levels decrease as a function of the recycle number, while the Cm-244 level accumulates and gradually transforms into Cm-245. Long cooling times (10 + 2 years) are necessary with aqueous processing. The paper discusses the problems associated with multiple reprocessing of highly active fuel types and particularly the impact of Pu-238, Am-241 and Cm-244 on the fuel cycle operations. The calculations were performed with the zero-dimensional ORIGEN-2 code. The validity of the results depends on that of the code and its cross section library. The time span to reduce the initial inventory of Pu + MA by a factor of 10, amounts to 255 years when average burn ups are limited to 150 GWd t-1. (orig.)

  12. High conversion burner type reactor

    International Nuclear Information System (INIS)

    Purpose: To simply and easily dismantle and reassemble densified fuel assemblies taken out of a high conversion ratio area thereby improve the neutron and fuel economy. Constitution: The burner portion for the purpose of fuel combustion is divided into a first burner region in adjacent with the high conversion ratio area at the center of the reactor core, and a second burner region formed to the outer circumference thereof and two types of fuels are charged therein. Densified fuel assemblies charged in the high conversion ratio area are separatably formed as fuel assemblies for use in the two types of burners. In this way, dense fuel assembly is separated into two types of fuel assemblies for use in burner of different number and arranging density of fuel elements which can be directly charged to the burner portion and facilitate the dismantling and reassembling of the fuel assemblies. Further, since the two types of fuel assemblies are charged in the burner portion, utilization factor for the neutron fuels can be improved. (Kamimura, M.)

  13. Economic Analysis of Symbiotic Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

    International Nuclear Information System (INIS)

    A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: 1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, 2) a 'singletier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and 3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall bus-bar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

  14. Cooperative Russian-French experiment on plutonium-enriched fuels for fast burner reactor

    International Nuclear Information System (INIS)

    Various kinds of nuclear fuels with an increased plutonium content are under study according to the program including three stages: fabrication, irradiation in BOR-60 reactor, post-irradiation examination. Flowsheets for fabricating pelletized and vibrocompacted fuels of UPu0.45O2, UPu0.45N, UPu0.6N, PuN + ZrN, PuO2 + MgO are presented along with basic fuel properties. The irradiation of oxide fuel is carried out in an individual irradiation device at rated maximum temperature of the fuel at the beginning of irradiation equal to 2100 deg C. The irradiation of nitride fuel and the fuel based on inert matrices is performed in the other device with the aim of limitation of maximum temperature by the value of 1550 deg C. The duration of irradiation for all fuel types constitutes 750 EFPD. Fuel element charge in Bor-60 reactor core was realized in 2000

  15. Economic Analyiss of "Symbiotic" Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Kent Alan [ORNL; Shropshire, David E. [Idaho National Laboratory (INL)

    2009-01-01

    A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a

  16. Comparative study of fast critical burner reactors and subcritical accelerator driven systems and the impact on transuranics inventory in a regional fuel cycle

    International Nuclear Information System (INIS)

    Research highlights: → Double-strata fuel cycle has a potential to minimize transuranics mass in Europe. → European Minor Actinides legacy can be reduced down to 0 before the end of century. → 40% higher capacity needed to burn MA for fast critical reactor then for EFIT fleet. → Na cooled fast reactor cores with high content of MA and low CR have been assessed. → Fast critical and ADS-EFIT reactors show comparable MA transmutation performance. - Abstract: In the frame of Partitioning and Transmutation (P and T) strategies, many solutions have been proposed in order to burn transuranics (TRU) discharged from conventional thermal reactors in fast reactor systems. This is due to the favourable feature of neutron fission to capture cross section ratio in a fast neutron spectrum for most TRU. However the majority of studies performed use the Accelerator Driven Systems (ADS), due to their potential flexibility to utilize various fuel types, loaded with significant amounts of TRU having very different Minor Actinides (MA) over Pu ratios. Recently the potential of low conversion ratio critical fast reactors has been rediscovered, with very attractive burning capabilities. In the present paper the burning performances of two systems are directly compared: a sodium cooled critical fast reactor with a low conversion ratio, and the European lead cooled subcritical ADS-EFIT reactor loaded with fertile-free fuel. Comparison is done for characteristics of both the intrinsic core and the regional fuel cycle within a European double-strata scenario. Results of the simulations, obtained by use of French COSI6 code, show comparable performance and confirm that in a double strata fuel cycle the same goals could be achieved by deploying dedicated fast critical or ADS-EFIT type reactors. However the critical fast burner reactor fleet requires ∼30-40% higher installed power then the ADS-EFIT one. Therefore full comparative assessment and ranking can be done only by a

  17. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  18. Thermal-hydraulics of actinide burner reactors

    International Nuclear Information System (INIS)

    As a part of conceptual study of actinide burner reactors, core thermal-hydraulic analyses were conducted for two types of reactor concepts, namely (1) sodium-cooled actinide alloy fuel reactor, and (2) helium-cooled particle-bed reactor, to examine the feasibility of high power-density cores for efficient transmutation of actinides within the maximum allowable temperature limits of fuel and cladding. In addition, calculations were made on cooling of actinide fuel assembly. (author)

  19. Advanced Burner Reactor Preliminary NEPA Data Study

    International Nuclear Information System (INIS)

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  20. Advanced Burner Reactor Preliminary NEPA Data Study.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  1. Fuel cycle analysis of TRU or MA burner fast reactors with variable conversion ratio using a new algorithm at equilibrium

    Energy Technology Data Exchange (ETDEWEB)

    Salvatores, Massimo [CEA Cadarache, 13108 St-Paul-Lez-Durance (France); Argonne National Laboratory, NE Division, Argonne, IL 60439 (United States)], E-mail: massimo.salvatores@cea.fr; Chabert, Christine [CEA Cadarache, 13108 St-Paul-Lez-Durance (France); Fazio, Concetta [Forschungszentrum Karlsruhe GmbH, P.O. Box 3640, 76021 Karlsruhe (Germany); Hill, Robert [Argonne National Laboratory, NE Division, Argonne, IL 60439 (United States); Peneliau, Yannick; Slessarev, Igor [CEA Cadarache, 13108 St-Paul-Lez-Durance (France); Yang, Won Sik [Argonne National Laboratory, NE Division, Argonne, IL 60439 (United States)

    2009-10-15

    Partitioning and Transmutation (P and T) strategies assessment and implementation play a key role in the definition of advanced fuel cycles, in order to insure both sustainability and waste minimization. Several options are under study worldwide, and their impact on core design and associated fuel cycles are under investigation, to offer a rationale to down selection and to streamline efforts and resources. Interconnected issues like fuel type, minor actinide content, conversion ratio values, etc. need to be understood and their impact quantified. Then, from a practical point of view, studies related to advanced fuel cycles require a considerable amount of analysis to assess performances both of the reactor cores and of the associated fuel cycles. A physics analysis should provide a sound understanding of major trends and features, in order to provide guidelines for more detailed studies. In this paper, it is presented an improved version of a generalization of the Bateman equation that allows performing analysis at equilibrium for a large number of systems. It is shown that the method reproduces very well the results obtained with full depletion calculations. The method is applied to explore the specific issue of the features of the fuel cycle parameters related to fast reactors with different fuel types, different conversion ratios (CR) and different ratios of Pu over minor actinide (Pu/MA) in the fuel feed. As an example of the potential impact of such analysis, it is shown that for cores with CR below {approx}0.8, the increase of neutron doses and decay heat can represent a significant drawback to implement the corresponding reactors and associated fuel cycles.

  2. Fuel cycle analysis of TRU or MA burner fast reactors with variable conversion ratio using a new algorithm at equilibrium

    International Nuclear Information System (INIS)

    Partitioning and Transmutation (P and T) strategies assessment and implementation play a key role in the definition of advanced fuel cycles, in order to insure both sustainability and waste minimization. Several options are under study worldwide, and their impact on core design and associated fuel cycles are under investigation, to offer a rationale to down selection and to streamline efforts and resources. Interconnected issues like fuel type, minor actinide content, conversion ratio values, etc. need to be understood and their impact quantified. Then, from a practical point of view, studies related to advanced fuel cycles require a considerable amount of analysis to assess performances both of the reactor cores and of the associated fuel cycles. A physics analysis should provide a sound understanding of major trends and features, in order to provide guidelines for more detailed studies. In this paper, it is presented an improved version of a generalization of the Bateman equation that allows performing analysis at equilibrium for a large number of systems. It is shown that the method reproduces very well the results obtained with full depletion calculations. The method is applied to explore the specific issue of the features of the fuel cycle parameters related to fast reactors with different fuel types, different conversion ratios (CR) and different ratios of Pu over minor actinide (Pu/MA) in the fuel feed. As an example of the potential impact of such analysis, it is shown that for cores with CR below ∼0.8, the increase of neutron doses and decay heat can represent a significant drawback to implement the corresponding reactors and associated fuel cycles.

  3. Advanced fuel cycle scenario study in the European context using different burner reactor concepts

    International Nuclear Information System (INIS)

    Different types of fast spectrum dedicated burners have been proposed for the management of radioactive wastes in the frame of various advanced fuel cycle scenarios. Accelerator-driven systems (ADS) and critical low conversion ratio fast reactors have been studied, e.g. within the European context. A potential alternative system is a fusion-fission hybrid (FFH). In the present study, a sodium-cooled fast reactor driven by a D-T fusion neutron source, the subcritical advanced burner reactor (SABR) system is considered. In order to intercompare the different systems, a systematic study is under way. The performances of the two types of systems (SABR, ADS) will be compared from a minor actinide (MA) or transuranic (i.e. Pu+MA) burning potential point of view. The present paper reports preliminary results of the first phase of study, i.e. the comparison of SABR and ADS when used as minor actinides burners. (authors)

  4. Ball plasma dynamics for FBX BURNER reactor

    International Nuclear Information System (INIS)

    The authors have been conducting fundamental experiments on the moving plasma balls in the major axis direction. This has a fundamental importance for the quasi-steady fusion reactor scheme FBX BURNER. This configuration is the descendants of Spheromak type nuclear fusion scheme but with long toroidal field coils. The main issues have been focused onto the dynamic stability of the moving plasma balls. The second issue is the collision between two successive plasma balls as shown. This shows a basic result from an optical measurement. The comparison with magnetic measurements shows peculiar but interesting features of the system. The dimension of the plasma injector is 2 m in length as a whole. It is in a coaxial configuration. The external electrode is 150 mm in inner diameter and the inner electrode is 50 mm in diameter. This is attached to a 1 m insulation reservoir with same inner diameter. An axial magnetic field up to 0.1 Tesla is applied on the discharge with a current of up to 10 kA in few ms. The authors show their experimental and numerical simulation results on these problems

  5. Linear accelerator for burner-reactor

    International Nuclear Information System (INIS)

    Future development of nuclear power engineering depends on the successful solution of two key problems of safety and utilization of high level radioactive wastes (HLRW) of atomic power plants (APP). Modern methods of HLRW treatment involve solidification, preliminary storing for a period of 30-50 years necessary for the decay of long-living nuclides and final burial in geological formations several hundred meters below the ground surface. The depth burial of the radioactive wastes requires complicated under ground constructions. It's very expensive and doesn't meet modern ecological requirements. Alternative modern and more reasonable methods of APP HLRW treatment are under consideration now. One of the methods involves separation of APP waste radionuclides for use in economy with subsequent transmutation of the long-living isotopes into the short-living ones by high-intensity neutron fluxes generated by proton accelerators. The installation intended for the long-living radionuclides transmutation into the short-living ones is called burner-reactor. It can be based on the continuous regime proton accelerator with 1.5 GeV energy, 0.3 A current and beam mean power of 450 MW. The preferable type of the proton accelerator with the aforementioned parameters is the linear accelerator

  6. An improved method for fuel cycle analysis at equilibrium and its application to the study of fast burner reactors with variable conversion ratio

    International Nuclear Information System (INIS)

    Studies related to advanced fuel cycles require a considerable amount of analysis to assess performances both of the reactor cores and of the associated fuel cycles. A physics analysis should provide a sound understanding of major trends and features, in order to provide guidelines for detailed studies. In this paper we present an improved version of a generalization of the Bateman equation that allows performing analysis at equilibrium for a large number of systems. It is shown that the method reproduces very well the results obtained with full depletion calculations. The method is applied to explore the features of the fuel cycles parameters related to fast reactors with different fuel types, different conversion ratios (CR) and different MA/Pu ratios in the fuel feed. It is shown that for cores with CR below ∼0.8, the increase of neutron doses and decay heat can represent a significant drawback to implement the corresponding reactors and associated fuel cycles. (authors)

  7. Fast reactors and nonproliferation

    International Nuclear Information System (INIS)

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

  8. Fast Reactor Development Strategy in China

    International Nuclear Information System (INIS)

    As one of the largest developing countries, China needs a reliable energy supplement. At the same time, China should improve the energy structure to decrease CO2 emissions. Nuclear and renewable energies are the main solutions to these issues. According to the research results, the nuclear capacity should increase to 400 GW(e) up to 2050. Fast reactors must be developed considering the limitation of uranium resources. In order to deploy fast reactor technology, the ‘experimental reactor, demonstration reactor and commercial reactor’ strategy has been suggested. China has finished the construction of the China Experimental Fast Reactor (CEFR) and gained necessary experience about fast reactors. The China Institute of Atomic Energy (CIAE) has begun to design the CFR-600, a 600 MW(e) demonstration fast reactor. This reactor will be put into operation before 2025. After that, a larger commercial reactor will be constructed. Besides fast reactors, all of other key sectors of fuel cycle will be developed at the same time such as reprocessing, fast reactor fuel, etc. There are two main tasks of fast reactors, one of which is to raise the utility ratio of uranium, and the other one is to transmute the long life waste of light water reactors. The fast reactor will be designed as a breeder and burner, respectively. (author)

  9. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  10. Analysis of Reactor Deployment Scenarios with Introduction of SFR Breakeven Reactors and Burners Using DANESS Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2008-01-15

    Using the DANESS code newly employed for future scenario analysis, reactor deployment scenarios with the introduction of sodium cooled fast reactors(SFRs) having different conversion ratios in the existing PWRs dominant nuclear fleet have been analyzed to find the SFR deployment strategy for replacing PWRs with the view of a spent fuel reduction and an efficient uranium utilization through its reuse in a closed nuclear fuel cycle. Descriptions of the DANESS code and how to use are briefly given from the viewpoint of its first application. The use of SFRs and recycling of TRUs by reusing PWR spent fuel leads to the substantial reduction of the amount of PWR spent fuel and environmental burden by decreasing radiotoxicity of high level waste, and a significant improvement on the natural uranium resources utilization. A continuous deployment of burners effectively decreases the amount of PWR spent fuel accumulation, thus lightening the burden for PWR spent fuel management. An introduction of breakeven reactors effectively reduces the uranium demand through producing excess TRU during the operation, thus contributing to a sustainable nuclear power development. With SFR introduction starting in 2040, PWRs will remain as a main power reactor type till 2100 and SFRs will be in support of waste minimization and fuel utilization.

  11. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  12. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  13. Commercialization of fast reactors

    International Nuclear Information System (INIS)

    Comparative analysis has been performed of capital and fuel cycle costs for fast BN-type and pressurized light water VVER-type reactors. As a result of materials demand and components costs comparison of NPPs with VVER-1000 and BN-600 reactors, respectively, conclusion was made, that under equal conditions of the comparison, NPP with fast reactor had surpassed the specific capital cost of NPP with VVER by about 30 - 40 %. Ways were determined for further decrease of this difference, as well as for the fuel cycle cost reduction, because at present it is higher than that of VVER-type reactors. (author)

  14. Use of fast reactors for actinide transmutation

    International Nuclear Information System (INIS)

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

  15. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  16. Carbide and Nitride Fuels for Advanced Burner Reactor

    International Nuclear Information System (INIS)

    The impacts of the mixed carbide and nitride fuels on the core performances and passive safety features of TRU burner were assessed and comapred with the metallic and oxide fuels. Targeting the potential design goals adopted in the Advanced Burner Reactor core concepts, the alternative TRU burner concepts were developed by loading carbide and nitride fuels. The neutron spectrum is softer than that of the metal core, but harder than that of the oxide core, and the core performance parameters such as fuel residence time, discharge burnup, flux level, etc are generally between the values of the metal and oxide cores. The margin to fuel melt was significantly increased because of the high thermal conductivity and high melting temperature, and hence there is an additional room to improve the thermal efficiency by increasing the operating temperature. The changed fuel composition affected the kinetics parameters and reactivity feedback coefficients, but the variations were minimal. The reduced core height decreases the sodium void worth, and the high thermal conductivity decreases the fuel temperature and Doppler constant. As a result, both carbide and nitride cores have favorable passive safety features without additional design fixes that are required in the oxide core concepts. (author)

  17. The Integral Fast Reactor

    International Nuclear Information System (INIS)

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

  18. Investigation of CANDU reactors as a thorium burner

    International Nuclear Information System (INIS)

    Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium can be used as a booster fissile fuel material in the form of mixed ThO2/PuO2 fuel in a CANDU fuel bundle in order to assure reactor criticality. The paper investigates the prospects of exploiting the rich world thorium reserves in CANDU reactors. Two different fuel compositions have been selected for investigations: (1) 96% thoria (ThO2) + 4% PuO2 and (2) 91% ThO2 + 5% UO2 + 4% PuO2. The latter is used for the purpose of denaturing the new 233U fuel with 238U. The behavior of the reactor criticality k ∞ and the burn-up values of the reactor have been pursued by full power operation for >∼8 years. The reactor starts with k ∞ = ∼1.39 and decreases asymptotically to values of k ∞ > 1.06, which is still tolerable and useable in a CANDU reactor. The reactor criticality k ∞ remains nearly constant between the 4th year and the 7th year of plant operation, and then, a slight increase is observed thereafter, along with a continuous depletion of the thorium fuel. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn-up can be achieved with the same fuel (>160,000 MW D/MT). The reactor criticality would be sufficient until a great fraction of the thorium fuel is burned up, provided that the fuel rods could be fabricated to withstand such high burn-up levels. Fuel fabrication costs and nuclear waste mass for final disposal per unit energy could be reduced drastically

  19. Exposure calculation code module for reactor core analysis: BURNER

    International Nuclear Information System (INIS)

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules

  20. A blueprint for GNEP advanced burner reactor startup fuel fabrication facility

    International Nuclear Information System (INIS)

    Research highlights: → This article discusses use of WG-plutonium as the startup fuel for Advanced Burner Reactor. → The presence of gallium in WG fuel may compromise the fuel integrity. → There is no facility exists to remove gallium from plutonium except at laboratory scale. → This article discusses the processes and issues associated with the gallium removal. → The article provides realistic scenario to all stack-holders involved in designing and operating ABR. - Abstract: The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu)-239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the

  1. China experimental fast reactor

    International Nuclear Information System (INIS)

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*1015 n/cm2/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  2. Fast breeder reactor research

    International Nuclear Information System (INIS)

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  3. Fast reactor programme

    International Nuclear Information System (INIS)

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

  4. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

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

  5. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarises the fast reactor research carried out at the Netherlands Energy Research Centre during the year 1981. The neutron and fission product cross sections of various isotopes have been evaluated. In the fuel performance programme, some preliminary results are given and irradiation facilities described. Creep experiments on various stainless steel components are reported

  6. Preliminary safety evaluation of the advanced burner test reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division

    2006-09-15

    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  7. Carbide and nitride fuels for advanced burner reactor

    International Nuclear Information System (INIS)

    Full text: Under the U.S. fast reactor program, reference and alternative 1000 MWth Advanced Burner Reactor (ABR) core concepts were developed using ternary metallic (U-TRU-Zr) and mixed oxide (UO2+TRUO2) fuels. Recently, mixed carbide and nitride fuels have been considered as fast reactor fuels on the basis of their high density, compatibility with coolant, high melting temperature, and excellent thermal conductivity although they are ceramic fuel like a mixed oxide fuel. Thus, the performance of the ABR core loaded with carbide and nitride fuels was evaluated in this study with an expectation that the carbide and nitride fuels can mitigate disadvantages of both metallic and oxide fuels in the ABR: favorable passive safety features in a severe accident compared to the oxide core, a higher discharge burnup compared to the metallic core, and a potential to increase thermal efficiency. All calculations performed in this study were focused on the neutronics characteristics, although the fabrication and irradiation experiences for carbide and nitride fuels are limited and some problems were observed in the reprocessing and irradiation of these fuels. The mixed monocarbide and mixed mononitride fuels were selected as the alternative fuel forms and the ABR core concepts with these fuels were developed based on the reference 1000 MWth ABR core concepts. For consistency, the potential design goals used in the reference ABR core concepts were also employed in this study: a 1000 MWth power rating, medium TRU conversion ratio of ∼0.75, a compact core, one-year operational cycle length at least with a capacity factor of 90%, sufficient shutdown margin with a limited maximum single control assembly fault, and possible use of either metallic or any ceramic fuels in the same core layout. The core layout and outer assembly dimensions of the reference 1000 MWth ABR core were kept, but the intra assembly design parameters were varied to maximize the discharge burnup within the

  8. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  9. Fast reactor database

    International Nuclear Information System (INIS)

    This publication contains detailed data on liquid metal cooled fast reactors (LMFRs), specifically plant parameters and design details. Each LMFR power plant is characterized by about 400 parameters, by design data and by relevant materials. The report provides general and detailed design characteristics including structural materials, data on experimental, demonstration, prototype and commercial size LMFRs. The focus is on practical issues that are useful to engineers, scientists, managers and university students and professors. The report includes updated information contained in IAEA previous publications on LMFR plant parameters: IWGRF/51 (1985) and IWGFR/80 (1991) and reflects experience gained from two consultants meetings held in Vienna (1993,1994). This compilation of data was produced by members of the IAEA International Working Group on Fast Reactors (IWGFR)

  10. Fast breeder reactor

    International Nuclear Information System (INIS)

    This paper outlined the present status of FBR development in six countries and reviewed Japanese activities on FBR development. Joyo experimental FBR has accumulated a lot of technical data including irradiation tests of advanced fuels and was now long shut down due to the partial obstruction of rotating plug movement. Monju prototype FBR reactor experienced a sodium leakage in its secondary heat transfer system during performance tests in December 1995 and had been shut down until May 2010. Feasibility study on commercialized FBR cycle system ended in March 2006 and proposed the concept of commercialized FBR cycle technologies. In order to plan a demonstration reactor, research and development of innovative technologies are conducted as the FaCT (Fast Reactor Cycle Technology Development) Project. In connection with the results of this research and development, a 5-party council of Japan was established to discuss processes of demonstration and commercialization of FBR cycle systems in Japan. Joint efforts were made for a demonstration reactor to be committed in 2015, in addition to start operation around 2025 aiming at the commercialization of FBR before 2050. (T. Tanaka)

  11. The energy analysis of burner reactor power systems

    International Nuclear Information System (INIS)

    Currently most commercial nuclear power stations are based on thermal reactor designs called burner reactors which are net consumers of fissile material. These power stations form one part of a larger system that generates electricity from uranium. However, in addition to producing energy, such systems also consume energy, in the form of various fuels, during construction and operation. This thesis describes the use of energy analysis to determine the total energy required by these systems. A number of factors are shown to influence energy consumption and, in particular, the effect of extracting uranium from different sources is studied in detail. For ores, an important inverse relationship between energy use and ore grade is investigated and quantified. The physical limit at which the energy input to the system is equal to its output is shown to correspond to an average grade of 15 parts per million of 'triuranium octoxide'. Analysis of proposals for extracting uranium from seawater indicates that the only schemes giving a positive energy balance are costly (500 dollars/lb U3O8) and limited to low production rates. The effects of feedback within fuel systems are analysed and the results are used to formulate an economic model in which nuclear electricity prices determine uranium ore costs as well as vice versa. The model demonstrates that, with present techniques, the average economic limit to ore grade is 50 ppm U3O8 with subsequent resources, on current assessment, of only 107 tonnes U3O8. This contradicts most traditional studies which, by assuming fixed, non-dependent fuel costs, suggest an ore grade limit of less than 4 ppm U3O8 and economically recoverable resources in excess of 1010 tonnes U3O8. (author)

  12. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  13. Fast reactor operating experience

    International Nuclear Information System (INIS)

    At the beginning of electricity generation from nuclear power there was the breeder, which fulfilled its duty in a number of smaller test and experimental reactors within national programs. Over the years, some of those reactors have attained impressive availabilities, while others have helped to improve our knowledge by the negative results they contributed. Worldwide a decisive step was taken by the mid- to late sixties in the planning and construction of medium sized demonstration fast breeder power plants (250 to 350 MW). In the Federal Republik of Germany, this step is taken belatedly in building the SNR-300. BN-350 in the USSR, Phenix in France, and PFR in the United Kingdom have now been in operation for some ten years. Over that period, valuable experience has been accumulated in sodium technology. The operating behavior of all components and systems working in sodium is called excellent; the hazards associated with sodium, the fire hazard in particular, thus often seem to be greatly overrated. Leakages have been brought under control. It has always been possible so far to trace them back to systemic faults produced in the welding process. The ability of fast sodium cooled reactors to produce more nuclear fuel than they consume has been demonstrated in Phenix, whose breeding ration has been measured to be 1.16. The first true large breeder, Super Phenix in France, is to be commissioned already in 1985. In building another three breeder power plants the European partners in an association hope to achieve the commercial breakthrough of the breeder line. (orig.)

  14. Fast breeder reactor

    International Nuclear Information System (INIS)

    The fluid-cooled fast breeder reactor described includes an outer cylindrical boundary wall, a plurality of canless fuel elements and breeder material elements received within the boundary wall and being in an array therein forming a fissionable fuel zone and a breeder material zone coaxially surrounding the fissionable fuel zone, a coolant supply system for applying fluid coolant at uniform pressure to the entire cross section within the cylindrical boundary wall, and flow guide devices extending substantially horizontally and disposed at different levels one above the other within the breeder material zone which coaxially surrounds the fissionable fuel zone, means for elastically securing the flow guide devices at alternate levels within the breeder material to the boundary wall, the flow guide devices at the levels intermediate the alternate levels being spaced by an annular gap from the boundary wall. 7 claims, 7 drawing figures

  15. Use of freeze-casting in advanced burner reactor fuel design

    Energy Technology Data Exchange (ETDEWEB)

    Lang, A. L.; Yablinsky, C. A.; Allen, T. R. [Dept. of Engineering Physics, Univ. of Wisconsin Madison, 1500 Engineering Drive, Madison, WI 53711 (United States); Burger, J.; Hunger, P. M.; Wegst, U. G. K. [Thayer School of Engineering, Dartmouth College, 8000 Cummings Hall, Hanover, NH 03755 (United States)

    2012-07-01

    This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models

  16. Use of freeze-casting in advanced burner reactor fuel design

    International Nuclear Information System (INIS)

    This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO2) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary results

  17. United States fast reactor programme

    International Nuclear Information System (INIS)

    The fast reactor programs in USA deal with: EBR-II termination program; fast flux test facility; nuclear energy research initiative; accelerator transmutation of waste; and other non-DOE funded activities. These are concerned with preliminary concept development activities of a secure transportable autonomous reactor -liquid metal cooled

  18. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  19. Sodium fast reactor safety and licensing research plan - Volume II

    International Nuclear Information System (INIS)

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  20. Fast reactor technology development in china status and prospects

    International Nuclear Information System (INIS)

    China has decided to speed-up the nuclear power development. It is programmed that the nuclear power capacity will reach 40 GWe in 2020 and envisaged 60 GWe and 240 GWe in 2030 and 2050 respectively. The basic strategy of PWR-FBR matched development with Fast reactor metal fuel closed cycle for a sustainable and quick increasing nuclear energy supply is adopted. Another strategy also decided is that the partitioning and transmutation of MA will be realized using fast burner and ADS. The fast reactor engineering development will be divided into three steps: China Experimental Fast Reactor (CEFR 65 MWt/20 MWe), 3China Prototype/Demonstration Fast Reactor (CPFR/CDFR ≥1 500 MWt/600 MWe) and China Demonstration Fast Breeder Reactor (CDFBR 1 000-1 500 MWe). The CEFR is under installation and pre-operation testing with it's first criticality planned in 2009. The design study of CPFR is just started in 2006. Recently a discussion for the second step is under way to faster the fast reactor development by a larger than 600 MWe CPFR and as a role of CDFR. (authors)

  1. Fast reactor database. 2006 update

    International Nuclear Information System (INIS)

    Liquid metal cooled fast reactors (LMFRs) have been under development for about 50 years. Ten experimental fast reactors and six prototype and commercial size fast reactor plants have been constructed and operated. In many cases, the overall experience with LMFRs has been rather good, with the reactors themselves and also the various components showing remarkable performances, well in accordance with the design expectations. The fast reactor system has also been shown to have very attractive safety characteristics, resulting to a large extent from the fact that the fast reactor is a low pressure system with large thermal inertia and negative power and temperature coefficients. In addition to the LMFRs that have been constructed and operated, more than ten advanced LMFR projects have been developed, and the latest designs are now close to achieving economic competitivity with other reactor types. In the current world economic climate, the introduction of a new nuclear energy system based on the LMFR may not be considered by utilities as a near future option when compared to other potential power plants. However, there is a strong agreement between experts in the nuclear energy field that, for sustainability reasons, long term development of nuclear power as a part of the world's future energy mix will require the fast reactor technology, and that, given the decline in fast reactor development projects, data retrieval and knowledge preservation efforts in this area are of particular importance. This publication contains detailed design data and main operational data on experimental, prototype, demonstration, and commercial size LMFRs. Each LMFR plant is characterized by about 500 parameters: physics, thermohydraulics, thermomechanics, by design and technical data, and by relevant sketches. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors with complete technical information of a total of 37 LMFR

  2. Fast reactors and nuclear nonproliferation

    International Nuclear Information System (INIS)

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (author)

  3. Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Wegst, Ulrike G.K. [Dartmouth College, Hanover, NH (United States). Thayer School of Engineering; Allen, Todd [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States)

    2014-04-07

    Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

  4. Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

    International Nuclear Information System (INIS)

    Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

  5. Sodium fast reactors (SFRs) and recyclers

    International Nuclear Information System (INIS)

    This presentation is about Sodium Fast Reactor (SFRs) and Recyclers. Their pursuit has been going on in the United States (U.S.) since 1941 and that development work could help support the penetration of SFRs into the current nuclear power market in three forms: 1. A breeding SFR to increase the supply of fissile material. It will not happen for many decades because of increased uranium (U) resources, nuclear market ability to absorb increased U prices, and/or switch to a Thorium (Th) fuel cycle (under development in India) until the anticipated stringent regulations for breeding SFRs are defined and tested. 2. An economic SFR capable of competing with the Advance Light Water Reactor (ALWR) expected to produce electricity in the near future. The Generation IV (Gen IV) program is pursuing that goal under conceptual studies in South Korea (1) and, particularly under the demonstration Japan Sodium Fast Reactor (JSFR) (2) forecasted to start up by 2025 followed by the deployment of commercial JSFRs before 2050. 3. To use the pyro-processing and electro refining methodology developed under the Integral Fast Reactor (IFR) (3) to separate the Light Water Reactor (LWR) spent nuclear fuel (SNF) Transuranics (TRUs) and to burn them in SFRs referred to as Advanced Burner Reactors (ABR). That innovative approach can significantly increase the capacity of geological repositories for disposition of LWR SNF. That last form of SFR is needed urgently to cope with the continued increase in U.S. inventories of recyclable fissile and fertile materials and, particularly, with the projected growth in LWR SNF. According to a recent Electrical Power Research Institute (EPRI) study (4) to reduce CO2 emissions, the U.S. nuclear generated electricity will increase by 64 Gigawatt electrical (GWe) by 2030. While it is realized that additional long term interim storage can alleviate this need, it is not a long term solution because it will have to be followed eventually by final disposal or

  6. Fast reactors using molten chloride salts as fuel

    International Nuclear Information System (INIS)

    This report deals with a rather exotic 'paper reactor', in which the fuel is in the form of molten chlorides. (a) Fast breeder reactor with a mixed fuel cycle of thorium/uranium-233 and uranium 238/plutonium in which all of the plutonium can be burned in situ and in which a denatured mixture of uranium-233 and uranium-238 is used to supply further reactors. The breeding ratio is relatively high, 1.58 and the specific power is 0.75 GW(th)/m3 of core. (b) Fast breeder reactor with two and three zones (internal fertile zone, intermediate fuel zone, external fertile zone) with an extremely high breeding ratio of 1.75 and a specific power of 1.1 GW(th)/m3 of core. (c) Extremely high flux reactor for the transmutation of the fission products: strontium-90 and caesium-137. The efficiency of transmutation is approximately 15 times greater than the spontaneous beta decay. This high flux burner reactor is intended as part of a complex breeder/burner system. (d) Internally cooled fast breeder in which the cooling agent is the molten fertile material, the same as in the blanket zone. This reactor has a moderate breeding ratio of 1.38, a specific power of 0.22 GW(th)/m3 of core and very good inherent safety properties. All of these reactors have the fuel in the form of molten chlorides: PuCl3 as fissile, UCl3 as fertile (if needed) and NaCl as dilutent. (Auth.)

  7. Fast reactors using molten chloride salts as fuel

    International Nuclear Information System (INIS)

    This report deals with a rather exotic ''paper reactor'' in which the fuel is in the form of molten chlorides. (a) Fast breeder reactor with a mixed fuel cycle of thorium/uranium-233 and uranium 238/plutonium in which all of the plutonium can be burned in situ and in which a denatured mixture of uranium-233 and uranium-238 is used to supply further reactors. The breeding ratio is relatively high, 1.58 and the specific power is 0.75 GW(th)/m3 of core. (b) Fast breeder reactor with two and three zones (internal fertile zone, intermediate fuel zone, external fertile zone) with an extremely high breeding ratio of 1.75 and a specific power of 1.1 GW(th)/m3 of core. (c) Extremely high flux reactor for the transmutation of the fission products: strontium-90 and caesium-137. The efficiency of transmutation is approximately 15 times greater than the spontaneous beta decay. This high flux burner reactor is intended as part of a complex breeder/burner system. (d) Internally cooled fast breeder in which the cooling agent is the molten fertile material, the same as in the blanket zone. This reactor has a moderate breeding ratio of 1.38, a specific power of 0.22 GW(th)/m3 of core and very good inherent safety properties. All of these reactors have the fuel in the form of molten chlorides: PuCl3 as fissile, UCl3 as fertile (if needed) and NaCl as dilutent. The fertile material can be 238UCl3 as fertile and NaCl as dilutent. In mixed fuel cycles the 233UCl3 is also a fissile component with 232ThCl4 as the fertile constituent

  8. Fuel cycle of actinide burner-reactor. Review of investigations by > program

    International Nuclear Information System (INIS)

    The problem of long-lived minor-actinides (Np, Am, Cm) transmutation is one of major part of problem of nuclear power ecological safety. The problem of Pu surpluses burning-out adjoins to this problem. Existing and perspective reactor systems could be used for it, but task of optimum organization of the external closed cycle for actinide burner reactor becomes the important aspect of transmutation problem. Since 1992, SSC RIAR has proposed the demonstration program-concept DOVITA (Dry reprocessing, Oxide fuel, Vibropac, Integral, Transmutation of Actinides), which should demonstrate opportunities of new technologies for realization of the optimized fuel cycle for actinide burner reactor. The brief review of study on DOVITA program for 5 years is given in this paper. (J.P.N.)

  9. Fast reactor fuel cycle facility

    International Nuclear Information System (INIS)

    An integrated fuel cycle facility named Fast Reactor Fuel Cycle Facility (FRFCF) is planned to be set up at Kalpakkam to close the fuel cycle of the Prototype Fast Breeder Reactor (PFBR) that is already under construction there. FRFCF is the first project of its kind in India. Closure of fuel cycle of PFBR will be a significant milestone of the second stage of nuclear power programme of the Department of Atomic Energy. The facility would be ready for operation in 2014. Design work and safety review of FRFCF are presently in progress. (author)

  10. Fast reactor designs: Commercial size fast reactors (unforeseen events)

    International Nuclear Information System (INIS)

    This chapter contains detailed design data and main operational data on the following commercial fast reactors (unforeseen events): Super-Phenix-1; Super-Phenix-2; SNR-2; BN-800; DFBR; CDFR; EFR; BN-1600; BN-1800; BREST-1200; JSFR-1500

  11. Fast reactor research in Switzerland

    International Nuclear Information System (INIS)

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

  12. Fast reactor savants take stock

    International Nuclear Information System (INIS)

    Some of the argument grew almost fierce when the Royal Society, one of the world's premier scientific institutions, held an international meeting in London last May on the fast neutron breeder reactor. Discussion skipped between broad scientific principles, technical minutiae, economics and politics. Some impressions are given by an independent writer on energy affairs. (author)

  13. Fast reactor savants take stock

    Energy Technology Data Exchange (ETDEWEB)

    Conway, Arthur

    1989-08-01

    Some of the argument grew almost fierce when the Royal Society, one of the world's premier scientific institutions, held an international meeting in London last May on the fast neutron breeder reactor. Discussion skipped between broad scientific principles, technical minutiae, economics and politics. Some impressions are given by an independent writer on energy affairs. (author).

  14. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  15. Status of national programmes on fast reactors

    International Nuclear Information System (INIS)

    Based on the International Working Group on Fast reactors (IWGFR) members' request, the IAEA organized a special meeting on Fast Reactor Development and the Role of the IAEA in May 1993. The purpose of the meeting was to review and discuss the status and recent development, to present major changes in fast reactor programmes and to recommend future activities on fast reactors. The IWGFR took note that in some Member States large prototypes have been built or are under construction. However, some countries, due to their current budget constraints, have reduced the level of funding for research and development programmes on fast reactors. The IWGFR noted that in this situation the international exchange of information and cooperation on the development of fast reactors is highly desirable and stressed the importance of the IAEA's programme on fast reactors. These proceedings contain important and useful information on national programmes and new developments in sodium cooled fast reactors in Member States. Refs, figs and tabs

  16. Utilization of fast reactor excess neutrons for burning minor actinides and long lived FPs

    International Nuclear Information System (INIS)

    An evaluation is made on a large MOX fuel fast reactor's capability of burning minor actinides and long lived fission products (FPs) without imposing penalties on core nuclear and safety characteristics. The excess neutrons generated in the fast reactor core are fully utilized not only to generate the fissile material but also to transmute the minor actinides and long lived FPs. The FP target assemblies which consist of Tc-99 and I-129 are loaded into the selected blanket positions whereas the minor actinides are loaded to the rest of the blanket. A long term FP accumulation scenario is also considered in the mix of FP burner fast reactor and non-burner LWRs. (author)

  17. Integral fast reactor safety features

    International Nuclear Information System (INIS)

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

  18. On fast reactor kinetics studies

    Energy Technology Data Exchange (ETDEWEB)

    Seleznev, E. F.; Belov, A. A. [Nuclear Safety Inst. of the Russian Academy of Sciences IBRAE (Russian Federation); Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F. [Inst. for Physics and Power Engineering IPPE (Russian Federation)

    2012-07-01

    The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)

  19. 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,

  20. Heterogeneous Recycling in Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  1. A review of fast reactor activities in Italy

    International Nuclear Information System (INIS)

    In this paper, Italian activities on liquid metal fast reactors are shown for the period May 1995 - April 1996. During this period the ENEA collaboration with General Electric on ALMR came to an end as a consequence of the reduced effort on the design development. Nevertheless ENEA completed the studies on the PRISM Mod B oxide burner core, the neutronic configuration of which was presented at last year's meeting. Some results of the dynamic calculations are shown. ENEA participated in the IAEA/EC benchmark on the comparative calculations for severe accident in BN-800 reactor. A complete neutronic evaluation has been made including power distribution, Doppler, sodium void and material coefficients. Activities on seismic isolation are also outlined. The Italian contribution to SPX restart and operation is described; some information about the complementary convention of the Nersa society is given, together with the Italian industry participation in the SPX restart. (author)

  2. Innovative Fast Reactors: impact of fuel composition on reactivity coefficients

    International Nuclear Information System (INIS)

    Innovative fast reactors will have to comply with the major objectives as stated, e.g. within the Generation-IV initiative, namely sustainability and waste minimization, keeping their safety characteristics at least as good as Generation-III evolutionary reactors. As a consequence, a specific feature of most innovative fuels envisaged for future fast reactors is the potential high content of Minor Actinides (MA: Np, Am, Cm etc). This is the case whatever the mission assigned to a fast reactor: to be a breeder, an 'isogenerator' or a TRU burner, i.e. fast reactors with very different conversion ratios (CRs), and whatever the strategy adopted for the spent fuel processing. The MA content can vary in a wide range, according to the core design characteristics and the fuel type. In practice, the MA content can vary from a few percent (e.g. <10%) in a core with a CR ≥ 1 and with the homogeneous recycle of not-separated TRU, aiming to a MA inventory stabilisation in the fuel cycle, to ∼20-30% in the case of the fuel of a burner fast reactor (CR<1)with a Pu/MA ratio ∼1. It is well known that the introduction of such significant amounts of MA can have an impact on the core reactivity coefficients (e.g. coolant void and Doppler coefficients) and, in the case of cores with a very low CR value, on the delayed neutron fraction. Several recent studies have been devoted to quantify these effects. However, it is important to use a flexible and powerful physics tool, in order to understand the physics features of the calculated effects, to predict trends and potential limitations during preliminary design phases. This flexible and powerful physics tool is provided by the Equivalent Perturbation Theory (EGPT), as implemented in the ERANOS code system. The full paper will give a detailed analysis of these phenomena for both Na-void and Doppler reactivity coefficients in a variety of different FRs with different core characteristics. It will also be shown how to apply this type

  3. Fundamental experiments for FBX burner linear fusion reactor core with FBX plasma flow

    International Nuclear Information System (INIS)

    FBX is a production and confinement scheme of a spherical torus that carries a strong plasma current with both toroidal and poloidal components. On the other hand HI-I is a fundamental experiment on moving plasma. With two of them, a new type of fusion reactor scheme FBX-III BURNER (III) is established. In this paper, the fundamental results of the first two types of experiments are introduced to find out a total plasma behavior of the long term project. 9 refs., 5 figs

  4. A review of the UK fast reactor programme. March 1977

    International Nuclear Information System (INIS)

    This paper reports on the Fast Reactor Programme of United Kingdom. These are the main lines: Dounreay Fast Reactor; Prototype Fast Reactor; Commercial Fast Reactor; engineering development; materials development; chemical engineering/sodium technology; fast reactor fuel; fuel cycle; safety; reactor performance study

  5. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  6. Interfacial effects in fast reactors

    International Nuclear Information System (INIS)

    The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed to measure U-238 capture rates near th blanket--reflector interface in the MIT Blanket Test Facility. Prior MIT experiments on a thorium--uranium interface in a blanket assembly were also reanalyzed. Extremely localized fertile capture rate increases of on the order of 50% were measured immediately at the interfaces relative to extrapolation of asymptotic interior traverses, and relative to state-of-the-art (LIB-IV, SPHINX, ANISN/2DB) calculations which employ infinite-medium self-shielding throughout a given zone. A method was developed to compute a spatially varying background scattering cross section per absorber nucleus which takes into account both homogeneous and heterogeneous effects on the interface flux transient

  7. Molten salt related extensions of the SIMMER-III code and its application for a burner reactor

    International Nuclear Information System (INIS)

    Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16-20 November 2003]. The molten salt fuel is a ternary NaF-LiF-BeF2 system fuelled with ca. 1 mol% typical compositions of transuranium-trifluorides (PuF3, etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP' 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as to

  8. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    International Nuclear Information System (INIS)

    Highlights: • Thorium as support fertile material for TRU transmutation in Fast Reactors. • Comparative analysis of Th and U based breakeven and burner Fast Reactors. • Thorium fosters significant advantages in terms of safety parameters. • Inherent safety is investigated through quasi-static reactivity and energy balances. • Th use in low-CR Fast Reactors does not reduce fuel decay heat and neutron sources. - Abstract: The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232

  9. Aspects of the fast reactors fuel cycle

    International Nuclear Information System (INIS)

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

  10. Status of fast reactor research in Germany

    International Nuclear Information System (INIS)

    The paper gives a short survey of fast reactor activities in Germany. The fast reactor activities of FZK are part of the Nuclear Safety Projects. The R and D program include neutron physical and safety calculated, and post-irradiated examination of structural materials. The key issues and tasks of the program concerned safety and transmission of minor activities and fission products. (author)

  11. Startup operational tests of fast reactors

    International Nuclear Information System (INIS)

    This paper is mainly concerned with the experiences of the two main phases of startup operational tests of fast reactors: (1) The general tests and Sodium filling before core loading. (2) The core loading,approach to criticality and power build up operational tests, taking for example a large and middle demonstrating integrated-type fast reactor. (author)

  12. Improved structural materials for fast breeder reactors

    International Nuclear Information System (INIS)

    Electricity plays a crucial role in the economic development of our country. Coal is the primary fuel for generation of electricity in India as in many other countries. In India, generation of power by nuclear reactors is very important because of (i) availability of large thorium resource, (ii) constraints on setting up of fossil fuel based power plants and (iii) the negligibly small green house gas emissions by nuclear energy. The nuclear programme of the country is being implemented in three stages: (i) pressurized heavy water reactors of the CANDU type, (ii) sodium-cooled fast reactors and (iii) thorium-based reactors. Sodium-cooled fast reactor (SFR) technology is envisioned to make use of the large thorium reserves available. India has undertaken and made rapid strides in developing SFR technology and building of fast reactors for energy generation. A Fast Breeder Test Reactor (FBTR) of 40 MWt is operating successfully for over 25 years at Indira Gandhi Centre for Atomic Research. Based on the design, construction and operational experience, a 500 MWe Prototype Fast Breeder Reactor (PFBR) has been designed indigenously and is in an advanced stage of construction. Its design is being further optimised for enhanced economy with respect to cost of electricity production, for use in commercial reactors. Currently, several R and D programmes are under implementation for the development of new materials required for improved economy of commercial fast reactors

  13. Fast reactor strategy in European Union

    International Nuclear Information System (INIS)

    The tendency and strategy of fast reactors development in European Union are considered. The advantages and disadvantages of sodium, lead-bismuth and gas cooled fast reactors are discussed. It is shown that development of such reactors is the further sustainable development of nuclear power engineering. All three tendencies have clear structure and tasks, all prototypes will appear by 2020 and NPP - towards the middle of the century. It is pointed out that sodium coolant is the leading tendency in fast reactor development in European Union

  14. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  15. ELSY - The European Lead Fast Reactor

    OpenAIRE

    Alemberti, Alessandro; Carlsson, Johan; Malambu, Edouard; ORDEN Alfredo; CINOTTI Luciano; STRUWE Dankward; Agostini, Pietro; Monti, Stefano

    2009-01-01

    The European Lead Fast Reactor is being developed starting from September 2006, in the frame of the ELSY (European Lead SYstem) project funded by the Sixth Framework Programme of EURATOM. The project, coordinated by Ansaldo Nucleare, involves a wide consortium of European organizations. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered te...

  16. Status of fast reactor activities in Russia

    International Nuclear Information System (INIS)

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

  17. A CFD Simulation Process for Fast Reactor Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Kurt D. Hamman; Ray A. Berry

    2010-09-01

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k–e and SST (Menter) k–? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  18. A CFD simulation process for fast reactor fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hamman, Kurt D., E-mail: Kurt.Hamman@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Berry, Ray A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States)

    2010-09-15

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly 'benchmark' geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-{epsilon} and SST (Menter) k-{omega} were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  19. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

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

  20. Sodium fast reactors with closed fuel cycle

    CERN Document Server

    Raj, Baldev; Vasudeva Rao, PR 0

    2015-01-01

    Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

  1. New fast-reactor approach

    International Nuclear Information System (INIS)

    The design parameters for a 1000 MW LMFBR type reactor are presented. The design requires the multiple primary coolant pumps and heat exchangers to be located around the core within the reactor vessel

  2. Decommissioning of fast reactors after sodium draining

    International Nuclear Information System (INIS)

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

  3. Fast molten salt reactor-transmuter for closing nuclear fuel cycle on minor actinides

    International Nuclear Information System (INIS)

    Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle is the most perspective and actual direction. The reactor on melts salts - molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed nuclear fuel cycle, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides in molten salt reactor is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF - BeF2; NaF - LiF - BeF2; NaF-LiF ; NaF-ZrF4 ; LiF-NaF -KF; NaCl. RRC 'Kurchatov institute' together with other employees have developed the basic design reactor installations with molten salt reactor - burner long-living nuclides for fluoride fuel composition with the limited solubility minor actinides (MAF3 10 mol %) allows to develop in some times more effective molten salt reactor with fast neutron spectrum - burner/ transmuter of the long-living radioactive waste. In high-flux fast reactors on melts salts within a year it is possible to burn ∼300 kg minor actinides per 1 GW thermal power of reactor. The technical and economic estimation given power

  4. The status and prospects of fast reactor technology development in china

    International Nuclear Information System (INIS)

    China has decided to speed-up the nuclear power development. It is programmed that the nuclear power capacity will reach 40 GWe in 2020 and envisaged 60 GWe and 240 GWe in 2030 and 2050 respectively. The basic strategy of PWR-FBR matched development with fast reactor metal fuel closed cycle for a sustainable and quick increasing nuclear energy supply is adopted. Another strategy also decided is that the partitioning and trans-mutation of MA will be realized using fast burner and ADS. The fast reactor engineering development will be di-vided into three steps: China Experimental fast Reactor (CEFR 65 MWt/20 MWe), China Prototype/Demonstration Fast Reactor (CPFR/CDFR≥1 500 MWt/600 MWe) and China Demonstration Fast Breeder Reactor(CDFBR100-1500 MWe). The CEFR is under installation and pre-operation testing with its first criticality panned in 2009. The design study of CPFR is just started in 2006. Recently a discussion for the second step is under way to faster the fast reactor development by a larger than 600 MWe CPFR and as a role of CDFR. (authors)

  5. Liquid Metal Coolant Technology for Fast Reactors

    International Nuclear Information System (INIS)

    In the paper presented are results of comparative analysis and the choice of liquid metal coolants for fast reactors, the current status of studies on the physical chemistry and technology of sodium coolants for fast neutron reactors and heavy liquid metal coolants, namely, lead-bismuth and lead for fast reactors and accelerator driven systems. There are descriptions of devices designed for control of the impurities in sodium coolants and their removal as well as methods of heavy liquid metal coolant quality control, removal of impurities from heavy liquid metal coolants and the steel surface of components of nuclear power plants (NPPs) and relevant equipment. Attention is given to the issues of modelling of impurity mass transfer in liquid metal coolants and designing new liquid metal coolants for NPPs. Results of the analysis of NPP abnormal operating conditions are presented. The adopted design approaches assure reliable protection against accidents. Up to now, about 200 reactor-years of sodium cooled fast reactor operation and about 80 reactor-years of submarine reactor operation have been gained. The new goals for sodium and heavy liquid metal coolant technology have been formulated as applied to the new generation fast reactors. (author)

  6. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)

    2015-11-04

    This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and

  7. Review of fast reactor activities in India

    International Nuclear Information System (INIS)

    It may be recalled that In the presentation at the last meeting of the IWGFR (13th Annual meeting), a broad outline of India's nuclear energy programme and the role of fast breeders in the programme has been provided. The steps taken to enable the fast breeders to fulfil their role have also been described. In brief, fast breeder reactors are considered as an essential and integral part of the programme of nuclear energy and constitute the second step in the programme, the first being the construction of natural uranium heavy water moderated reactors which will consume natural uranium but will produce plutonium to fuel fast breeder reactors. This basic position has remained unchanged and the Government is now taking steps to build a large number of heavy water reactors, say 10 million Kw capacity in the next 20 years. This defines the time frame for developing the fast breeder technology in the country. It has therefore been decided to mobilise the efforts towards design, construction and operation of a medium sized (about 500 M We) reactor by mid-nineties. Thus, the climate for fast breeder reactors is good and there is a good deal of enthusiasm amongst the scientists and engineers working in the field although the actual implementation of the programme during the year had to face certain difficulties

  8. Development in UK commercial fast reactor design

    International Nuclear Information System (INIS)

    The design of the CDFR commercial demonstration fast reactor which should be put into operation early in the 90-ties is described. Basic elements of the reactor components are considered. The choice of the integrated primary coolant circuit, and design of intermediate heat exchangers, sodium pumps and charging machines is substantiated. The reactor power is 1320 MW(e), or 3300 MW(t). The sodium temperature at the reactor inlet is 370 deg C, at its outlet 540 deg C. Linear loading per fuel element length is 40 W/mm. The conclusion is drawn that the described design of the demonstration reactor fully corresonds to requirements of a full-scale commercial NPP with a fast reactor

  9. A glossary of terms for fast reactors

    International Nuclear Information System (INIS)

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

  10. The fast breeder reactor Rapsodie (1962)

    International Nuclear Information System (INIS)

    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)

  11. Fast reactor research activities in Brazil

    International Nuclear Information System (INIS)

    Fast reactor activities in Brazil have the objective of establishing a consistent knowledge basis which can serve as a support for a future transitions to the activities more directly related to design, construction and operation of an experimental fast reactor, although its materialization is still far from being decided. Due to the present economic difficulties and uncertainties, the program is modest and all efforts have been directed towards its consolidation, based on the understanding that this class of reactors will play an important role in the future and Brazil needs to be minimally prepared. The text describes the present status of those activities, emphasizing the main progress made in 1996. (author)

  12. Stationary Liquid Fuel Fast Reactor

    International Nuclear Information System (INIS)

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  13. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  14. EC-FP7 ARCAS: technical and economical comparison of Fast Reactors and Accelerator Driven Systems for transmutation of Minor Actinides

    International Nuclear Information System (INIS)

    The ARCAS project aims to compare, on a technological and economical basis, Accelerator Driven Systems and Fast Reactors as Minor Actinide burners. It is split in five work packages: the reference scenario definition, the fast reactor system definition, the accelerator driven system definition, the fuel reprocessing and fabrication facilities definition and the economical comparison. This paper summarizes the status of the project and its five work packages. (author)

  15. Thermomechanical analysis of fast-burst reactors

    International Nuclear Information System (INIS)

    Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor

  16. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

  17. JNC viewpoint on fast reactor knowledge preservation

    International Nuclear Information System (INIS)

    JNC is undertaking a major program of research and development on liquid-metal cooled fast breeder reactors, which is fully supported by the government of Japan and the electrical utilities. Hence, the perspective of JNC on knowledge preservation is rather different from that of organizations where the fast reactor project has been scaled down or discontinued. Within JNC, there is a statutory obligation to preserve documentary records of the fast reactor project. Over time the method of archiving has changed from optical (microfilm, microfiche etc.) to digital storage. It is the long-term objective of JNC to convert all its records to digital format and make them available to staff over its intranet. JNC is also attempting to preserve 'human knowledge', that is, the expertise of staff who have been involved in the fast reactor project over a long period and who are now nearing retirement. Based on this information, two computerized systems are currently being constructed: one which records in,a readily accessible manner the background to key design decisions for the Monju plant; and a second which uses simple relationships between design parameters to aid designers understand the knock-on effects of design choices (joint project with Mitsubishi). To its partners in international cooperation - the US/DoE and the organizations of the Euro-Japan collaboration - JNC is proposing a joint approach to knowledge preservation and retrieval. The proposed concept, dubbed the International Super-Archive Network (ISAN), would make use of the standardized software the new technologies of the internet increase the mutual accessibility of fast reactor information. JNC considers it extremely important to reflect the lessons learnt from previous experience in the fast reactor field to the operation and maintenance of Monju and the design of future reactors. (author)

  18. Spatial Kinetics in Fast Reactors

    International Nuclear Information System (INIS)

    Reactor neutronic calculations designed for calculating of unsteady processes in a real 3D geometry require processing of a large amount of information. They cannot consist of simple models, as they should reflect the processes of variations of all local reactor characteristics. The model complexity and the significant time needed for numerical solution of neutron-transport equations limit the choice of methods that can achieve the required accuracy. Thus there is an urgent need for the development of various methods enabling the solution of unsteady neutron-transport equations and estimates of their errors, spent time and consistency with the experimental data. (author)

  19. Design challenges for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    It is of vital importance for commercialized fast reactor to achieve component design with excellent integrity and economics. In the phase II of feasibility study till 2005, a system design for commercialized fast reactor for sodium cooling was achieved. For economical improvement, the system design was undertaken along the guideline including innovative technology for system simplification and new material development. In this paper, the results from the design for shortening of cooling pipings, new components and three dimensional seismic isolation are described, which are design challenges for the sodium cooled fast reactor. Furthermore, in-service inspection and repair is mentioned. Finally, economics for the simplification and the mass reduction employing above technologies are examined

  20. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

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

  1. Multi-group calculations for fast reactors

    International Nuclear Information System (INIS)

    The paper deals with various causes of error in calculations. The first part sets out the mathematical approximations (diffusion approximation, Sn method, etc.), the numerical resolution methods (effect of integration step), the models used, and the implications of these various factors in the determination of the principal characteristics of a fast neutron reactor. The second part studies the effect on reactivity of variations of element cross-sections, using various fuels, in a reactor of rather hard spectrum. (author)

  2. Removing the heat from fast reactor cores

    International Nuclear Information System (INIS)

    Whatever the view about the time when fast breeder reactors will reach the commercial and industrial stage, there is a growing and widespread interest in developing their technology. The reactors are called breeders because they can produce more fissile material than they use in their own cores. As part of an Agency programme related to their technology and economics a symposium on Alkali Metal Coolants - Corrosion Studies and System Operating Experience was held in Vienna from 28 November to 2 December

  3. The fast breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    This paper outlines the current national fast reactor program in France and U.K and describes the increasing plant operational experience being acquired in the two countries for fuel reprocessing and the European project of a series of demonstration reprocessing plants of sufficient capacity to serve the needs of several commercially sized fast reactors. The key futures of France and U.K. programs are: fuel dismantling and pin cropping, dissolution, fuel dissolvers, liquor clarification, plutonium accountancy, solvent extraction, product preparation and packaging, wastes and emissions and fuel fabrication (initial blending, milling, pellet pressing, etc...)

  4. Fast Reactor Development Strategy in China

    International Nuclear Information System (INIS)

    • China devotes herself to the peaceful use of nuclear to meet the growing energy demand. Proper amount of nuclear power plants could provide clean energy with low risk. • Fast reactor is a promising technology to ensure the sustainable development of nuclear energy, which can produce new fuel from depleted uranium and burn the long-life radioactive waste at the same time. It is expected that fast reactor will provide enough clean power to people for a long term in the future

  5. Fast Reactor Knowledge Management at IGCAR, India

    International Nuclear Information System (INIS)

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

  6. Fast-acting nuclear reactor control device

    International Nuclear Information System (INIS)

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

  7. The fast breeder reactor. v. 1

    International Nuclear Information System (INIS)

    The Energy Committee's report was prepared after hearing evidence (the minutes of which are published in Volume II) from the Central Electricity Generating Board, the United Kingdom Atomic Energy Authority and the Department of Energy. Memoranda received from other interested bodies or individuals were also considered and members of the Committee visited fast breeder projects in France, West Germany and Japan. As well as the development of the fast reactors, the economics and timescale were reviewed. The particular case of the fast breeder reactor and proposed fuel reprocessing plant at Dounreay was considered. The main conclusion is that major expenditure on fast reactor programmes can only be justified if there is a potential economic case, i.e. if the fuel cycle costs are lower than for PWRs. This would only be the case if uranium costs increased greatly. It is not considered worthwhile to participate in the European Fast Reactor although this should be reviewed in 1993 and 1997. The Committee agree with the Government's decision to cease funding the PFR in 1994 and endorses the need to regenerate the local economy which will be affected by this decision. (UK)

  8. TITAN program and direct cycle fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Yasuyoshi; Yoshizawa, Yoshio; Nitawaki, Takeshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

    2000-07-01

    In December 1999, the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (TIT) started a new program for the development of advanced nuclear reactors with small and medium size. TITAN is the acronym for the program. A novel concept of a carbon dioxide cooled direct cycle fast reactor with a Rankin cycle has been proposed as the advanced nuclear reactors and evaluated for an alternative option to liquid metal cooled fast reactors (LMFRs). The use of carbon dioxide as coolant eliminates major safety related problems of sodium cooled fast reactors: positive sodium void reactivity, hazardous reaction between sodium and water or air. The decay heat is passively removed by allocating a storage tank of liquidized carbon dioxide between the regenerator and the condenser, and by introducing naturally the carbon dioxide vaporized from the tank into the core in the event of the depressurization accident. The direct cycle results in considerable simplification of the heat transport system owing to the absence of intermediate cooling and water-steam loops comparing with the LMFRs. The thermal efficiency of the direct cycle is evaluated as 34.3 %, which is slightly higher than those in the current BWRs and PWRs. (author)

  9. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    A review of the United Kingdom Fast Reactor Programme is introduced. Operational experience with the Prototype Fast Reactor (PFR) is briefly summarized. The design concept of the Commercial Demonstration Fast Reactor (CDFR) is given in some detail. The emphasis is on materials development, chemical engineering/sodium technology, fuel reprocessing and fuel cycle, engineering component development and reactor safety

  10. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    International Nuclear Information System (INIS)

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

  11. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  12. Safety Aspects of Thorium Fuel in Sodium-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    While detrimental to the breeding performance, use of thorium (Th) in Fast Reactors (FRs) has certain positive aspects: in principle, it has better thermal properties and irradiation performance than uranium (U)-based fuels, and it increases the potential burning rate of legacy transuranic waste for a burner design. The impact of thorium on the top-level operational and safety characteristics of a sodium-cooled transuranic-burner FR design is the focus of this paper. Different fuel cycle burning schemes are considered, including the development of a breakeven design to cover the long-term transition to a self-sufficient cycle. For completeness, a comparison with the counterpart U-based cores is also given. Reactivity feedback coefficients and other parameters are employed to predict the impact on the reactor passive-safety features of the various fuel cycle options. Reactivity decomposition techniques are employed to show the key contributors to the reactivity effects of the different fuels. The results show significant safety improvements fostered by implementing thorium fuel, with notable reductions in the reactivity insertion in case of core voiding, ~2$ for the burner designs and ~6$ for the breakeven design (which shows overall negative voids), while preserving a relatively simple core configuration. (author)

  13. Liquid metal fast reactor transient design

    International Nuclear Information System (INIS)

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

  14. Fast Reactor and ADS development in China

    International Nuclear Information System (INIS)

    Conclusion: • The Fukushima accident influence China deeply. “The 12th five years plan and 2020 perspective goal of nuclear safety and radioactive pollution prevention” has been approved which means the nuclear may restart in the near future. • A demonstration fast reactor is under design. • More and more research works will be executed on CEFR

  15. Fast neutron flux in heavy water reactors

    International Nuclear Information System (INIS)

    The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author)

  16. Fast reactor systems for deep sea research

    International Nuclear Information System (INIS)

    Fast reactor (FR) systems have been studied as power units for unmanned bases and research submersibles to monitor various phenomena and as a thermal source for the unmanned base to feed useful microorganisms in the deep sea region. The systems, which are set in pressure hulls, comprise of the FR's and secondary gas loops. Concepts and arrangements of the systems are presented. (author)

  17. Thermophysical properties of fast reactor fuel

    International Nuclear Information System (INIS)

    This paper identifies the fuel properties for which more data are needed for fast-reactor safety analysis. In addition, a brief review is given of current research on the vapor pressure over liquid UO2 and (U,PU)O/sub 2-x/, the solid-solid phase transition in actinide oxides, and the thermal conductivity of molten urania

  18. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

    The invention, in which the United States Department of Energy has participated as co-inventor, relates to breeder reactor fuel elements, and specifically to such elements incorporating 'getters', hereafter designated as fission product traps. The main object of the invention is the construction of a fast breeder reactor fuel pin, free from local stresses induced in the cladding by reactions with cesium. According to the invention, the fast breeder fuel element includes a cladding tube, sealed at both ends by a plug, and containing a fissile stack and a fertile stack, characterized by the interposition of a cesium trap between the fissile and fertile stacks. The trap is effective at reactor operating temperatures in retaining and separating the cesium generated in the fissile material and preventing cesium reaction with the fertile stack. Depending on the construction method adopted, the trap may consists of a low density titanium oxide or niobium oxide pellet

  19. Development of plutonium: Fast Neutrons Reactors option

    International Nuclear Information System (INIS)

    Phenix reactor is shortly described with combustible assembly with some operational data. 'CAPRA'(Plutonium Enhance Consumption in Fast Reactors) is an R and D program for the development of an optimized combustible for fast reactors for burning more plutonium. Three ways are tested: a 45% Pu concentration in an oxide fuel keeping actual fabrication and reprocessing options giving a 80 kg/TWh Pu consumption, a fuel without U238 but with a W or a Mo matrix with problems of reprocessing and core reactivity giving a 110 kg/TWh Pu consumption, and a nitride fuel with an up to 65% Pu concentration giving a 90 to 100 kg/TWh Pu consumption. (A.B.)

  20. The development of fast reactors in France

    International Nuclear Information System (INIS)

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

  1. Fast Reactor Fuel Development in Europe

    International Nuclear Information System (INIS)

    Research and development of minor-actinide-bearing fuels in Europe has made significant progress, with a number of scoping irradiation tests made on a number of candidate fuels foreseen for fast reactors and dedicated minor actinide transmutation systems, e.g. the accelerator driven system. Currently, efforts concentrate on uranium based fuels, as the deployment of fast reactor fleets requires Pu generation in order to achieve sustainability. Both homogeneous and heterogeneous concepts for minor actinide reactor recycling are considered. In the former, the minor actinides are added in small quantities to the mixed oxide fuel, while in the latter, the minor actinides are loaded in significant quantities in UO2. Irradiation programmes to test these concepts for pellet and SPHEREPAC fuel configurations are under way. (author)

  2. Status of fast reactor activities in Brazil

    International Nuclear Information System (INIS)

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

  3. Bowing and interaction of fast reactor subassemblies

    International Nuclear Information System (INIS)

    Deformations of the subassembly structural components, in particular the bowing of the hexagonal wrapper which encloses the pin bundles, due to stainless steel swelling as a result of fast neutron irradiation give rise to operational and safety problems especially in large breeder reactors where the neutron flux is much larger than in smaller reactors. The restraint on bowing induces heavy restraint loads and high stresses in the wrapper, which tend to limit the target burn-up of the fast reactor fuels. Therefore, a realistic analysis has to include the phenomenon of creep to determine the extent to which the stresses in the wrapper would be relaxed due to both thermally induced and irradiation induced creep. Apart from this, determination of deformations of the subassemblies in the core due to the interaction among them is also necessary. (author)

  4. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The existing sodium cooled fast reactors (SFR) have two types of designs--loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphenix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL's Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed

  5. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Young; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of)

    2015-05-15

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity.

  6. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity

  7. Nuclear data needs for fast reactors

    International Nuclear Information System (INIS)

    The nuclear data, i.e., the numerical information about every nuclide - especially those representing the probabilities of various nuclear interactions and of radioactivity - of interest in a nuclear fission reactor are among the most essential inputs to be known a priori, to the best possible accuracy, for the design of nuclear reactor. The nuclides of interest cover not just (1) the fuel nuclides, the containers, the coolant, the moderator (if any), etc., that are initially inserted, but also (2) the actinides, the fission products, etc. that would be produced from the moment the reactor goes into operation and (3) the decay products that are produced even while the reactor is shutdown. The nuclide-list is known to cover a few hundreds. The neutron-nuclear interaction cross-section data, required for a few tens of reactions, very sensitively depend on the nuclide species and the neutron energy. Hence the data requirement significantly varies between thermal and fast reactors. The present talk is intended to touch upon the kinds and forms of nuclear data needed in the design and analysis of fast reactors. The recent variants available in the databases and some inter-comparison results will also be presented. (author)

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

    International Nuclear Information System (INIS)

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

  9. Advanced simulation for fast reactor design

    International Nuclear Information System (INIS)

    Full text: This talk broadly reviews recent research aimed at applying advanced simulation techniques specifically to fast neutron reactors. By advanced simulation we generally refer to attempts to do more science-based simulation - that is, to numerically solve the three-dimensional governing physical equations on fine scales and observe and study the holistic phenomena that emerge. In this way simulation is treated more akin to a traditional physical experiment, and can can be used both separately and in conjunction with physical experiments to develop more accurate predictive theories on reactor behavior. Many existing fast reactor modeling tools were developed for last generation's computational resources. They were built by engineers and physicists with deep physical insight - insight that both shaped and was informed by existing theory, and was underpinned by a vast repository of experimental data. Their general approach was to develop models that were tailored to varying degrees to the details of the reactor design, using free model parameters that were subsequently calibrated to match existing experimental data. The resulting codes were thus extremely useful for their specific purpose but highly limited in their predictive capability (neutronics to a lesser degree). They tended to represent more the state-of-the-art in our understanding rather than tools of exploration and innovation. Recently, a number of researchers have attempted to study the feasibility of solving more fundamental governing equations on realistic, three-dimensional geometries for different fast reactor sub-domains. This includes solving the Navier-Stokes equations for single-phase sodium flow (Direct Numerical Simulation, Large Eddie Simulation, and Reynolds Averaged Navier Stokes Equations) in the core, upper plenum, primary and intermediate loop, etc.; the non-homogenized transport equations at very fine group, angle, and energy discretization, and thermo-mechanical feedback based on

  10. Advanced sodium fast reactor unit concept

    International Nuclear Information System (INIS)

    The paper presents status of development for 1200 MW power unit with sodium fast reactor for commercial construction in the Russian Federation. General characteristics of the reactor plant (RP) and power unit as well as goals that shall be achieved because of design development are described. The power unit design is based on technical decisions, which have been partially proven during sodium reactor operation in Russia and partially have been validated by R and D work for BN-800 RP. At the same time, new technical decisions are applied that improve safety and technical-and-economic indices. To validate them, the corresponding R and D work shall be performed. It is planned to construct the pilot power unit in 2020 and to put into operation the next commercial power units of this type using plutonium generated in the thermal reactors. (author)

  11. Homogeneous recycling of minor actinides in an EFR type fast reactor

    International Nuclear Information System (INIS)

    The capability of the European Fast Reactor (EFR) as a typical large fast reactor to incinerate the minor actinides Np and Am, produced in LWRs, has been investigated for the case of homogeneous recycling of Np, Am and Pu. Detailed account is taken for a three-region reload scheme in which every two years 1/3 of the core is refuelled. The fraction of admixed minor actinides is varied from 2.8 to 7.5%. Results are given in terms of the number of clean-up LWRs by one burner, the nuclide inventories during recycling, and the risk potential of the waste from the incineration system compared to the non-incineration case. It is concluded that an enforced research should be done in chemistry to prove that a satisfactory separation especially of Am and also of Cm from the rare earths in the waste is possible on large scale. (author). 4 refs, 7 figs, 1 tab

  12. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  13. Review of the Italian fast reactor programme

    International Nuclear Information System (INIS)

    The Caorso power station (860 MWe) underwent its third reloading at the end of 1985. The construction work on the power station at Montalto di Castro, a twin 1,000 MWe BWR reactor plant has continued according to plan. ENEL has agreed with Ansaldo the contract for the NSSS of the Trino Vercellese nuclear plant. This will be the first of the so-called ''Unified Design Nuclear Power Station'', twin PWR units, Westinghouse type, 950 MWe each, with 3 cooling loops. The value of the NSSS order is 1400 billion lire, while the complete cost will be 5000 billion lire. Civil engineering work will begin in July 1987 and completion is planned for 1995. In the fast reactor field, the Italian effort has been operating in the framework of the European R and D agreement; during 1985 the sum assigned by ENEA to fast reactors, excluding PEC realization, was about 100 billion lire. Worthy of note was the signing of a formal agreement between ENEA and ENEL with the aim of coordinating their activities on fast reactor development

  14. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  15. The different facilities of the reactor PHENIX for radio isotope production and fission product burner

    International Nuclear Information System (INIS)

    During the last few years different tests have been made to optimize the blanket of the reactor. Year after year the breeding ratio has lost a part of interest regarding the production and availability of plutonium in the world. A characteristic of a fast reactor is to have important neutron leaks from the core. The spectrum of those neutrons is intermediate, the idea was to find a moderator compatible with sodium and stable in temperature. After different tests we kept as a moderator the calcium hydride and as a samply support, a cluster which is separated from the carrier. At the end we present the model used for thermalized calculations. The scheme is then applied to a heavy nuclide transmutation example (Np237 Pu238) and to fission product transmutation (Tc99). (author)

  16. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

    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

  17. Fuel management codes for fast reactors

    International Nuclear Information System (INIS)

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

  18. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    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)

  19. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    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

  20. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    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

  1. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  2. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward

    International Nuclear Information System (INIS)

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high

  3. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

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

  4. Review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    This report covers the activities of the experimental fast reactor JOYO from April 1985 to March 1986. After completion of the 7th duty cycle operation at the end of March 1985, special operation was carried out for the in-vessel performance test of the failed fuel detection and location system by irradiating slitted pins, natural circulation test from 30 MWt, and in-core measurement of coolant flow rate of each core subassembly during April 1985

  5. Safety problems in fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Fast neutron reactors fuels have a high proportion of plutonium and undergo severe irradiation. Risks during spent fuel reprocessing and subsequent fabrication will depend on isotopic composition, fission product content, physico-chemical form of products, quantities handled. These risks (criticality, contamination, irradiation) are listed for the different steps of the cycle and methods used to control the risks (chemical reaction yields, equipment reliability, intervention, conditions...) are indicated. Problem arising from wastes and effluents produced at each step are briefly given

  6. Alternative concept for a fast energy amplifier accelerator driven reactor

    International Nuclear Information System (INIS)

    Recently Rubbia et al. introduced a conceptual design of a Fast Energy Amplifier (EA) as an advanced innovative reactor which utilizes a neutron spallation source induced by protons as an external source in a subcritical array imbibed a molten lead coolant which, besides being breeder and waste burner, generates energy. This paper introduces some qualitative changes in Rubbia's concept such as more than one point of spallation, in order to reduce the requirement in the energy and current of the accelerator, and mainly to make a more flat neutron distribution. The subcritical core which in Rubbia's concept is an hexagonal array of pins immersed in a molten lead coolant is replaced by a concept of a solid lead calandria with the fuel elements in channels cooled by helium, allowing on line refueling or shuffling, and the utilization of a direct thermodynamic cycle (Brayton), which is more efficient than a vapor cycle. Although the calculations to demonstrate the feasibility of the EA alternative concept are underway and not yet finished, these ideas do not violate the basic physics of the EA, as showed in this paper, with evident advantages in the fuel cycle (on line refueling); reduced requirements in the accelerator complex, which is more realistic and economical in today accelerators technology; and finally the utilization of He as coolant compared with molten Pb is more close to the proved technology given the know how of gas cooled reactors and more efficient from the thermodynamic point of view, allowing simplification and the utilization in other process, besides electricity generation, as hydrogen generation. (author)

  7. Design study of high breeding fast reactor

    International Nuclear Information System (INIS)

    Aiming to increase fuel breeding capability as the most essential feature of fast breeders, an idea of the FP gas purge/tube-in-shell type metallic fuel assembly is proposed. It makes volume fraction of fuel high as more than 50% and realizes a very hard neutron spectrum in the core. The structure of the fuel assembly, its fabrication and the FP gas purging mechanism were assessed and it is clarified that the new concept of the fuel assembly is engineeringly feasible. FP gas purging does not affect shielding structure and can be managed by a small scale cover-gas treatment system because of good trapping characteristics of bonding sodium in the assembly as expected. The fuel handling system without forced cooling is possible. Other reactor components such as IHX were also evaluated. Thus, a concept of the total reactor system of a fast breeding reactor of 670 MWe with the ultra-high breeding ratio of 1.84 and the short reactor doubling time of 6.7 years was obtained. (author)

  8. Sodium-cooled Fast Reactor Core Designs for the TRU burning with Thorium blanket

    International Nuclear Information System (INIS)

    In this study, the SFR(Sodium-cooled Fast Reactor) burner cores are designed with thorium blanket to have smaller burnup reactivity swing but higher TRU burning capability than the typical SFR burner cores using the TRU-U-10Zr fuel. Furthermore, we expect the SFR burner cores using thorium blanket have smaller coolant void reactivity because of the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. From the results, it is found that use of the thorium blanket both in inner and outer cores gives several desirable features such as the reduction of sodium void worth, small burnup reactivity swing but less negative Doppler coefficient and reduced control rod worth and that the use of thorium blanket only in the inner core gives much smaller sodium void worth but larger burnup reactivity swing than the cores using thorium blanket both in the inner and outer cores

  9. European lead fast reactor-ELSY

    International Nuclear Information System (INIS)

    Highlights: → ELSY, the European Lead Fast Reactor (LFR) design is presented. → Presentation of Main Components design. → Core design, safety systems and safety analysis. → Future development activities for Lead-cooled system. - Abstract: The conceptual design of the European Lead Fast Reactor is being developed starting from September 2006, in the frame of the EU-FP6-ELSY project. The ELSY (European Lead-cooled System) reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, while fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative Decay Heat Removal (DHR) systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; it has been demonstrated that the high density of lead can be mitigated by compact solutions and adoption of seismic isolators

  10. Risk Management for Sodium Fast Reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R; Groth, Katrina; Cardoni, Jeffrey N; Wheeler, Timothy A.

    2015-01-01

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

  11. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

    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.)

  12. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

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

  13. Advanced Fast Reactor - 100 - Design Overview

    International Nuclear Information System (INIS)

    The Advanced Fast Reactor-100 (AFR-100) is a small modular sodium-cooled fast reactor with an electrical power output of 100MWe. The AFR-100 has a long-lived core that does not require refueling for 30 years. The concept contains various innovations such as a small compact modular core (both vented and non-vented fuel pins), advanced core shielding materials, a compact fuel handling system, advanced electromagnetic pumps, compact intermediate heat exchangers, and a direct reactor auxiliary cooling system. These advanced systems and components were adopted in order to reduce the overall size of the primary heat transport system (and therefore the overall commodities and cost), enhance safety, and to improve overall plant performance. This paper presents the summary results of a year-long study that culminated in the design of two primary heat transport configurations for the AFR-100. The paper describes those innovations and shows how they are integrated into the overall AFR-100 primary heat transport system design. (author)

  14. ELSY - The European lead fast reactor

    International Nuclear Information System (INIS)

    The European Lead Fast Reactor is being developed starting from September 2006, in the frame of the ELSY (European Lead SYstem) project funded by the Sixth Framework Programme of EURATOM. The project, coordinated by Ansaldo Nucleare, involves a wide consortium of European organizations. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. The main objectives of the ELSY project are to show that: the adopted innovative design and technology achieves a very high safety standards; the fuel cycle can be closed; the non-proliferation resistance is enhanced; a high availability factor is reached; the economic competitiveness target is reached; the design is compliant with GIF goals. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied and compared. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative passively operated DHR (Decay

  15. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  16. Status of the DEBENE fast breeder reactor development, March 1979

    International Nuclear Information System (INIS)

    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

  17. Fast reactor passive shutdown system: LIM

    International Nuclear Information System (INIS)

    To enhance the inherent safety of the fast breeder reactor (FBR), unique attempts are being made in reactivity control systems design to achieve maintenance-free and reliable performance. The design proposed is the lithium injection module (LIM) for inherent ultimate shutdown. Reactor physics calculation revealed the reactivity worth of LIM in a 60 MWe metal-fueled FBR and a 1,000 MWe mixed-oxide-fueled FBR. An experimental verification on the freeze seal design assured an accurate injection temperature of LIM. Reliability, maintainability, and real time monitoring for LIM is also discussed. A definite advantage over the conventional self-actuated shutdown system (SASS) has been presented. LIM offers substantial inherent safety with improved maintainability. (author)

  18. Fast critical experiment data for space reactors

    International Nuclear Information System (INIS)

    Data from a number of previous critical experiments exist that are relevant to the design concepts being considered for SP-100 and MMW space reactors. Although substantial improvements in experiment techniques have since made some of the measured quantities somewhat suspect, the basic criticality data are still useful in most cases. However, the old experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, we have calculated about 20 fast benchmark critical experiments with the latest ENDF/B data and modern transport codes. These calculations were undertaken as a part of the planning process for a new series of benchmark experiments aimed at supporting preliminary designs of SP-100 and MMW space reactors

  19. Fuel systems for compact fast space reactors

    International Nuclear Information System (INIS)

    About 200 refractory metal clad ceramic fuel pins have been irradiated in thermal reactors under the 1200 K to 1550 K cladding temperature conditions of primary relevance to space reactors. This paper reviews performance with respect to fissile atom density, operating temperatures, fuel swelling, fission gas release, fuel-cladding compatibility, and consequences of failure. It was concluded that UO2 and UN fuels show approximately equal performance potential and that UC fuel has lesser potential. W/Re alloys have performed quite well as cladding materials, and Ta, Nb, and Mo/Re alloys, in conjunction with W diffusion barriers, show good promise. Significant issues to be addressed in the future include high burnup swelling of UN, effects of UO2-Li coolant reaction in the event of fuel pin failure, and development of an irradiation performance data base with prototypically configured fuel pins irradiated in a fast neutron flux

  20. Fast reactor technology innovation and visualization

    International Nuclear Information System (INIS)

    Innovations in safety, operations, and maintenance for improving the availability, reliability, and capital cost of the sodium fast reactor are described. Concerning safety these innovations deal with on-line limiting safety settings, inherent core protection, detection of subassembly coolant mis-allocation. Concerning reactor operations these innovations deal with advanced energy conversion, adapting non-base load nuclear plants and on-line diagnostics. Other innovations concern inspection, servicing, refueling. The development of these innovations rely on visualization technology for their use and for demonstration of improvements achievable. A visualization platform for running these innovations and the nuclear plant thermal-hydraulic, structure, and process codes that underlie them are described. The platform hardware consists of a large-scale tiled display and a haptic hand-controller and in the future will grow to include a high-speed network and multiple graphics-client systems

  1. Chemistry for fast reactor fuel cycle

    International Nuclear Information System (INIS)

    The fuel cycle for the fast reactors poses several challenging chemistry issues. The use of fuels with high plutonium content, the variety of fuel matrices (oxides, carbides, metal alloys), the high burn-up to which the fuel is driven and the need to close the fuel cycle with minimum out-of-pile inventory are examples of special features of fast reactors. The need to reduce waste generation and the need to identify matrices for safe long term disposal of waste are additional issues that need a chemist's attention. As a chemist, the subject of actinide separations has been very stimulating to me, with a myriad of interesting possibilities and at the same time, demanding careful attention to the unique chemistry of the actinides including multiplicity of oxidation states. The presence of high concentrations of plutonium in the reprocessing streams introduces issues such as third phase formation, which provides an incentive for the development of candidates for solvent extraction as alternatives to tri-n-butyl phosphate, currently used for the Purex reprocessing scheme. With the advent of supercritical fluid extraction as a tool for actinide recovery from a variety of matrices, and the potential of room temperature ionic liquids to offer significant advantages in actinide processing, actinide separations is an element of fast reactor fuel cycle that is full of opportunities and challenges. The need to process metallic alloy fuels using molten salt electrorefining as the route, adds further to the challenges. The presentation will highlight some of the recent progress achieved in this area at IGCAR. (author)

  2. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author)

  3. Fast Reactor Programme. Third Quarter 1969. Progress Report

    International Nuclear Information System (INIS)

    The RCN research programme on fast spectrum nuclear reactors comprises reactor physics, fuel performance, radiation damage in canning materials, corrosion behaviour in canning materials, aerosol research and heat transfer and hydraulics. An overview is given of the fast reactor experiments at the STEK critical facility in Petten, the Netherlands, in the third quarter of 1969

  4. A review of the UK fast reactor programme, March 1979

    International Nuclear Information System (INIS)

    The Status report of the UK activities related to fast-breeder reactor activities includes the following: summary of the operating experience of the prototype Fast Reactor (PFR) during 1978; design studies of the commercial demonstration fast reactor (CDFR); design studies of later advanced LMFBR; engineering developments of high temperature sodium loop, steam generators and instrumentation; materials development; corrosion problems; sodium technology; fuel elements development; PFR fuel reprocessing; safety issues molten fuel-coolant interaction; core structure test; accident analysis; reactor performance studies; experimental reactor physics; fuel management and general neutronics calculation for CDFR; reactor instruments

  5. Safe Management Of Fast Reactors: Towards Sustainability

    International Nuclear Information System (INIS)

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

  6. Actinide management with commercial fast reactors

    Science.gov (United States)

    Ohki, Shigeo

    2015-12-01

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GWey if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

  7. The Argentine-Brazilian fast reactor programme

    International Nuclear Information System (INIS)

    This paper summarizes the Argentine-Brazilian Fast Reactor Programme and gives reasons for the decision of a binational venture. The work carried out by both countries is described, showing how they complement each other, with the corresponding saving of resources. The main objectives of the Programme and tentative schedules in three progressing integrating stages are given and the present nuclear know-how in each country is identified as a good starting point. The paper also gives some details regarding the economical and human resources involved. (author). 1 graph

  8. Actinide management with commercial fast reactors

    International Nuclear Information System (INIS)

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GWey if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel

  9. Integral Fast Reactor fuel pin processor

    International Nuclear Information System (INIS)

    This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves

  10. Actinide management with commercial fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ohki, Shigeo [Japan Atomic Energy Agency, 4002, Narita-cho, O-arai-machi, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

    2015-12-31

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GW{sub e}y if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

  11. Small power sodium cooled fast nuclear reactors

    International Nuclear Information System (INIS)

    1.5 MW(e), 12 MW(e) and 170 MW(e) small power sodium cooled fast reactors have been developed. The reactor plants were developed as universal power units for economically effective energy and industrial steam generation and heat supply. The main features increasing the power unit economic efficiency are: serial fabrication of standard RPs at the factory and delivery of reactor vessels in ready made form; realization of self-protection principles and use of passive systems in RP; use of standard machine room equipment, fabricated in accordance with the rules of conventional heat power engineering; use of turbine plant with thermodynamic coefficient, exceeding the corresponding value for the plants of PWR type. For MBRU-1.5 and MBRU-12 RPs it is proposed to use a core without FA replacement during the whole service life (30 years) and for BMN-170 RP it is proposed to use a core with a 4 year operating period and 1 year between the refueling shutdowns. During the whole service life a minimal number of operating personnel will be needed for the plant servicing. The personnel functions will be periodically to observe the parameters of technological process. Passive principles are used in the main RP safety systems: a passive type system of emergency residual heat removal system provides heat removal directly through the reactor vessel forced air cooling due to the natural air chimney effect; an emergency reactor shut-down system is provided by emergency protection rods with active-passive action. (author)

  12. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

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

  13. Fast reactors and advanced light water reactors for sustainable development

    International Nuclear Information System (INIS)

    Complete text of publication follows: The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been re-acknowledged worldwide. In response to this international movement, the papers compiling the most recent findings in the fields of fast reactors (FR) and advanced light water reactors (LWR) were gathered and published in this special issue. This special issue compiles six articles, most of which are very meticulously performed studies of the multi year development of design and assessment methods for large sodium-cooled FRs (SFRs), and two are related to the fuel cycle options that are leading to a greater understanding on the efficient utilization of energy resources. The Japanese sodium-cooled fast reactor (JSFR) is addressed in two manuscripts. H. Yamano et al. reviewed the current design which adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. Their safety assessments of both design basis accidents and severe accidents indicate that the devised JSFR satisfies well their risk target. T. Takeda et al. discussed the improvement of the modeling accuracy for the detailed calculation of JSFR's features in three areas: neutronics, fuel materials, and thermal hydraulics. The verification studies which partly use the measured data from the prototype FBR Monju are also described. Two of these manuscripts deal with those aspects of advanced design of SFR that have hitherto not been explored in great depth. The paper by G. Palmiotti et al. explored the possibility of using the sensitivity methodologies in the reactor physics field. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described. F. Baque et al. reviewed the evolution of the in

  14. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-07

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

  15. Research activities on fast reactors in Switzerland

    International Nuclear Information System (INIS)

    The current domestic Swiss electricity supply is primarily based on hydro power (approximately 61%) and nuclear power (about 37%). The contribution of fossil systems is, consequently, minimal (the remaining 2%). In addition, long-term (but limited in time) contracts exist, securing imports of electricity of nuclear origin from France. During the last two years, the electricity consumption has been almost stagnant, although the 80s recorded an average annual increase rate of 2.7%. The future development of the electricity demand is a complex function of several factors with possibly competing effects, like increased efficiency of applications, changes in the industrial structure of the country, increase of population, further automation of industrial processes and services. Due to decommissioning of the currently operating nuclear power plants and expiration of long-term electricity import contracts there will eventually open a gap between the postulated electricity demand and the base supply. The assumed projected demand cases, high and low, as well as the secured yearly electric energy supply are shown. The physics aspects of plutonium burning fast reactor configurations are described including first results of the CIRANO experimental program. Swiss research related to residual heat removal in fast breeder reactors is presented. It consists of experimental ana analytic investigations on the mixing between two horizontal fluid layers of different velocities and temperatures. Development of suitable computer codes for mixing layer calculation are aimed to accurately predict the flow and temperature distribution in the pools. A satisfactory codes validation based on experimental data should be done

  16. History of fast reactor fuel development

    Science.gov (United States)

    Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

    1993-09-01

    The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

  17. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    International Nuclear Information System (INIS)

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

  18. A small modular fast reactor as starting point for industrial deployment of fast reactors

    International Nuclear Information System (INIS)

    The current commercial reactors based on light water technology provide 17% of the electricity worldwide owing to their reliability, safety and competitive economics. In the near term, next generation reactors are expected to be evolutionary type, taking benefits of extensive LWR experience feedbacks and further improved economics and safety provisions. For the long term, however, sustainable energy production will be required due to continuous increase of the human activities, environmental concerns such as greenhouse effect and the need of alternatives to fossil fuels as long term energy resources. Therefore, future generation commercial reactors should meet some criteria of sustainability that the current generation cannot fully satisfy. In addition to the current objectives of economics and safety, waste management, resource extension and public acceptance become other major objectives among the sustainability criteria. From this perspective, two questions can be raised: what reactor type can meet the sustainability criteria, and how to proceed to an effective deployment in harmony with the high reliability and availability of the current nuclear reactor fleet. There seems to be an international consensus that the fast spectrum reactor, notably the sodium-cooled system is most promising to meet all of the long term sustainability criteria. As for the latter, we propose a small modular fast reactor project could become a base to prepare the industrial infrastructure. The paper has the following contents: - Introduction; - SMFR project; - Core design; - Supercritical CO2 Brayton cycle; - Near-term reference plant; - Advanced designs; - Conclusions. To summarize, the sodium-cooled fast reactor is currently recognized as the technology of choice for the long term nuclear energy expansion, but some research and development are required to optimize and validate advanced design solutions. A small modular fast reactor can satisfy some existing near-term market niche

  19. Minor actinide burning in dedicated lead-bismuth cooled fast reactors

    International Nuclear Information System (INIS)

    The destruction of minor actinides (MA) in dedicated burners is of contemporary interest in Europe and Japan because it requires the deployment of smaller number of special transmutation facilities. A major fraction of Pu from spent LWR fuel can be then burned in PWRs (or fast reactors) using dedicated fertile-free fuel assemblies. However, the design of MA burning fast spectrum cores poses significant challenges because of deterioration of key safety parameters, in particular of the coolant void coefficient. This study proposes the concept of an lead-bismuth eutectic (LBE)-cooled dedicated MA burner having metallic fuel (MA-Pu-Zr) and streaming assemblies to attain acceptable coolant void worth performance. It is shown that a large 1800 MWth fertile-free core containing 37 wt% TRU with very high fraction of MA(59 wt%) from LWR spent fuel can be burned in a first cycle for 700 EFPDs with a very small reactivity swing: less than βeff. Moreover, the reactivity void worth is negative for a fully voided core when all surrounding coolant is kept at reference density. However, the core reactivity increases as coolant density falls from the reference value of 10.25 to 6 g/cm3. Because its coolant density coefficient value is less than that of a sodium cooled IFR, the concept provides good potential for the achievement of self-regulation characteristics in unprotected events, provided that small negative fuel temperature feedback can be maintained. (authors)

  20. 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)

    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

  1. International Experience with Fast Reactor Operation and Maintenance

    International Nuclear Information System (INIS)

    This paper reviews the most important lessons learned from operation of the world’s sodium cooled fast reactors, both test reactors and power producing reactors, which represent nearly 400 reactor-years of cumulative operating experience. The first reactor in the world to produce electricity was a fast reactor, the Experimental Breeder Reactor I, in December 1951. International experience with fast reactor technology exists in France, Germany, India, Japan, the Russian Federation, the United Kingdom and the United States of America. The operating experience with these reactors has been mixed; early problems were associated with fuel cladding, steam generators, fuel handling and sodium leakage. Excellent experience has been gained, however, that demonstrates the robust nature of the technology, the potential for exceedingly safe designs, ease of maintenance, ease of operation and the ability to effectively manage waste from spent fuel. It is a mature technology. (author)

  2. Seminar on Heat-transfer fluids for fast neutron reactors

    International Nuclear Information System (INIS)

    This book reports the content of a two-day meeting held by the Academy of Sciences on the use of heat-transfer fluids in fast neutron reactors. After a first part which proposes an overview of scientific and technical problems related to these heat-transfer fluids (heat transfer process, nuclear properties, chemistry, materials, risks), a contribution proposes a return on experience on the use of heat-transfer fluids in the different design options of reactors of fourth generation: from mercury to NaK in the first fast neutron reactor projects, specific assets and constraints of sodium used as heat-transfer fluid, concepts of fast neutron reactors cooled by something else than sodium, perspectives for projects and research in fast neutron reactors. The next contribution discusses the specifications of future fast-neutron reactors: expectations for fourth-generation reactors, expectations in terms of performance and of safety, specific challenges. The last contribution addresses actions to be undertaken in the field of research and development: actions regarding all reactor types or specific types as sodium-cooled reactors, lead cooled reactors, molten salt reactors, and gas-cooled fast reactors

  3. LFR "Lead-Cooled Fast Reactor"

    Energy Technology Data Exchange (ETDEWEB)

    Cinotti, L; Fazio, C; Knebel, J; Monti, S; Abderrahim, H A; Smith, C; Suh, K

    2006-05-11

    The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small, transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology

  4. Indian fast reactor technology: Current status and future programme

    Indian Academy of Sciences (India)

    S C Chetal; P Chellapandi

    2013-10-01

    The paper brings out the advantages of fast breeder reactor and importance of developing closed nuclear fuel cycle for the large scale energy production, which is followed by its salient safety features. Further, the current status and future strategy of the fast reactor programme since the inception through 40 MWt/13 MWe Fast Breeder Test Reactor (FBTR), is highlighted. The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast Reactor (SFRs) and timely introduction of metallic fuel reactors with emphasis on breeding gain and enhanced safety are being brought out in this paper.

  5. The economics of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Specific technical features of fast reactors make them more expensive than water-cooled reactors in terms of initial investment, an over cost of 30% is acknowledged in this study. Their consumption of natural uranium is negligible being fed on depleted uranium (except for the very first cycle when an important quantity of plutonium is necessary). In the context of the scarcity of natural uranium, fast reactors could provide a competitive KWh compared with PWR. The study shows that sodium-cooled fast reactor could be economically competitive somewhere in the second part of the 21. century. The development of fast reactors could be accelerated by other arguments than economic competitiveness, for instance some governments might value more the energy independence given by a fleet of fast reactors or by considerations linked to non-proliferation or to the burning of actinides. In addition the article details the worldwide resource in natural uranium. (A.C.)

  6. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  7. Sodium fast neutron reactors. Status and perspective of development

    International Nuclear Information System (INIS)

    This report reveals data on development history of domestic fast neutron reactors cooled with sodium (BN reactors). It also shows BN reactors' unique role in expanding source of nuclear power raw materials and in solving ecological problems relating to radioactive wastes. There is brief information on characteristics and operation experience of research reactors BR-10, BOR-60, pilot-industrial reactors BN-350 and BN-600. As well there is data on BN-800 reactor designing that obtained a license for building. There are considered BN reactor peculiarities in regard of safety and design decisions on safety provision at the level meeting standard document requirements. BN reactor technical and economic indices and the ways of their improvement are evaluated. There is brief information on alternative perspective technologies of fast reactors, in particular regarding 'BREST-300' reactor cooled with lead coolant

  8. Trial visualization of fast reactor design knowledge

    International Nuclear Information System (INIS)

    In design problems of large-scale systems like fast breeder reactors, inter-relations among design specifications are very important where a selected specification option is transferred to other specification selections as a premise to be taken account in engineering judgments. These inter-relations are also important in design case studies with the hypothetical adoption of rejected design options for the evaluation of deviation propagations among design specifications. Some of these rejected options have potential worth for future reconsideration by some circumstance changes (e.g., advanced simulations to exclude needs for mock-up tests, etc.), to contribute to flexibility in system designs. In this study, a computer software is built to visualize a design problem structure by representing engineering knowledge nodes on individual specification selections along with inter-relations of design specifications, to validate the knowledge representation method and to derive open problems. (author)

  9. Fast Reactor Fuel Development in Japan

    International Nuclear Information System (INIS)

    The future fast reactor and its fuel cycle system under development in Japan uses oxide fuel with simplified pelletizing fuel fabrication technology as a reference concept. Its driver fuel consists of large diameter annular fuel pellets, oxide dispersion strengthened ferritic steel cladding fuel pins with a ferritic-martensitic steel subassembly wrapper tube and minoractinide- bearing oxide fuel. The target burnup of the driver fuel is 150 GW.d/t in discharge average, which corresponds to 250 GW.d/t of peak burnup and 250 dpa of peak neutron dose. Fuel developmental efforts, including out-of-pile studies such as material characteristics experimental evaluation and fuel property measurements, various irradiation tests and fuel fabrication technology developments were planned and are in progress. Future fuels will be realized through Joyo irradiation tests and Monju demonstrations. International collaborative efforts are also an important part of such activities. (author)

  10. Safeguards in Prototype Fast Breeder Reactor Monju

    International Nuclear Information System (INIS)

    The assemblies loaded in the core and stored in the ex-vessel storage tank (EVST) are in liquid sodium in the Japanese prototype fast breeder reactor (FBR) Monju. Since it is difficult to apply a direct verification procedure for the fuel assemblies in these areas, a dual containment and surveillance system consisting of two monitoring devices such as surveillance camera and radiation monitor that are functionally independent has been applied. In addition, the Monju Remote Monitoring System was developed to strengthen the continuous surveillance and to reduce the load of the inspection activities. Furthermore, the ex-vessel transfer machine radiation monitor (EVRM) and the exit gate monitor (EXGM) were upgraded to strengthen the monitoring of spent blanket fuel assemblies and to improve the reliability of distinguishing between fuel assemblies and non-fuel items. As the result, the integrated safeguards was introduced in November 2009, and the effective safeguards activities have been implemented in Monju. (author)

  11. Sodium technology for fast breeder reactors

    International Nuclear Information System (INIS)

    Sodium, because of its good heat transfer and nuclear properties, is used as a coolant in fast reactors. It is also used largely as a reducing agent in pharmaceutical, perfumery and general chemical industries. Its affinity to react with air and water is a strong disadvantage. However, this is fully understood and the design of engineering systems take care of this aspect. With several experimental and test facilities established over the years in this country as well as abroad, the 'sodium technology' has reached a level of maturity. The design of sodium systems considering all the physical and chemical properties and the developmental work carried out at Indira Gandhi Centre for Atomic Research are broadly covered in this report. (author)

  12. Fabrication of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    Mixed oxide (MOX) (U,Pu)O2, and metallic (U,Pu ,Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity , low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion. The higher coefficient of linear expansion is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burnup, fuel cladding interaction and lower margin between operating and melting temperature. The optimal solution may lie in cermet fuel (U, PuO2), where PuO2 is dispersed in U metal matrix and combines the favorable features of both the fuel types. The advantages of this fuel include high thermal conductivity, larger margin between melting and operating temperature, ability to retain fission product etc. The matrix being of high density metal the advantage of high breeding ratio is also maintained. In this report some results of fabrication of cermet pellet comprising of UO2/PuO2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

  13. Methane reforming with fast nuclear reactor steam

    International Nuclear Information System (INIS)

    The paper considers the concept of utilizing nuclear fast reactor (FR) with a sodium coolant for methane steam reforming. Steam conditions of a power FR, e.g. the BN-600 now operating in Russia: steam pressure P=13.2 MPa and steam temperature T=500degC, do not absolutely comply with the catalytic reactor working parameters, which produces a synthetic gas (syngas), a mix of hydrogen and carbon oxide. In this connection, the present paper addresses a possibility of utilizing steam produced in one of three independent the BN-600 loops in an amount of 640 t/h for preparing a gas-steam mixture with T=500degC and its additional heating in a converter up to the operating temperature, T=850degC, at the expense of natural gas burning or electrical energy supplying. In this case, the fraction of burned natural gas burning or electrical energy supplying. In this case, the fraction of burned natural gas significantly decreases. It is estimated that steam parameters of the BN-600 afford to obtain ∼3·105 nm3/h of hydrogen. It is also considered a concept of nuclear heat transfer to remote regions to be achieved with the aid of syngas incoming from the converter, its cooling further and transmitting through a pipeline to the place of its utilization, where it is restored into methane with the heat extraction. (author)

  14. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  15. History of fast reactor fuel development

    International Nuclear Information System (INIS)

    The first fast breeder eactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s. (orig.)

  16. Relationship between core size, coolant choice, fuel type, and neutron flux in a fast irradiation test reactor

    International Nuclear Information System (INIS)

    Currently, the United States has no domestic capability for large volume irradiation testing in a fast-spectrum system to support the development of advanced fuels or materials. The recently-proposed Global Nuclear Energy Partnership includes provisions for a sodium-cooled Advanced Burner Test Reactor which could provide this testing capability. In addition to sodium, lead-bismuth eutectic and helium coolants are being considered for future energy systems. In this paper, sodium, lead-bismuth eutectic, and helium-cooled systems are evaluated to determine the impact on fast flux and fuel enrichment resulting from varying core diameter, fuel volume fraction, fuel type, and coolant. While fast flux is most strongly influenced by core diameter at fixed power, fuel enrichment is a more complicated function of all four parameters. In the end, the combination of high fast flux and low enrichment can best be achieved by a sodium-cooled system. (authors)

  17. A review of fast reactor progress in Japan

    International Nuclear Information System (INIS)

    The fast reactor development project in Japan is continuing at a slightly increased scale of effort in budget. The total budget for LMFBR development for fiscal year 1978 was 24 billion yen. In August 1977 major industries engaged in LMFBR have set up an office where design work can be jointly conducted. Highlights and topics of the fast reactor development activities cover description of JOYO reactor, its first criticality experiment, and the prototype fast breeder MONJU. Research and development programmes dealt with fission products release and its possible interaction with the soodium coolant, inspection of reactor components, experiments simulating sodium leakage, development of steam generator

  18. Materials development and materials selection for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    As applied to operational conditions of fast reactors a combined investigations of structural materials are accomplished. Steels 10Kh18N9, 08Kh16N11M3 are recommended to be used for reactor vessels, steel 10Kh2M - for steam generators, a strain hardened steel 08Kh16N11M3T and steel 05Kh12N2M - for fuel assembly cans. The investigations provided for designing such sodium cooled fast reactors as Bor-60, BN-350, BN-600. The investigation results are now in use in construction of a new fast reactor BN-800

  19. Multigroup fast fission factor treatment in a thermal reactor lattice

    International Nuclear Information System (INIS)

    A multigroup procedure for the studies of the fast fission effects in the thermal reactor lattice and the calculation of the fast fission factor was developed. The Monte Carlo method and the multigroup procedure were combined to calculate the fast neutron interaction and backscattering effects in a reactor lattice. A set of probabilities calculated by the Monte Carlo method gives a multigroup spectrum of neutrons coming from the moderator and entering the fuel element. Thus, the assumptions adopted so far in defining and calculating the fast fission factor has been avoided, and a new definition including the backscattering and interaction effects in a reactor lattice have been given. (author)

  20. Simulating the Behaviour of the Fast Reactor Joyo (Draft)

    International Nuclear Information System (INIS)

    Motivated by the development of fast reactors the behaviour of the Japanese experimental fast reactor Joyo is simulated with two Monte Carlo codes: Monte Carlo NParticle (MCNP) and Probabilistic Scattering Game (PSG). The simulations are based on the benchmark study 'Japan's Experimental Fast Reactor Joyo MKI core: Sodium-Cooled Uranium-Plutonium Mixed Oxide Fueled Fast Core Surrounded by UO2 Blanket'. The study is focused on the criticality of the reactor, control rod worth, sodium void reactivity and isothermal temperature coefficient of the reactor. These features are calculated by applying both homogeneous and heterogeneous reactor core models that are built according to the benchmark instructions. The results of the two models obtained by the two codes are compared with each other and especially with the experimental results presented in the benchmark. (author)

  1. Implications of Fast Reactor Transuranic Conversion Ratio

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays

    2010-11-01

    Theoretically, the transuranic conversion ratio (CR), i.e. the transuranic production divided by transuranic destruction, in a fast reactor can range from near zero to about 1.9, which is the average neutron yield from Pu239 minus 1. In practice, the possible range will be somewhat less. We have studied the implications of transuranic conversion ratio of 0.0 to 1.7 using the fresh and discharge fuel compositions calculated elsewhere. The corresponding fissile breeding ratio ranges from 0.2 to 1.6. The cases below CR=1 (“burners”) do not have blankets; the cases above CR=1 (“breeders”) have breeding blankets. The burnup was allowed to float while holding the maximum fluence to the cladding constant. We graph the fuel burnup and composition change. As a function of transuranic conversion ratio, we calculate and graph the heat, gamma, and neutron emission of fresh fuel; whether the material is “attractive” for direct weapon use using published criteria; the uranium utilization and rate of consumption of natural uranium; and the long-term radiotoxicity after fuel discharge. For context, other cases and analyses are included, primarily once-through light water reactor (LWR) uranium oxide fuel at 51 MWth-day/kg-iHM burnup (UOX-51). For CR<1, the heat, gamma, and neutron emission increase as material is recycled. The uranium utilization is at or below 1%, just as it is in thermal reactors as both types of reactors require continuing fissile support. For CR>1, heat, gamma, and neutron emission decrease with recycling. The uranium utilization exceeds 1%, especially as all the transuranic elements are recycled. exceeds 1%, especially as all the transuranic elements are recycled. At the system equilibrium, heat and gamma vary by somewhat over an order of magnitude as a function of CR. Isotopes that dominate heat and gamma emission are scattered throughout the actinide chain, so the modest impact of CR is unsurprising. Neutron emitters are preferentially found

  2. Thermal-hydraulic analyses of transients in an actinide-burner reactor cooled by forced convection of lead-bismuth

    International Nuclear Information System (INIS)

    The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed

  3. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

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

  4. Development of studies on helium cooled fast reactors

    International Nuclear Information System (INIS)

    A necessity is shown of developing breeders with high reproductive properties. Helium cooled fast reactor is considered. The reactor performances, heating circuit with the use of a steam turbine unit in the secondary circuit is outlined. The reactor design and fuel assemblies are described

  5. The design of the Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    India has a moderate uranium reserve and a large thorium reserve. The primary energy resource for electricity generation in the country is coal. The potential of other resources like gas, oil, wind, solar and biomass is very limited. The only viable and sustainable resource is the nuclear energy. Presently, Pressurised Heavy Water Reactors utilizing natural uranium are in operation/under construction and the plutonium generated from these reactors will be multiplied through breeding in fast breeder reactors. The successful construction, commissioning and operation of Fast Breeder Test Reactor at Kalpakkam has given confidence to embark on the construction of the Prototype Fast Breeder Reactor (PFBR). This paper describes the salient design features of PFBR including the design of the reactor core, reactor assembly, main heat transport systems, component handling, steam water system, electrical power systems, instrumentation and control, plant layout, safety and research and development

  6. Regenerative burner

    Energy Technology Data Exchange (ETDEWEB)

    Davies, T.E.; Quinn, D.E.; Watson, J.E.

    1986-08-05

    A regenerative burner is described operable in fire and flue modes comprising: a burner shell having first and second internal chambers, the first chamber being disposed on the flame axis of the burner and the second chamber surrounding the radial perimeter of the first chamber; a gas permeable annular regenerative bed separating the first and second chambers such that gas flow between the first and second chambers must travel through the regenerative bed in a generally radial direction with respect to the flame axis; means for supplying combustion air to the second chamber when the burner is in the fire mode and for exhausting the products of combustion from the second chamber when the burner is in the flue mode; and means for supplying fuel in the vicinity of the flame axis for mixing with combustion air to support combustion when the burner is in the fire mode.

  7. Model of fast reactor knowledge preservation system

    International Nuclear Information System (INIS)

    Despite lack of the energy market today, fast reactors (FR) in the closed nuclear fuel cycle are the basis of a full-scale development of nuclear power in future. However, there are serious problems concerning the future R and D of these reactor technologies related to the following obstacles. All research on FR was stopped in Germany, Italy, United Kingdom and the United States and the work performed only dealt with the decommissioning of FR. Many experts who participated in R and D programs to create FR have retired or are approaching retirement age. In France, Japan and Russia work on the development of FR still continues, but there is a lack of young scientists and engineers. Due to all this factors IAEA launched the initiative to combine efforts of the leading nuclear countries to develop a project for the preservation of knowledge in the field of scientific and technological problems of FR development. Efforts of IAEA and national experts resulted in a model of FR information search and classification (so called ). This work has initiated a systematic process of creation and filling of information data bank on various aspects of FR design and operation. As the next step it would be logical to develop self-consistent mathematical models of FR-based NPP and closed NFC with their subsequent introduction into the system of knowledge preservation. So, it will serve as an important step towards preservation of knowledge in the field of FR design through joint development and to ensure open access to software. Such a project may lay the groundwork for the future development of distance learning courses and training on the optimal FR design, with the participation of leading specialists in this field. The report provides a mathematical and logical model for the preservation of knowledge concerning FR science and technology: taxonomy, an engineering model of FR-based NPP, a FR NFC model

  8. Regenerative burner

    Energy Technology Data Exchange (ETDEWEB)

    Gitman, G.M.

    1990-05-08

    This patent describes a method of combusting fuel in a furnace having a pair of regenerative burners, each burner having a combustion chamber. It comprises: supplying fuel and oxygen alternatively to each burner to create alternating firing burners wherein the oxygen is supplied from two sources providing first and second oxidizing gases having different oxygen concentrations and simultaneously alternating the application of negative pressure to the remaining non-firing burner to recover heat from flue gases exhausted by the regenerative bed of the non-firing burner to be used further to preheat at least part of the oxygen being supplied to the firing burner; mixing the fuel with a fraction of the oxygen under substoichiometric combustion condition to create products of incomplete combustion to form a hot, luminous flame core containing partially pyrolized fuel; and mixing the partially pyrolyzed fuel with a remaining fraction of the oxygen to complete combustion of the pyrolized fuel; and controlling the total flow of fuel and oxygen supplied to each burner to provide each burner with a desired flame stoichiometry.

  9. A dry process fuel cycle analysis for a sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Joon Jeong; Gyu, Hong Roh; Hangbok, Choi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2005-07-01

    The Korean nuclear fuel cycle with the dry-processed oxide fuel sodium-cooled fast reactor (SFR) has been studied by the dynamic analysis method. In addition to the SFR, the current operating Pressurized Water Reactor (PWR) and Canadian Deuterium Uranium (Candu) reactors were considered. For the analysis, the equilibrium fuel cycle of the reference core was chosen from a modified BN-600 core. In the reactor scenario, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 1%. In this study, the spent fuel inventory as well as the amount of plutonium, minor actinides (MA) and fission products (FP) of the recycling fuel cycle were estimated and compared to that of the once-through fuel cycle. The results of the once-through fuel cycle have shown that the demand increases up to 64 GWe and total amount of spent fuel is about 102 kt in 2100. When the dry-processed fuel SFR scenario is implemented, it is expected that the total spent fuel inventory can be reduced by about 80%. Also, the plutonium inventory can be reduced by about 25%. However it was found that the SFR scenario does not contribute to the reduction of the MA and FP, which are important when designing a repository. For a further destruction of the MA, an actinide burner can be considered in a future nuclear fuel cycle. (authors)

  10. Immobilization of Fast Reactor First Cycle Raffinate

    International Nuclear Information System (INIS)

    This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method

  11. Creep buckling problems in fast reactor components

    International Nuclear Information System (INIS)

    Creep buckling analyses for two important components of 500 M We Prototype Fast Breeder Reactor (PFBR), viz. Intermediate Heat Exchanger (IHX) and Inner Vessel (IV), are reported. The INCA code of CASTEM system is used for the large displacement elasto-plastic-creep analysis of IHX shell. As a first step, INCA is validated for a typical benchmark problem dealing with the creep buckling of a tube under external pressure. Prediction of INCA is also compared with the results obtained using Hoff's theory. For IV, considering the prohibitively high computational cost for the actual analysis, a simplified analysis which involves only large displacement elastoplastic buckling analysis is performed using isochronous stress strain curve approach. From both of these analysis is performed using isochronous stress strain curve approach. From both of these analysis, it has been inferred that creep buckling failure mode is not of great concern in the design of PFBR components. It has also been concluded from the analysis that Creep Cross Over Curve given in RCC-MR is applicable for creep buckling failure mode also. (author). 8 refs., 9 figs., 1 tab

  12. Immobilization of Fast Reactor First Cycle Raffinate

    Energy Technology Data Exchange (ETDEWEB)

    Langley, K. F.; Partridge, B. A.; Wise, M.

    2003-02-26

    This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.

  13. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  14. A review of the U.K. fast reactor programme: March 1978

    International Nuclear Information System (INIS)

    The review of the UK fast reactor programme covers the description of Dounreay Fast Reactor shut down after seventeen years of successful operation; description of prototype fast reactor (PFR); core design parameters safety features and plant design for commercial demonstration fast reactor (CDFR). Engineering development is related to large sodium rigs, coolant circuit hydraulics and vibration, instrumentation and components. The subjects of interest are material development, sodium technology, fast reactor fuel, fuel cycle, reactor safety, reactor performance studies

  15. Technology Options for a Fast Spectrum Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Wachs; R. W. King; I. Y. Glagolenko; Y. Shatilla

    2006-06-01

    Idaho National Laboratory in collaboration with Argonne National Laboratory has evaluated technology options for a new fast spectrum reactor to meet the fast-spectrum irradiation requirements for the USDOE Generation IV (Gen IV) and Advanced Fuel Cycle Initiative (AFCI) programs. The US currently has no capability for irradiation testing of large volumes of fuels or materials in a fast-spectrum reactor required to support the development of Gen IV fast reactor systems or to demonstrate actinide burning, a key element of the AFCI program. The technologies evaluated and the process used to select options for a fast irradiation test reactor (FITR) for further evaluation to support these programmatic objectives are outlined in this paper.

  16. The United States of America fast breeder reactor program

    International Nuclear Information System (INIS)

    The reasons for the development of the fast breeder reactor in the United States are outlined, and the LMFBR program is discussed in detail, under the following headings: program objectives, reactor physics, fuel and materials development, fuel recycle, safety, components, plant experience program (Near Commercial Breeder Reactor). The special facilities to be used at each stage of the program are described. It is planned that the Near Commercial Breeder Reactor will be complete in 1986, and commercial plants should follow in rapid succession. An alternate fast reactor concept (Gas Cooled Fast Reactor) is outlined. The Environmental Impact Statement for the proposed program is summarized, and the cost benefit analysis supplied as part of the Environment Statement is also summarized. (U.K.)

  17. Numerical modelling of the CHEMREC black liquor gasification process. Conceptual design study of the burner in a pilot gasification reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marklund, Magnus

    2001-02-01

    The work presented in this report is done in order to develop a simplified CFD model for Chemrec's pressurised black liquor gasification process. This process is presently under development and will have a number of advantages compared to conventional processes for black liquor recovery. The main goal with this work has been to get qualitative information on influence of burner design for the gas flow in the gasification reactor. Gasification of black liquor is a very complex process. The liquor is composed of a number of different substances and the composition may vary considerably between liquors originating from different mills and even for black liquor from a single process. When a black liquor droplet is gasified it loses its organic material to produce combustible gases by three stages of conversion: Drying, pyrolysis and char gasification. In the end of the conversion only an inorganic smelt remains (ideally). The aim is to get this smelt to form a protective layer, against corrosion and heat, on the reactor walls. Due to the complexity of gasification of black liquor some simplifications had to be made in order to develop a CFD model for the preliminary design of the gasification reactor. Instead of modelling droplets in detail, generating gas by gasification, sources were placed in a prescribed volume where gasification (mainly drying and pyrolysis) of the black liquor droplets was assumed to occur. Source terms for the energy and momentum equations, consistent with the mass source distribution, were derived from the corresponding control volume equations by assuming a symmetric outflow of gas from the droplets and a uniform degree of conversion of reactive components in the droplets. A particle transport model was also used in order to study trajectories from droplets entering the reactor. The resulting model has been implemented in a commercial finite volume code (AEA-CFX) through customised Fortran subroutines. The advantages with this simple

  18. Fast reactor development programme in France during 1992

    International Nuclear Information System (INIS)

    The present position with respect to the development of fast reactors in France and prospects for future R and D is summarized. The paper gives an overview on the status of the fast reactors Phenix and Super Phenix. In addition to the studies in support of the EFR project, which are presented in a separate report, CEA and NOVATOME have conducted exploratory studies to evaluate the potential of fast reactors to burn plutonium and long lived wastes with the objective to maintain the acceptable values of two important parameters for safety, namely the sodium void worth and the Doppler coefficient. (author). 1 fig

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

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

  1. Analysis of fast reactor scenario with different conversion ratios

    International Nuclear Information System (INIS)

    Korean fast reactor scenarios have been analyzed for various kinds of conversion ratio by the DANESS system dynamic analysis code. The once-through fuel cycle analysis was modeled based on the Korean 'National Energy Basic Plan' up to 2030 and a postulated nuclear demand growth rate until 2150. The fast reactor scenario analysis has been performed for three kinds of conversion ratios such as 0.3, 0.61 and 1.0. Through the calculations, the nuclear reactor deployment scenario, front-end cycle, back-end cycle, and long-term heat load have been investigated. From the once-through results, it is shown that the nuclear power demand would be ∼70 GWe and the total amount of the spent fuel accumulated by 2150 would be ∼168000 t. Also, the fast reactor scenario analysis results show that the spent fuel inventory and out-pile transuranic element can be reduced by increasing the fast reactor conversion ratio. Furthermore, the long-term heat load of spent fuel decreases with increasing the conversion ratio. However, it is known that the deployment of a fast reactor of low conversion ratio does not much reduce the spent fuel and out-pile transuranic element inventory due to the fast reactor deployment limitation which is related to the availability of transuranic elements. (author)

  2. The 'RB' Reactor as a Source of Fast Neutrons

    International Nuclear Information System (INIS)

    A study of the RB reactor as possible source of fast neutrons began in 1976 and four different version of fast neutron sources are designed up to 1990: an external neutron converter - ENC (1976), an experimental fuel channel - EFC (1982), an internal neutron converter - INC (1983), and a coupled fast-thermal core - HERBE (1990). An overview of applications and characteristics of each particular source of fast neutrons, including available irradiation space, neutron spectra and equivalent neutron and gamma dose rates is presented in the paper. Control and safety-related implications of these modifications of the reactor are emphasised. Computer codes and nuclear data libraries, used in calculations, are described. (author)

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

    International Nuclear Information System (INIS)

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

  4. OECD Nuclear Energy Agency Activities Related to Fast Reactor Development

    International Nuclear Information System (INIS)

    The OECD Nuclear Energy Agency (NEA), whose role is to assist its member countries to develop, through international cooperation, the scientific and technological bases required for the safe, environmentally friendly and economical use of nuclear energy, conducts work related to fast reactor systems along two areas of activity: one focused on scientific research and technology development needs and one dedicated to strategic and policy issues. The paper summarizes recent and ongoing NEA activities in each of these areas of activity, including: improved evaluations of basic nuclear data needed for the development of fast reactor systems, expansion of integral experiments databases to provide improved validation for fast reactor modelling methods, modelling of transients in SFRs, creation of an innovative fuels expert group, a series of information exchange meetings on actinide and fission product partitioning and transmutation, study on homogeneous versus heterogeneous recycle of transuranic isotopes in fast reactors, studies on research needs and the availability of experimental facilities for fast reactor safety studies, and a study on trends towards sustainability in the nuclear fuel cycle. The NEA is also an active player in many other international activities related to fast neutron systems, such as the Generation-IV International Forum where the NEA acts as technical secretariat for the project. The NEA will continue to support member countries in the field of fast reactor development and related advanced fuel cycles by providing a forum for exchange of information and various other collaborative activities. (author)

  5. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    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.)

  6. Integral physics data for fast-reactor design

    International Nuclear Information System (INIS)

    Integral physics data for fast-reactor design. The recent compilation of the section on fast-reactor physics for the forthcoming second edition of 'Reactor Physics Constants' has necessitated a survey of the available experimental integral data. The choice of fast-reactor-physics integral data to be included in the compilation was based upon two criteria besides availability: (a) the data arise from relatively simple systems which lend themselves to simple theoretical analyses; and (b) complicated systems representing prototypes or mock-ups having general interest in terms of fast-power reactors. The first criterion was decided upon so as to list integral data for those systems of most general utility for the verification of cross-section parameters and calculational procedures. The second criterion is based upon presentation of current data on actual fast power breeder reactor systems. These are too complicated for simple theoretical analysis. They demonstrate the complexity of the actual reactor versus the more idealized and easily analysed critical experiment. Integral physics data for reactor design refer to measurements on reactor systems, critical or otherwise, of the various reactor physics quantities of practical and/or theoretical importance. These characterize and lead to an understanding of the system. The measurements are represented by critical mass, core shape factor, detector ratios, neutron spectra, material replacement experiments, reflector savings, neutron lifetime, Rossi-α, and similar quantities. These data are reviewed and the range of applicability is described. Limitations of experimental and analytical results are shown to exist in certain spectral and criticality analyses. Experimental and analytical investigations are suggested for future work. These will tend to narrow the gap between theory and experiment on 'known' systems. They also include investigations to 'firm up' the physics of large conceptual, fast power-breeder reactor

  7. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

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

    International Nuclear Information System (INIS)

    Aiming to assure the continued utilization of fission energy, the development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed, as the best type for the Brazilian nuclear system in the future. The inherent safety characteristics are superior to current fast breeder reactors and an efficient utilization of thorium can be realized. The construction and operation of an experimental fast reactor is the first step and a basic tool for the development of FBRs technologies. A serie of core design for an 90 MW FBR is studied and the possible options and sizes of the main parameters are identified. (E.G.)

  9. Recent progress of Gas Fast Reactor program

    International Nuclear Information System (INIS)

    The GFR is considered by the French Atomic Energy Commission as a promising concept which combines the benefits of fast spectrum and high temperature, using helium as coolant. He properties are interesting with respect to safety: it is single phase (no threshold effect due to phase changing), chemically inert, and non toxic. It affords an optical transparency allowing potential improvements in temperature measurement, management for dismantling, and in-service-inspection. The voiding effect is limited, less than 1$, providing quasi- decoupling of the reactor physics from the state of the coolant. Nevertheless, Helium is a poor coolant, so that the GFR viability includes development of a refractory and dense fuel, and robust management of accidental transients, especially cooling accidents. GFR feasibility is essentially linked to three demonstrations: the feasibility (fabrication, thermo-mechanical behaviour) of a refractory fuel; the safety architecture with appropriate systems for the prevention and a robust mitigation of accidental scenarios (especially depressurization); economic competitiveness. The first one includes an experimental activity at the laboratory scale: completion of the results is expected by 2012-2015. The next step afterward will be the design, construction and the operation of a 50-100 MWth experimental reactor, the Allegro project (former ETDR), possibly as a European Joint Undertaking. The full paper will recall the 2007 design choices and it will give an overview of the progress performed so far regarding the safety architecture and the safety evaluation. The 2007 reference fuel technology is a ceramic plate type fuel element. It combines a high enough core power density (minimization of the Pu inventory), plutonium and minor actinides recycling capabilities. Innovative to many aspects, the fuel element is a key issue in the GFR feasibility. It is supported already by a significant R and D effort also applicable to a pin concept that is

  10. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    The fast reactor programme in the United Kindom is reviewed under the following headings: Progress with PFR; Reprocessing: Commercial Design Studies; Structural Integrity; Engineering and Components; Materials; Sodium Chemistry; Core and Fuel; Safety; Plant Performance. (author)

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

    International Nuclear Information System (INIS)

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

  12. Advances in sodium technology, testing and diagnostics of fast reactors

    International Nuclear Information System (INIS)

    The collection contains a selection of 29 papers from three international specialists' meetings: the CMEA conference ''Control and measuring instruments and diagnostic systems of fast reactors'' held in the GDR in April 1983; the IAEA conference on nuclear power experience held in Austria in September 1982; and the conference ''Problems of technology and corrosion in sodium coolant and protective gas'' held in the GDR in April 1977. Three papers on operating experience with Soviet fast reactors and their safety have a general character; they are followed up by three papers on sodium technology. Five papers deal with the diagnostics of fast sodium cooled reactors and nine papers are devoted to the diagnostics of steam generators. Eight papers relate to detectors for the diagnostics of fast reactors. Safety regulations for work with alkali metals are added. (A.K.)

  13. Fast-breeder-power reactor records in the INIS database

    International Nuclear Information System (INIS)

    This report presents a statistical analysis of more than 19,700 records of publications concerned with research and technology in the field of fast breeder power fission reactors which are included in the INIS Bibliographic Database for the period from 1970. to 1999. The main objectives of this bibliometric study were: to make an inventory of the fast breeder power reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of fast breeder power reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in fast breeder reactors research and technology. The quantitative data in this report are obtained for various properties of relevant INIS records such as year of publication, secondary subject categories, countries of publication, language, publication types, literary types, etc. (author)

  14. Radioisotopes in the primary circuit of a fast reactor

    International Nuclear Information System (INIS)

    In the frame of the research performed to understand the behaviour of the radioactive isotopes of iodine in the primary coolant circuit of fast reactor, a simple theoretical model is proposed. Results concerning PHENIX and RAPSODIE are given

  15. A code to calculate multigroup constants for fast neutron reactor

    International Nuclear Information System (INIS)

    KQCS-2 code is a new improved version of KQCS code, which was designed to calculate multigroup constants for fast neutron reactor. The changes and improvements on KQCS are described in this paper. (author)

  16. Review of fast reactor operating experience gained in 1998 in Russia. General trends of future fast reactor development

    International Nuclear Information System (INIS)

    Review of the general state of nuclear power in Russia as for 1998 is given in brief in the paper. Results of operation of BR-10, BOR-60 and BN-600 fast reactors are presented as well as of scientific and technological escort of the BN-350 reactor. The paper outlines the current status and prospects of South-Urals and Beloyarskaya power unit projects with the BN-800 reactors. The main planned development trends on fast reactors are described concerning both new projects and R and D works. (author)

  17. Operating experience from the BN600 sodium fast reactor

    International Nuclear Information System (INIS)

    Conclusion: The operating experience from the BN600 reactor power unit for more than 32 years is positive in terms of the demonstration of the feasibility of the utilization of a sodium-cooled fast reactor for commercial electric generation. The BN600 reactor is an important key link ensuring the continuity and succession of the development of the fast reactors in Russia of which the reliable and steady operation confirms good prospects of this line of the nuclear power industry. In the course of the BN600 power unit operation the valuable operating experience from the individual systems and components which should be preserved and utilized when developing the advanced designs of the sodium-cooled fast reactors was accumulated

  18. New modelling method for fast reactor neutronic behaviours analysis

    International Nuclear Information System (INIS)

    Due to safety rules running on fourth generation reactors' core development, neutronics simulation tools have to be as accurate as never before. First part of this report enumerates every step of fast reactor's neutronics simulation implemented in current reference code: ECCO. Considering the field of fast reactors that meet criteria of fourth generation, ability of models to describe self-shielding phenomenon, to simulate neutrons leakage in a lattice of fuel assemblies and to produce representative macroscopic sections is evaluated. The second part of this thesis is dedicated to the simulation of fast reactors' core with steel reflector. These require the development of advanced methods of condensation and homogenization. Several methods are proposed and compared on a typical case: the ZONA2B core of MASURCA reactor. (author)

  19. Shuffling strategy study of breeding-burning integrated fast reactor

    International Nuclear Information System (INIS)

    The breeding-burning integrated fast reactor uses burning assemblies to generate thermal power, meanwhile, converts 238U into 239Pu in the fertile assemblies. With periodical shuffling of assemblies, the reactor can maintain criticality for decades of years. To maintain long-term stability of the core reactivity, the core layout and shuffling strategy should balance the burning and the breeding of the assemblies. The scattered core layout and shuffling strategy ensures fast breeding of the fertile assemblies, and keeps stable core power distribution in whole life of the reactor. Moreover, at the end of the reactor life, the discharge burnups of different fuel assemblies are close to each other, which are about 250300 GW · d/t. This is important for breeding-burning integrated fast reactor to achieve very efficient utilization of uranium resource without reprocessing. (authors)

  20. Recycle Strategies for Fast Reactors and Related Fuel Cycle Technologies

    International Nuclear Information System (INIS)

    Fast reactors and related fuel cycle (hereafter referred to as 'fast reactor cycle') technologies have the potential to contribute to long term energy security owing to their effective use of uranium and plutonium resources, and to a reduction in the heat generation and potential toxicity of high level radioactive wastes by burning long lived minor actinides recovered from spent fuel from light water reactors and fast reactors. Further, it is likely that fast reactor cycle technologies can play a certain role in non-proliferation as addressed in the Global Nuclear Energy Partnership. With these features, the research and development towards their commercialization has been promoted vigorously and globally as a future vision of nuclear energy. The introduction of fast reactor cycle systems will be carried out independently in each country according to its national conditions and nuclear energy policy. It should then be considered important to have a globally common consensus relating to safety philosophy, concepts of proliferation resistance, transuranic element burnup and recycling and so on. For the development and utilization of fast reactor cycle systems, while respecting each country's concept, it is essential to organize the technologies and concepts which countires should have in common globally and build a framework to make them standardized. The use of existing frameworks such as the Generation IV International Forum and the International Project on Innovative Nuclear Reactors and Fuel Cycles is considered effective to achieving this. Furthermore, a vigorous promotion such as international cooperative developments enables the formation of international consensus on major technologies for the fast reactor cycle as well as the saving of resources by infrastructure sharing. (author)

  1. Fast Neutron Detector for Fusion Reactor KSTAR Using Stilbene Scintillator

    OpenAIRE

    Lee, Seung Kyu; Kang, Byoung-Hwi; Kim, Gi-Dong; Kim, Yong-Kyun

    2011-01-01

    Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spe...

  2. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    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

  3. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  4. Improve Design of Fuel Shear for Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    GAO; Wei; OUYANG; Ying-gen; LI; Wei-min

    2012-01-01

    <正>Due to the deeper burnup and higher fuel swelling, fast reactor metal fuel rod using 316 stainless steel cladding, replacing the traditional zirconia cladding. The diameter of fuel rod of fast reactor is much longer than that of PWR, and the cladding of stainless steel has better ductility than zirconia cladding. Using the existing shear still will cause several aspects of problem: 1) Longer diameter of rod leads to

  5. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

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

  6. A review of calculation methods for fast and intermediate reactors

    International Nuclear Information System (INIS)

    This paper discusses the development of methods for calculating intermediate and fast reactors. It deals with various approaches to the problems of physical calculation. The calculation of resonance effects is discussed. Consideration is given to multi-group systems of fundamental and conjugate equations, various applications of perturbation theory to the problems of physical reactor calculation, and numerical methods of solving fundamental and conjugate reactor equations, which approximate the method of spherical harmonics. The paper describes an application of the response method to the solution of critical-mass problems, and methods of calculating reactors with hydrogeneous moderators. The fundamental features of an effective one-group reactor model are described. (author)

  7. High burnup fast reactor fuel: processing and waste management experiences

    International Nuclear Information System (INIS)

    The routine processing of mixed Plutonium/Uranium oxide fuels from the Prototype Fast Reactor (PFR) at Dounreay began in September 1980 and the design features of the modified Dounreay Fast Reactor (DFR) reprocessing plant and experience of the first active campaign were described in a paper to the British Nuclear Engineering Society in November 1981 (1). Since then progress in processing the fuel discharged from PFR has been covered briefly in a number of papers to international conferences and the Public Inquiry held in 1986 into the outline planning application for the proposed European Demonstration Reprocessing Plant. During this decade considerable experience in the operation of fast reactors and associated fuel plants has been accumulated providing confidence in the system before entering the next development phase - that of its commercial demonstration. Confidence in the UK draws on the successful operation of the PFR and the associated Dounreay fuel reprocessing and BNF Sellafield fabrication plants. Of equal importance is public confidence in safe operation and in the management of wastes generated by a fast reactor system. The present paper is a review of fast reactor reprocessing and waste management at the Dounreay Nuclear Establishment (DNE) as a contribution to the present status of the fast reactor system

  8. Methods for quantifying uncertainty in fast reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H.; Fischer, P. F.

    2008-04-07

    Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.

  9. Preliminary Conceptual Design for a Multipurpose Experimental Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    A conceptual nuclear design was done for a 100MWe multi-purpose sodium cooled fast reactor core for both research and experiment as a pre-feasibility study. Challenging goals for a small sized core were set up. Cycle length should be longer than 4 month with low enriched uranium-only fuels in a U-Zr metal alloy form. Overall design features are chosen from previous design concept - advanced burner test reactor (ABTR). Number of fuel assemblies and reflector assemblies were increased from 54 to 66 and from 78 to 207 respectively, resulting in increase of core radius to 180cm. In the reflector zone, there was enough space to install three independent fuel test loops in which different type of coolants and fuels might be tested. Fuel pin size was selected to be 0.737cm through a parametric study with database for U-Zr metal fuels tested in EBR-II. Core height was optimized to be 87cm after a parametric study on core size. Target flux levels at both fuel test assemblies and material test assemblies were 3.0×1015n/cm2-sec. Most of performance parameters as well as safety parameters were satisfied with goals except a flux level at the MTA. Overall safety aspects of the reactor is better than any other previous SFR design concepts

  10. Plutonium disposition in the BN-600 fast-neutron reactor at the Beloyarsk nuclear power plant

    Science.gov (United States)

    Moses, D. L.; Chebeskov, A. N.; Matveev, V. I.; Vasiliev, B. A.; Maltsev, V. V.

    In 1996, the United States and the Russian Federation completed an initial joint study that evaluated the candidate options for the disposition of surplus weapons-derived plutonium in both countries. While Russia advocates building new reactors for converting weapons-derived plutonium to spent fuel, the cost is high, and the continuing joint study of the Russian options is considering only the use of the existing VVER-1000 LWRs in Russia (and possibly in Ukraine) and the existing BN-600 fast-neutron reactor at the Beloyarsk Nuclear Power Plant in Russia. The BN-600 reactor, which currently uses enriched uranium fuel, is capable with certain design modifications of converting up to 1.3 metric tons (MT) of surplus weapons-derived plutonium to spent fuel each year. The steps needed to convert BN-600 to a plutonium-burner core will be discussed. The step involving the hybrid core allows an early and timely start that takes advantage of the limited capacity for fabricating uranium-plutonium mixed-oxide fuel early in the disposition program. The design lifetime of BN-600 must safely and reliably be extended by 10 yr to at least 2020 so that a sufficient amount of plutonium (˜20 MT) can be converted to spent fuel.

  11. Plutonium disposition in the BN-600 fast-neutron reactor at the Beloyarsk nuclear power plant

    International Nuclear Information System (INIS)

    In 1996, the United States and the Russian Federation completed an initial joint study that evaluated the candidate options for the disposition of surplus weapons-derived plutonium in both countries. While Russia advocates building new reactors for converting weapons-derived plutonium to spent fuel, the cost is high, and the continuing joint study of the Russian options is considering only the use of the existing VVER-1000 LWRs in Russia (and possibly in Ukraine) and the existing BN-600 fast-neutron reactor at the Beloyarsk nuclear power plant in Russia. The BN-600 reactor, which currently uses enriched uranium fuel, is capable with certain design modifications of converting up to 1.3 metric tons (MT) of surplus weapons-derived plutonium to spent fuel each year. The steps needed to convert BN-600 to a plutonium-burner core will be discussed. The step involving the hybrid core allows an early and timely start that takes advantage of the limited capacity for fabricating uranium-plutonium mixed-oxide fuel early in the disposition program. The design lifetime of BN-600 must safely and reliably be extended by 10 yr to at least 2020 so that a sufficient amount of plutonium (∝20 MT) can be converted to spent fuel. (orig.)

  12. Fast Pyrolysis of Lignin Using a Pyrolysis Centrifuge Reactor

    DEFF Research Database (Denmark)

    Trinh, Ngoc Trung; Jensen, Peter Arendt; Sárossy, Zsuzsa;

    2013-01-01

    Fast pyrolysis of lignin from an ethanol plant was investigated on a lab scale pyrolysis centrifuge reactor (PCR) with respect to pyrolysis temperature, reactor gas residence time, and feed rate. A maximal organic oil yield of 34 wt % dry basis (db) (bio-oil yield of 43 wt % db) is obtained at...

  13. Progress of Research on Demonstration Fast Reactor Main Pipe Material

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The main characteristics of the sodium pipe system in demonstration fast reactor are high-temperature, thin-wall and big-caliber, which is different from the high-pressure and thick-wall of the pressurized water reactor system, and the system is long-term

  14. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  15. Some policy aspects of the fast reactor question. Chapter 2

    International Nuclear Information System (INIS)

    The following aspects of energy policy in the UK are discussed: planning and forecasting, accuracy and relevance to government policies; economics; plant construction programmes; scope for electricity growth; arguments for and against fast reactor programme (and in relation to other types of reactor). The general discussions of energy policy cover coal, natural gas and oil in addition to nuclear power. (U.K.)

  16. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

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

  17. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  18. CP ESFR: Collaborative Project for a European Sodium Fast Reactor

    International Nuclear Information System (INIS)

    The Collaborative Project for a European Sodium Fast Reactor (CP ESFR) is performed (2009-2012) in the 7th European Framework Programme. It is devoted to the identification and study of innovations to be considered for the future in the core design, safety, reactor architecture, components and the dissemination of knowledge related to this technology among young European professionals. (author)

  19. Track 5: safety in engineering, construction, operations, and maintenance. Reactor physics design, validation, and operating experience. 5. A Negative Reactivity Feedback Device for Actinide Burner Cores

    International Nuclear Information System (INIS)

    per atmosphere increase in pressure. 4. This lifts the floats higher into the core above their equilibrium position at hot full power. 5. The increased neutron absorption produces a negative reactivity feedback. 6. The surrounding primary coolant keeps all boundaries at nearly constant temperature. The ex-core helium has very low energy absorption, plus good heat transfer, which helps maintain constant temperature and pressure. The neutron absorber floats are thin metal tubes that contain a rhenium slug, as a high-capture cross-section ballast, and an upper section of 10B4C pellets. The tops and bottoms of the floats are rounded to guard against sticking inside the riser tubes. The top of the float is vented through a porous disk into the cool helium plenum to allow the helium produced in 10B capture to escape. The absorber float is cooled by conduction through the LBE bath, and guide-tube wall, into the ambient LBE primary coolant. Whole-core Monte Carlo calculations for RFDs substituted for the central void tube in 20% of the streaming fuel assemblies proposed for actinide burner cores in Ref. 1 indicate a steady- state reactivity power feedback coefficient exceeding -1 c/% power, which is better than that of sodium-cooled integral fast reactor (IFR)-type cores (at approximately 20.5 c/%) and about half of that of oxide-fueled fast breeder reactors (FBRs). However, the RFD feedback is considerably slower following a step power increase: Preliminary estimates suggest a factor of 5 slower than the oxide fuel Doppler reactivity insertion rate. Nevertheless, this may be adequate since the reactors in question can be designed to have no obvious large, rapid reactivity insertion accidents to cope with. Much remains to be done to refine and optimize this concept. Among necessary evaluations are seismic response, the consequences of gas plenum failure, and reactivity insertion by the automatic RFD withdrawal following a power reduction, safety scram in particular. Various

  20. Recycle strategies for fast reactors and related fuel cycle technologies

    International Nuclear Information System (INIS)

    Full text: 1. Introduction Fast reactors and related fuel cycle (hereinafter referred to as 'Fast reactor cycle') technologies have the potential of contributing to long-term energy security due to effective use of uranium and plutonium resources, and reduction of the heat generation and potential toxicity of high-level radioactive wastes by burning long-lived minor actinides (MA) recovered from spent fuels of light-water reactors and fast reactors. Further, it is likely that fast reactor cycle technologies can play a certain role in non- proliferation as addressed in GNEP (Global Nuclear Energy Partnership). With these features, R and Ds toward their commercialization have been promoted vigorously and globally as a future vision of nuclear energy. 2. Recycle strategies in each country In Japan, it is determined that after burning uranium in light water reactors, plutonium is recovered from spent fuel and used for light water reactors at the moment and for fast reactors in the future. In order to make it possible, Fast Reactor Cycle Technology Development (FaCT) Project has been promoted with a combination of oxide-fueled sodium-cooled reactors, advanced aqueous reprocessing, and simplified pelletizing fuel fabrication adopted as a main concept aiming at startup of a demonstration reactor around 2025 and commercialization before around 2050. In France, a comparison of the basic specifications between an oxide-fueled sodium-cooled reactor and a carbide (or nitride)-fueled gas-cooled reactor has currently been promoted towards technological selection for a prototype reactor in 2012 in accordance with 'The 2006 planning act on the sustainable management of radioactive materials and waste (Act 2006- 739)' enacted in 2006. Based on the results, France aims at startup of the prototype reactor in 2020 and commercialization in around 2040. For reprocessing, methods which extract actinides collectively such as GANEX has been developed to enhance proliferation resistance

  1. Current status on fast reactor program in Kazakhstan

    International Nuclear Information System (INIS)

    Atomic scientific-industrial complex of Republic of Kazakhstan consist of: Uranium mining, production and power industry, Enterprises of uranium ores geological searching and number of natural mines (using the mining and underground leaching techniques); Two plants of U3O8 production at Aktau and Stepnogorsk towns; Metallurgical plant producing uranium fuel pellets for fuel assemblies of RBMK and VVER reactors types; Energy plant at Aktau (MAEK) is used for production of heat, electricity and desalination of water and based on three energy blocks using natural gas and one nuclear unit with fast breeder reactor BN-350. The fast breeder reactor BN-350 at Aktau was commissioned in November 1972 and finally shutdown in April 1999. Three different type of the research reactors and non reactor test facility on the territory of the former Semipalatinsk Nuclear Test Site and one research reactor and subcritical assembly nearly Almaty are exploiting for the investigation in field of reactors nuclear safety and other type of investigations. These are: VVR-K - light water reactor, power - 10 MW, EWG-1M - thermal light water heterogeneous vessel reactor with light water moderator and coolant, beryllium reflector, maximum thermal power - 35 MW, IGR - impulse homogeneous uranium-graphite thermal neutron reactor with graphite reflector, RA - thermal neutron high temperature gas heterogeneous reactor with air coolant, zirconium hydride moderator, and beryllium reflector, about 0.5 MW power, EAGLE - non reactor test facility for reactor fuel element melt process due to severe accident studding. Project on construction of experimental reactor TOKOMAK at city Kurchatov (in frame of International Thermonuclear Experimental Reactor) is going on (design and equipment manufacture and procurement stage). Accomplishment of the project is estimated for year 2007. Works on construction of the new cyclotron at Astana University started at the beginning of this year in co-operation with Dubna

  2. Fast Reactor Physics. Vol. II. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

    International Nuclear Information System (INIS)

    Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  3. Fast Reactor Physics Vol. I. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

    International Nuclear Information System (INIS)

    Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  4. A Review of Fast Reactor Progress in Japan

    International Nuclear Information System (INIS)

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1980 through March 1981, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1979. The 1980 year budget for R&D work and for construction of a prototype fast breeder reactor, MONJU, will be approximately 14 and 19 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, JOYO, power increase from 50 MWt to 75 MWt was made in July 1979 and an operational cycle at 75 MWt has been completed very recently. With respect to the prototype reactor MONJU, progress toward construction has been made and an environmental impact statement of the reactor is being reviewed by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extraporating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized below

  5. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report

  6. Fast reactors: the future of nuclear energy

    International Nuclear Information System (INIS)

    The main problems to be solved for FBR type reactors become viable economically, presenting the research programs of Europe, United States of America, Japan and Brazil are described. The cooperations between interested countries for improving FBR type reactors, and the financial and human resources necessaries for the development of programs, are evaluated. The fuel cycle is also analysed. (M.C.K.)

  7. What can fast breeders do for Ontario

    International Nuclear Information System (INIS)

    Fast reactors have the potential of significantly reducing Ontario's demand for natural resources while meeting virtually any requirements for nuclear power this province may have. The breeding efficiency of the fast reactors does not affect the overall uranium consumption of the system to any significant extent. It is, however, an important economic factor in a breeder/burner system. To minimize the resource consumption, the fast reactors should be introduced in Ontario at the onset of the next century. The 'breeder-burner' mix of reactors can effectively reduce the fissile inventory of the whole power system (including the inventory in irradiated fuel storage bays). For the nuclear capacity growth scenarios thought to be applicable in Ontario, the fast reactor systems have about the same or lower requirements for natural uranium as the best (self-sustaining thorium) CANDU cycles. Compared to all other advanced CANDU cycles, the fast reactors yield a substantial resource saving. (auth)

  8. Status of national programmes on fast reactors in Korea

    International Nuclear Information System (INIS)

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

  9. Review of the United Kingdom fast reactor programme - March 1986

    International Nuclear Information System (INIS)

    The UK programme in the field of fast reactors has continued successfully towards the following main objectives, details of which are contained in subsequent sections of this report: (2) progress with the prototype fast reactor (PFR) which achieved its design power on 4 March 1985; (3) nuclear fuel reprocessing; (4) commercial design studies; (5) structural integrity of LMFBR during its lifetime; (6) R and D work on components of LMFBR; (7) materials study; (8) sodium chemistry; (9) reactor core and fuel design philosophy; (10) safety problems; (11) plant performance studies

  10. Neutronic design of a plutonium-thorium burner small nuclear reactor

    International Nuclear Information System (INIS)

    A small nuclear reactor using thorium and plutonium fuel has been designed from the neutronic point of view. The thermal power of the reactor is 150 MWth and it is proposed to be used to supply electricity in an island in Indonesia. Thorium and plutonium fuel was chosen because in recent years the thorium fuel cycle is one of the promising ways to deal with the increasing number of plutonium stockpiles, either from the utilization of uranium fuel cycle or from nuclear weapon dismantling. A mixed fuel of thorium and plutonium will not generate the second generation of plutonium which will be a better way to incinerate the excess plutonium compared with the MOX fuel. Three kinds of plutonium grades which are the reactor grade (RG), weapon grade (WG), and spent fuel grade (SFG) plutonium, were evaluated as the thorium fuel mixture in the 17x17 Westinghouse PWR Fuel assembly. The evaluated parameters were the multiplication factor, plutonium depletion, fissile buildup, neutron spectrum, and temperature reactivity feedback. An optimization was also done to increase the plutonium depletion by changing the Moderator to Fuel Ratio (MFR). The computer codes TRITON (coupled NEWT and ORIGEN-S) in SCALE version 6 were used as the calculation tool for this assembly level. From the evaluation and optimization of the fuel assembly, the whole core was designed. The core was consisted of 2 types of thorium fuel with different plutonium grade and it followed the checkerboard loading pattern. A new concept of enriched burnable poison was also introduced to the core. The core life is 6.4 EFPY or 75 GWd/MTHM. It can burn up to 58% of its total mass of initial plutonium. VENTURE was used as the calculation tool for the core level

  11. Fast neutron benchmark proposal at TRIGA-ACPR Reactor

    International Nuclear Information System (INIS)

    The development of fast neutron benchmarks is a historical aim of reactor physics. The dry experimental tube situated in the central region of the core in TRIGA Annular-Core Pulsing Reactor (ACPR) offers a suitable neutron source for fast neutron benchmark development. Our proposal consists in mounting a high-enriched uranium annular converter into the dry channel of the core. Preliminary computations and measurements are presented in this paper. Neutron flux computations in the dry channel and the uranium converter were performed using MCNP and WIMS codes. Also neutron flux spectrum measurements and fast and thermal neutron flux distribution measurements were performed using foil activation techniques. (authors)

  12. The case for the gas cooled fast reactor

    International Nuclear Information System (INIS)

    Although gas-cooling for fast reactors had been the subject of consideration since the early days of nuclear power, it was when the concept of the prestressed concrete pressure vessel turned into practical fact, that convincing arguments could be made to overcome safety objections. In terms of hardware, the Gas Cooled Fast Breeder Reactor can rely on existing and available technologies; as far as fuel is concerned, valuable information will be derived from the Liquid Metal Fast Breeder Reactor programme. The GCFR can be made very flexible; its capital cost will not exceed by more than 20% the one for reactor built at present on commercial scale; the overall economy of its fuel cycle is good. It could play an important role in the future breeder family

  13. Fast-power-reactor optimization by the game theory

    International Nuclear Information System (INIS)

    In the first stage of the use of fast breeder reactor - because fissile-material amounts are small - we are interested in fast breeder reactors which achieve minimum fissile-material mass, with maximum power. This problem shows a two-matrix-game structure. First, we determine a competive-game solution and second, a cooperative-game solution, obtaining in this way the optimum distribution of the fissile and fertile materials in the multizone fast reactors. Another optimization problem which is solved in this paper is finding the reactor structure for which the power non-uniformity factor and the flux non-uniformity factor are minimum. This is, also, a mathematical two-matrix game and it is solved as above. The two optimization problems have different solutions. (author)

  14. Fast neutron reactors: a long experience facing a new future

    International Nuclear Information System (INIS)

    This article makes a survey of the different fast reactor programs throughout the world. The European fast reactor project (EFR) was launched in 1988 under the impulse of a partnership involving electricity producers, nuclear core manufacturers and research agencies. The fading prospect of an energy shortage has led to the freeze of EFR project by the end of 1998. In Europe most fast reactor programs have entered a waiting period whose duration could reach decades. Nevertheless the necessity of assuring the capability of designing and building such reactors stays a priority in order to benefit from the property of breeding which could strongly contribute to the future energetic autonomy of Europe. The improvement of technical performances, the integration of technological progress, the investigation of new concepts are the main tasks of the waiting period. Some prototypes will have to be built at regular time intervals in order to assure the feasibility of these evolutions. (A.C.)

  15. Simulator platform for fast reactor operation and safety technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  17. Fast Reactor Programme. Second Quarter 1969. Progress Report. RCN Report

    International Nuclear Information System (INIS)

    This progress report covers fast reactor research carried out by RCN during the second quarter 1969 forming part of the integrated fast breeder research and development programme also in progress at the national nuclear research centres Karlsruhe and Mol. The combined effort is based on a memorandum of co-operation in this field signed by the respective governments in 1968 and on a memorandum of understanding signed by the research centres. The RCN contribution is mainly concerned with the core of the fast breeder reactor and related safety aspects and, as such, must be looked upon as being complementary to the industrial research pro- field of fast reactors. The contribution comprises the following six items: - A Æéatîtôr , physics programme to determine the influence of fission products on several main characteristics of the reactor core such as void coefficient, Doppler coefficient and breeding ratio; - A fuel performance programme in which both stationary and transient irradiations are being carried out to establish the temperature and power limits of fuel rods; also the consequences of loss- of-cooling will be investigated; - Investigation into the change in mechanical properties of fuel canning materials due to high fast neutron doses; - A study of the corrosion behaviour of canning materials and their compatibility with the fuel under conditions of high temperature and high pressure; - Investigation into the behaviour of aerosols of fission products which could be formed after a fast reactor accident; a thorough understanding is of utmost importance for the reactor safety assessment ; - Studies on heat transfer in the reactor core. As fast breeders operate at high power densities, an accurate knowledge on the heat transfer phenomena is required

  18. OECD Nuclear Energy Agency Activities Related to Fast Reactor Development

    International Nuclear Information System (INIS)

    The OECD Nuclear Energy Agency (NEA), whose role is to assist its member countries to develop, through international cooperation, the scientific and technological bases required for the safe, environmentally friendly and economical use of nuclear energy, conducts work related to fast reactor systems in two areas of activity: one focused on scientific research and technology development needs and one dedicated to strategic and policy issues. Recent, scientifically oriented, fast reactor related activities coordinated by the NEA comprise: -A coordinated effort to evaluate basic nuclear data needed for the development of fast reactor systems; -A recently initiated review of Integral Experiments for Minor Actinide Management; -An ongoing study on Homogeneous versus Heterogeneous Recycle of Transuranic Isotopes in Fast Reactors; -A comparative analysis of the safety characteristics of sodium cooled fast reactors; -A series of workshops on Advanced Reactors with Innovative Fuels; -A series of information exchange meetings on actinide and fission product partitioning and transmutation. The NEA has also conducted two reviews on issues related to the transition from thermal to fast neutron nuclear systems. One study was devoted to technical issues, including benchmark studies on: (i) the performance of scenario analysis codes, (ii) a regional (European) scenario and (iii) a global transition scenario. The other study emphasized issues of interest to policymakers, such as key parameters affecting the cost-benefit analysis of transitioning, including the size and age of the nuclear reactor fleet, the expected future reliance on nuclear energy, access to uranium resources, domestic nuclear infrastructure and technology development, and radioactive waste management policy in place. The NEA is also an active player in many other international activities related to fast neutron systems, such as the Generation IV International Forum, where the NEA acts as technical secretariat for

  19. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

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

  20. Assumed mode approach to fast reactor core seismic analysis

    International Nuclear Information System (INIS)

    The need for a time history approach, rather than a response spectrum approach, to the seismic analysis of fast breeder reactor core structures is described. The use of a Rayleigh-Ritz/Assumed Mode formalism for developing mathematical models of reactor cores is presented. Various factors including structural nonlinearity, fluid inertia, and impact which necessitate abandonment of response spectrum methods are discussed. The use of the assumed mode formalism is described in some detail as it applies to reactor core seismic analysis. To illustrate the use of this formal approach to mathematical modeling, a sample reactor problem with increasing complexities of modeling is presented. Finally, several problem areas--fluid inertia, fluid damping, coulomb friction, impact, and modal choice--are discussed with emphasis on research needs for use in fast reactor seismic analysis

  1. 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)

    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

  2. A Review of the UK Fast Reactor Programme: March 1980

    International Nuclear Information System (INIS)

    Towards the end of 1979 the Government announced a new programme of thermal reactor stations to be built over ten years (totalling 15GW), in addition to the two AGR stations at Torness and Heysham 'B' which had been approved by the previous Government. The first station of the new programme will be based on a Westinghouse PWR, subject to safety clearance and the outcome of a public inquiry, and it is envisaged that the remaining stations of the programme would be split between PWRs and AGRs. The AEA Chairman wrote formally to the Secretary of State for Energy in December 1979, putting forward on behalf of the Electricity Supply Authorities, NNC, BNFL and the AEA a recommended strategy for building the Commercial Demonstration Fast Reactor (CDFR), subject to normal licensing procedure and to public inquiry, so as to ensure that the key options for introducing commercial fast reactors, when required, should remain open. A Government statement is expected during the next few months. Meanwhile the level of effort on fast reactor research and development in the UK has been maintained, the fast reactor remaining the largest of the UKAEA's reactor development projects with expenditure totalling somewhat over £80M per annum. The main feature of the UK fast reactor programme has continued to be the operation of PFR (Sections 2 and 7) which is yielding a wealth of experience and of information relevant to the design of commercial fast reactors. Bum-up of standard driver fuel has reached 6-7% by heavy atoms, while specially enriched lead fuel pins have reached 11 % without failure. An extensive programme of work in the reactor and its associated steam plant was completed in March 1980 and the reactor then started its fifth power run. The fuel reprocessing plant at DNE is being commissioned and has reprocessed some of the spent fuel remaining from the DFR. It will start soon on reprocessing fuel discharged from the PFR. During the year improvements to the design of the future

  3. Fast breeder reactor. The past, the present and the future. (7) History of fast reactor development in Japan - 2

    International Nuclear Information System (INIS)

    History and present state of fast breeder reactor was reviewed in series. As a history of fast reactor development in Japan - 2, this seventh lecture presented the development of the prototype FBR (MONJU) and design studies of the demonstration reactor. The MONJU started operation in 1994, but a sodium leakage in its secondary heat transfer system occurred during performance tests in 1995. It has not operated since and activities for restart are conducted. Since 1997 design studies of the demonstration FBR have been conducted to reflect the MONJU sodium leakage accident and also establish its economic competitiveness with advanced LWR. (T. Tanaka)

  4. Transmutation of /sup 90/Sr and /sup 137/Cs in a high-flux fast reactor with a thermalized central region

    Energy Technology Data Exchange (ETDEWEB)

    Taube, M.

    1976-10-01

    The fission products /sup 90/Sr and /sup 137/Cs produced by fission reactors of 30 GW(th) can be transmutated into stable nuclides by neutron irradiation with a thermal flux of 2 x 10/sup 16/ n cm/sup -2/ s/sup -1/. The rates of transmutation are 15 and 3.3 times greater, respectively, than that of spontaneous beta decay. The transmutation would take place in a central thermalized region of a high-flux fast burner reactor of 7 GW(th). In the case where the power reactors of 23 GW(th) are breeders with a high breeding gain of G = 0.38, the total system, inclusive of the high-flux burner, remains a breeding system, with G/sub total/ = 0.09. Details of the neutronics calculations and simplified thermohydraulics are given. The high-flux burner is fueled with a molten salt of chlorides of plutonium and sodium with a power density of 10 kW cm/sup -3/. The ''self-liquidation'' of such a system is discussed.

  5. Current status on fast reactor program in Kazakhstan

    International Nuclear Information System (INIS)

    Atomic scientific-industrial complex of Kazakhstan deals with uranium mining, production and power industry. The three types of research reactors existing are used for research mostly in the field reactor safety. The BN-350 reactor is being decommissioned within the frame of Kazakhstan-American project reviewed by IAEA. The EAGLE project is carried out dealing with design of an extremely safe fast reactor. Kazakstan is taking part in IAEA CRP on 'Fission product yield data required for transmutation of minor actinide nuclear waste'

  6. Status of fast reactor activities in the Russian Federation

    International Nuclear Information System (INIS)

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

  7. Towards the thorium fuel cycle with molten salt fast reactors

    International Nuclear Information System (INIS)

    Highlights: • Neutronic calculations for fast spectrum molten salt reactor. • Evaluation of the fissile matter to be used in such reactor as initial fissile load. • Capabilities to transmute transuranic elements. • Deployment scenarios of the Thorium fuel cycle. • Waste management optimization with molten salt fast reactor. - Abstract: There is currently a renewed interest in molten salt reactors, due to recent conceptual developments on fast neutron spectrum molten salt reactors (MSFRs) using fluoride salts. It has been recognized as a long term alternative to solid-fueled fast neutron systems with a unique potential (large negative temperature and void coefficients, lower fissile inventory, no initial criticality reserve, simplified fuel cycle, wastes reduction etc.) and is thus one of the reference reactors of the Generation IV International Forum. In the MSFR, the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this characteristic, the MSFR can operate with widely varying fuel compositions, so that the MSFR concept may use as initial fissile load, 233U or enriched uranium or also the transuranic elements currently produced by light water reactors. This paper addresses the characteristics of these different launching modes of the MSFR and the Thorium fuel cycle, in terms of safety, proliferation, breeding, and deployment capacities of these reactor configurations. To illustrate the deployment capacities of the MSFR concept, a French nuclear deployment scenario is finally presented, demonstrating that launching the Thorium fuel cycle is easily feasible while closing the current fuel cycle and optimizing the long-term waste management via stockpile incineration in MSRs

  8. A fast breeder reactor development scheme for Brazil

    International Nuclear Information System (INIS)

    Fast breeder reactors will be necessary in the next century in order to meet increasing demands for electricity resulting from industrialization and general improvement of standards of living. A scheme for the development of liquid metal fast breeder reactors in Brazil is proposed. Emphasis are placed on reactor safety in order to promote public acceptance, on utilization of thorium that is abundant in the country, and on consistency and smoothness of the development. The initial step is the construction and operation of a 5 MW experimental fast reactor in order to acquire basic experiences and technologies. The second step is the construction of a series of small power plants which should assure a ssound technological development. The reactor is designed with particular emphasis on safety and ease of operation. Demonstration of safety and reliability with small units would enhance public acceptance. In the final phase, when fast breeder reactors are to play a central role in electricity generation, large power plants that utilize both uranium and thorium fuel cycles will be built to establish a practically permanent power system. (Author)

  9. Plutonium utilization in thermal and fast reactors in Japan

    International Nuclear Information System (INIS)

    Nuclear power development in Japan is rather extensive, and the installed nuclear power capacity is predicted to reach 49,000MW(e) by 1985 and possibly 170,000MW(e) by 2000. Currently installed nuclear power is mainly based on the light-water reactor, and this trend is expected to persist for the time being. Plutonium produced by the LWR will reach 20t by 1985 and more than 200t by 2000. Should this plutonium be simply stored, it will cause economic pressure on utilities and the management, as well as physical protection problems associated with plutonium storing. Three ways of solving these problems are being worked out, the best solution being to use plutonium in fast reactors. To achieve this, an experimental fast reactor, JOYO, has been constructed and reached criticality in April 1977. A prototype fast breeder reactor, MONJU, designed for about 300MW(e), is nearing the final stages of design work. Its construction will commence in a few years. A demonstration fast breeder reactor will come after MONJU and the large-scale commercial use of a fast breeder reactor is expected around 1995. To meet the imminent need for plutonium utilization, two technologies, which are equally important to Japan, are currently being developed. One is the recycle use of plutonium into LWR, a technology which has long been jointly developed by research organizations and utilities. The other is to burn plutonium in an advanced thermal reactor (D2O-moderated, boiling-water cooled). The 160-MW(e) FUGEN is a prototype of this power reactor, and is almost finished. (author)

  10. Knowledge management in fast reactors and related fuel cycles

    International Nuclear Information System (INIS)

    Full text: The 21st century is ushering in a new phase of economic and social development which can be referred as 'Knowledge Economy', in which knowledge has become the key asset in determining the organization's success or failure. The IAEA defines knowledge management as an integrated, systematic approach to identify, manage and share an organization's knowledge collectively in order to help achieve the objectives of the organization. Nuclear technology is very complex and a highly technical endeavor. It relies on innovative creation, storage and dissemination of knowledge. The nuclear energy is characterized by long time scales and technological excellence. Nuclear knowledge management is a critical input to nuclear power industry, the associated fuel cycle activities and nuclear applications in medicine, industry and agriculture. Realizing the importance of knowledge preservation in the area of fast reactor technology, IAEA had given a consultancy work to Argonne National Laboratory to study and suggest the means of knowledge management. The IAEA initiative seeks to establish a comprehensive inventory of fast reactor data and knowledge for the fast reactor development in the coming years. It was suggested that the knowledge regarding important disciplines like fuels and materials, reactor physics and core design, operations, the demonstration of safety should be preserved. Various countries have initiated the fast reactor knowledge preservation activities. In France, CEA, EDF and Framatome ANP have initiated liquid metal cooled fast reactor knowledge preservation project that deals with R and D aspects and Superphenix design. European Fast Reactor collaboration (MASURCA,SNEAK,ZEBRA) has preserved the zero power critical experimental data in the SNEDAX database. Japan has started a comprehensive knowledge preservation program including the capture of 'Human Knowledge' based on interviews. In Russia steps are initiated to preserve fast reactor knowledge

  11. Cross section weighting spectrum for fast reactor analysis

    International Nuclear Information System (INIS)

    Preparation of a nuclear data library is the first task that a reactor analyst needs to perform a neutronic analysis of a reactor type. Today, in fast reactor area, the scheme used to generate this library includes the processing of an evaluated nuclear data file to obtain cross sections, in thousands of groups. Sequentially, the nuclear data are processed by a cell code to obtain neutron flux that is used to condense the large amount of energy groups to a practical calculation number of groups that can be used in reactor analysis. In the first step of the scheme it is necessary a weighting spectrum to generate the nuclear data. Here, it is proposed to use the flux estimated by Monte Carlo code using cell or the exact geometries and actual composition of the problem to obtain the main portion of the weighting spectrum instead of a code built-in function. As an example, it is presented the differences between selected pins spectrums obtained with MCNP5 calculation of a hexagonal fast reactor fuel assembly. Also, it is showed a comparison between these spectra and the one obtained in the representative unit-cell model of this fuel assembly. The comparisons support that the proposed procedure, problem dependent, may be more accurate and a good choice to generate weighting spectrum in ultra-fine energy structure for fast reactor analysis. The proposed method will be used in space reactor neutronic analysis. (author)

  12. Maintenance equipment for a fast reactor

    International Nuclear Information System (INIS)

    Object: To permit cleaning of through-holes of various diameters by mounting a spray nozzle on the lower end of a vertically movable extension tube connected to an Ar gas heater. Structure: After removing control rod drive mechanism and other apparatus mounted on the top of the reactor core for maintenance, a spray nozzle of a maintenance apparatus on a reactor top pit lid is inserted into a through-hole in a shield plug. Then, Ar gas heated by the heater is supplied through the extension tube and sprayed, thereby removing Na slug attached to and solidified on the inner surface of the hole by fusion. (Seki, T.)

  13. Preliminary evaluation of alternate-fueled gas cooled fast reactors

    International Nuclear Information System (INIS)

    A preliminary evaluation of various alternative fuel cycles for the Gas-Cooled Fast Reactor (GCFR) is presented. Both homogeneous and heterogeneous oxide-fueled GCFRs are considered. The scenario considered is the energy center/dispersed reactor concept in which proliferation-resistant denatured reactors are coupled to 233U production reactors operating in secure energy centers. Individual reactor performance characteristics and symbiotic system parameters are summarized for several possible alternative fuel concepts. Comparisons are made between the classical homogeneous GCFR and the advanced heterogeneous concept on the basis of breeding ratio, doubling time, and net fissile gain. In addition, comparisons are made between a three-dimensional reactor model and the R-Z heterogeneous configuration utilized for the depletion and fuel management calculations. Lastly, thirty-year mass balance data are given for the various GCFR fuel cycles studied

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

    International Nuclear Information System (INIS)

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

  15. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    International Nuclear Information System (INIS)

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

  16. Advanced Multiphysics Modeling of Fast Reactor Fuel Behavior

    International Nuclear Information System (INIS)

    Evaluation of fast reactor fuel thermo-mechanical performance using fuel performance codes is a key aspect of advanced fast reactors designs. Those fuel performance codes capture the multiphysics nature of fuel behavior during irradiation where different, mostly interdependent, phenomena are taking place. Existing fuel performance codes do not fully capture those interdependencies and present the different phenomena through de-coupled models. Recent developments in multiphysics simulation capabilities and availability of advanced computing platforms led to advancements in simulation of nuclear fuel behavior. This paper presents current experiences in applying different multiphysics simulation platforms to evaluation of fast reactors metallic fuel behavior. Full 3D finite element simulation platforms that include capabilities to fully couple key fuel behavior models are discussed. Issues associated with coupling metallic fuels phenomena, such as fission gas models and constituent distribution models, with thermo-mechanical finite element platforms, as well as different coupling schemes are also discussed. (author)

  17. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, Stephen Rhead [Idaho National Lab. (INL), Idaho Falls, ID (United States); Medvedev, Pavel G. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

  18. Capital cost: gas cooled fast reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design.

  19. Construction of Soviet fast reactor BN-600

    International Nuclear Information System (INIS)

    A sectional view is shown of the integral configuration of the 3rd unit reactor in the Beloyarsk nuclear power plant. The reactor vessel is a cylinder 12.8 m in diameter and 12.6 m in height. In view of overpressure in the vessel (40 kPa) the wall thickness is 30 to 40 mm. The reactor core contains 370 hexagonal fuel elements. Each element consists of 127 pins of an outer diameter of 6.9 mm. 27 positions are taken by regulating and scram rods. The fuel reserve in the core and the efficiency of reactivity control permits reactor operation for about 150 days such that one third of the fuel elements is exchanged during refuelling. A block diagram is shown of the power plant heat generating system. Core cooling is ensured by three circuits, i.e., the sodium primary and secondary circuits and one water and steam circuit. The progress of the power plant construction is briefly indicated. (J.P.)

  20. Capital cost: gas cooled fast reactor plant

    International Nuclear Information System (INIS)

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design

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

    International Nuclear Information System (INIS)

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

  2. Small size modular fast reactors in large scale nuclear power

    International Nuclear Information System (INIS)

    The report presents an innovative nuclear power technology (NPT) based on usage of modular type fast reactors (FR) (SVBR-75/100) with heavy liquid metal coolant (HLMC) i. e. eutectic lead-bismuth alloy mastered for Russian nuclear submarines' (NS) reactors. Use of this NPT makes it possible to eliminate a conflict between safety and economic requirements peculiar to the traditional reactors. Physical features of FRs, an integral design of the reactor and its small power (100 MWe), as well as natural properties of lead-bismuth coolant assured realization of the inherent safety properties. This made it possible to eliminate a lot of safety systems necessary for the reactor installations (RI) of operating NPPs and to design the modular NPP which technical and economical parameters are competitive not only with those of the NPP based on light water reactors (LWR) but with those of the steam-gas electric power plant. Multipurpose usage of transportable reactor modules SVBR-75/100 of entirely factory manufacture assures their production in large quantities that reduces their fabrication costs. The proposed NPT provides economically expedient change over to the closed nuclear fuel cycle (NFC). When the uranium-plutonium fuel is used, the breeding ratio is over one. Use of proposed NPT makes it possible to considerably increase the investment attractiveness of nuclear power (NP) with fast neutron reactors even today at low costs of natural uranium. (authors)

  3. Advanced monitoring and control systems for fast reactors

    International Nuclear Information System (INIS)

    One of the important aspects of nuclear power station (NPS) improvement with fast reactors is provision of safety. The safety conception of advanced fast power reactors is directed on elaborating such solutions where as much as possible properties of reactor self-protection and natural laws are used in which the self-protection of the nuclear reactor is realized. To these solutions we may refer the usage of hydraulically weighted rods of alarm protection, negative temperature and power coefficients, negative sodium empty effect, natural circulation without power sources, natural convection and other measures. Additionally special technological systems are envisaged, which start functioning with the coming of the initial event of the accident. 1 ref., 7 figs, 1 tab

  4. Fast Reactor Physics Parameters from a Pulsed Source

    International Nuclear Information System (INIS)

    One of the more important integral parameters in fast reactor physics analysis is the neutron spectrum of a particular composition reactor core. Various methods, such as proton recoil counters and nuclear emulsion analysis, have been used to study fast reactor spectra. With the development of high intensity short-duration pulsed neutron sources, the time-of-flight technique has become suitable for fast reactor spectrum determination. To evaluate the feasibility of measuring fast neutron spectra from a core using time-of-flight techniques, an experiment has been performed to measure the equilibrium spectmm in a large block of depleted uranium using the General Atomics Linac facilities. A ten-metric-ton block of depleted uranium was assembled to form a 81-cm cube. This block of uranium was pulsed by electron bombardment of a uranium target imbedded in the block. The spectra from various sections of the block were measured using time-of-flight techniques for a 50-m flight path. Spectral indices, such as the ratio of the fission rates of U238/U235, U233/U235, U234/U235, Np237/U235, Pu239/U235 were also measured. In addition, measurements of the U238 capture rates were obtained in various parts of the block. This paper describes the techniques used to obtain these reactor physics parameters. The experimental results such as the spectra and spectral indices are also compared with those obtained from theoretical considerations using multigroup transport theory analysis. The pulsed neutron technique is also applicable for the measurement of such parameters as: β/ℓ, where β is the effective delayed neutron fraction and ℓ is the lifetime; neutron importance; and keff. This paper concludes with a discussion on the proposed application of a pulsed neutron source for the measurement of some of these parameters on fast reactor cores constructed on ZPR-VI, the Argonne Fast Critical Facility. (author)

  5. Sodium components cleaning status in the Italian fast reactor program

    International Nuclear Information System (INIS)

    As a consequence of the Italian Fast Reactor Development, mainly aimed to the PEC project and to the participation in the French Superphenix project, it is of increasing importance to set up a reliable method for specific reactor components and related test loops. The first problem was the cleaning of the PEC fuelling machine. In order to perform the routine maintenance of the machine an alcohol cleaning method based on the use of 2-butoxyethanol-NN dimethylformamide mixture has been proposed

  6. Decommissioning of the Rapsodie fast reactor: developing a strategy

    International Nuclear Information System (INIS)

    The RAPSODIE experimental fast neutron reactor at Cadarache (France) was operated from 1962 to 1982. The initial decommissioning operations began immediately, reaching IAEA stage 2 in 1994. Since then, the facility has been maintained under surveillance pending final dismantling scheduled to begin in 2020. New studies are now in progress to accelerate the dismantling process. The present status of the reactor block is described, the advantages and drawbacks of early dismantling are considered, and various dismantling scenarios are discussed. (author)

  7. Studies on the transient operation and stability of fast reactors

    International Nuclear Information System (INIS)

    These studies form part of the general programme of perfecting calculation methods for fast reactors. The basic formulae are given for the layouts used, i.e. the classic kinetic and thermal exchange equations, etc. A description is then given of the digital computer methods employed for studying the stable functioning of the reactor and of the methods used for transient operation studies. Finally, some examples of application are discussed and a comparison is made with parallel studies on the same subject. (author)

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

    International Nuclear Information System (INIS)

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

  9. Fast reactor 3D core and burnup analysis using VESTA

    Energy Technology Data Exchange (ETDEWEB)

    Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

  10. UK fast reactor components - sodium removal decontamination and requalification

    International Nuclear Information System (INIS)

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

  11. Opening Address: Fast Neutron Reactors and Sustainable Development

    International Nuclear Information System (INIS)

    The aim of this presentation is to provide an insight into the challenges that lie ahead for the development of fast reactors. From the moment when the first fast reactor - EBR1 - lit up the city of Arco right up to Superphenix, by far the largest fast reactor ever built, there have been 40 years of fast reactor development, mainly centred on sodium cooled systems, leading to the successful operation of such plants. Therefore, the question could arise about the need for more R and D and the relevance of new prototype designs. There have been two major development steps in the history of fast reactors. During the 1960s and 1970s, their development was undertaken following concerns related to the energy supply, resulting mainly from the oil crisis, as well as from the need to use uranium resources more efficiently. In the 1980s, however, demand for nuclear energy declined after the Three Mile Island and Chernobyl accidents, as well as from the belief that fossil energy was plentiful and would remain cheap. It took about 20 years to realize that nuclear energy would expand, owing to the energy and climate challenges the world was faced with, and with that, the need for fast reactors became obvious in order to account for the constraints of such expansion. Currently, however, the context has changed since the 1970s, and the development of fast reactors needs to be made on a new basis, taking into account new criteria linked to economy, safety, reliability, resource saving, waste minimization and physical protection against terrorism or proliferation. Such huge technological challenges also require that the new fast reactor designs be developed internationally, within multinational cooperation frameworks. Such is the goal of the Generation IV International Forum (GIF), which is a gathering of the major key actors in the field of R and D, cooperating for the sustainable development of nuclear energy. A new way of thinking has emerged from this new context: the awareness

  12. Fabrication and characterisation of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    India is pursuing its three stage nuclear power programme. In 2nd stage it plans to erect number of fast reactors which will be requiring large amount of plutonium as one of the important feed materials. It plans to begin with Mixed Oxide (MOX) fuelled fast reactors and finally shift to metallic fuel based fast reactors. Mixed oxide (MOX) (U, PU)O2, and metallic (U, Pu, Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity, low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion which is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burn up, fuel cladding interaction and lower margin between operating and melting temperature

  13. China experimental fast reactor; Le reacteur rapide experimental chinois

    Energy Technology Data Exchange (ETDEWEB)

    Tianmin, X. [Institut d' Ingenierie Nucleaire de Pekin (China); Cunren, L. [Centre d' Etude de Surete de Pekin (China)

    2007-07-15

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*10{sup 15} n/cm{sup 2}/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  14. Testing stand for cosmic gas-cooling fast reactor's sample

    International Nuclear Information System (INIS)

    For carrying out of technical decision and nuclear, radiation and technological safety of gas-cooling space nuclear power plants is elaborating gas-cooling fast reactor's testing stand. In the base of its draft is taken conception of the reactor with filling up type reactor core on the base of ball fuel elements and radial coolant flowing. On the testing stand would suggested carrying out testing for study neutron and physical parameters of gas-cooling reactor, its behaviour under accident simulation. In the reactor core will suggest use carbon nitrides fuel elements with tungsten cover, provides under nominal regime relatively low fission products yield to first contour of device. Construction of fuel element was carrying out on reactor and non reactor testing and its calculated on working resource about 3000 hours. Constructive materials of reactor core have lower melting temperature, that provides organized in good time remove fuel element to containers placed under reactor in case connected with hypothetical accident. In the construction of reactor for seen tree-contours system of heat transfer and its provides multistage system of barriers against fission products yield to environment. tabs.1

  15. Spectrophotometric Procedure for Fast Reactor Advanced Coolant Manufacture Control

    Science.gov (United States)

    Andrienko, O. S.; Egorov, N. B.; Zherin, I. I.; Indyk, D. V.

    2016-01-01

    The paper describes a spectrophotometric procedure for fast reactor advanced coolant manufacture control. The molar absorption coefficient of dimethyllead dibromide with dithizone was defined as equal to 68864 ± 795 l·mole-1·cm-1, limit of detection as equal to 0.583 · 10-6 g/ml. The spectrophotometric procedure application range was found to be equal to 37.88 - 196.3 g. of dimethyllead dibromide in the sample. The procedure was used within the framework of the development of the method of synthesis of the advanced coolant for fast reactors.

  16. A new neutron noise technique for fast reactors

    International Nuclear Information System (INIS)

    This paper gives a new neutron noise technique for fast reactors, which is known as thermalization measurement technique of the neutron noise. The theoretical formulas of the technique were developed, and a digital delayed coincidence time analyzer consisted of TTL integrated circuits was constructed for the study of this technique. The technique has been tested and applied practically at Df-VI fast zero power reactor. It was shown that the provided technique in this work has a number of significant advantages in comparison with the conventional neutron noise method

  17. Two-lump fission product model for fast reactor analysis

    International Nuclear Information System (INIS)

    As a part of the Fast-Mixed Spectrum Reactor (FMSR) Project, a study was made on the adequacy of the conventional fission product lump models for the analysis of the different FMSR core concepts. A two-lump fission product model consisting of an odd-A fission product lump and an even-A fission product lump with transmutation between the odd- and even-A lumps was developed. This two-lump model is capable of predicting the exact burnup-dependent behavior of the fission products within a few percent over a wide range of spectra and is therefore also applicable to the conventional fast breeder reactor

  18. Fast reactor experiments with thorium at the PROTEUS facility

    International Nuclear Information System (INIS)

    The largescale utilization of thorium is usually linked to its introduction in fast breeder reactors and/or advanced converters. The present experiments were carried out in the zero-energy reactor facility, PROTEUS, at EIR. Six different configurations for the central fast test zone were considered in the current programme, the principal fuel/blanket materials used being in the form of rods of 15% PuO2/UO2, depleted UO2, ThO2 and Th-metal. For each configuration, measurements of the principal reaction rate ratios at the centre, as well as of reaction rate distributions across the test zone, were made. (Auth.)

  19. A review of the Indian fast reactor programme

    International Nuclear Information System (INIS)

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

  20. Fast Breeder Test Reactor: 15 years of operating experience

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 1985 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 1993 and power was raised to 10.5 MWt in Dec 1993. Turbine generator was synchronized to the grid in Jul 1997. The indigenously developed mixed carbide fuel has achieved a peak burn up of 88,000 MWd/t till now at a linear heat rating of 320 W/cm and reactor power of 13.4 MWt without any fuel-clad failure. The paper presents operating and decontamination experience, performance of fuel, steam generator and sodium circuits, certain unusual occurrences encountered by the plant and various improvements carried out in reactor systems to enhance plant availability. (author)

  1. Super fast reactor R and D projects in Japan

    International Nuclear Information System (INIS)

    The Japanese research project of the 'Research and Development of Super Fast Reactor' was conducted from December 2005 to March 2010, entrusted by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT). Aiming at a highly economical fast reactor, the plant concept was developed with quantitative characteristics. The databases of the thermal hydraulics and materials including water chemistry were accumulated by experiments. Based on the success of the project, the second phase of Super FR project was initiated in July 2010. The project consists of three subjects; (1) development of the plant concept: (2) thermal-hydraulics: (3) material-coolant interactions: The Super Fast Reactor has the same plant system of the once-through coolant cycle as the Super Light Water Reactor, the thermal reactor. The results of experimental R and D constitute the common database for the development of Super LWR and Super FR. This paper describes the principle of the reactor concept development and the R and D program of the second phase project. (author)

  2. Parameter analysis calculation on characteristics of portable FAST reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

  3. Parameter analysis calculation on characteristics of portable FAST reactor

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. The Integral Fast Reactor: A practical approach to waste management

    International Nuclear Information System (INIS)

    This report discusses development of the method for pyroprocessing of spent fuel from the Integral Fast Reactor (or Advanced Liquid Metal Reactor). The technology demonstration phase, in which recycle will be demonstrated with irradiated fuel from the EBR-II reactor has been reached. Methods for recovering actinides from spent LWR fuel are at an earlier stage of development but appear to be technically feasible at this time, and a large-scale demonstration of this process has begun. The utilization of fully compatible processes for recycling valuable spent fuel materials promises to provide substantial economic incentives for future applications of the pyroprocessing technology

  7. Operating safety experience of fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Operational safety criteria for nuclear reactors are very stringent and it is essential to incorporate adequate inherent and engineered safety features in the design to ensure safe operation of the reactor. Commissioning and operation of FBTR, being first of its kind in India based on nuclear chain reaction maintained by fast neutrons and use of high temperature liquid sodium as coolant, was a challenging task. Safe operation of the reactor for the past 17 years with good performance of sodium systems and the indigenous plutonium rich carbide fuel, touching a burn up level of 100 GWd/t has underlined the high level of design and operation competence achieved

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-05-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  12. Challenges and achievements - Prototype Fast Breeder Reactor construction

    International Nuclear Information System (INIS)

    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)

  13. Design Features and Operating Experience of Experimental Fast Reactors

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world'. One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The IAEA has begun an initiative to help coordinate Member State efforts in the field of fast neutron nuclear reactors. This initiative is primarily targeted at the preservation of knowledge in the areas of design, construction and operation, for both experimental and power fast reactors. The ultimate goal of this activity is to establish a comprehensive, international inventory of fast reactor data and knowledge, which will be an essential resource for the future development and deployment of fast reactor technology. In this project, carried out within the framework of the

  14. Current status of fast reactors and future plans in India

    International Nuclear Information System (INIS)

    In the Indian energy scenario projections for the future, the nuclear power through fast reactors is expected to play an important role of ∼ 20% of total installed capacity by 2052. Successful operation of 40 MWt/13 MWe capacity Fast Breeder Test Reactor(FBTR) since 1985, strong R and D executed in multidisciplinary domain backed up by manufacturing technology and construction of 500 MWe Prototype Fast Breeder Reactor (PFBR) based on indigenous design have provided high confidence on the success of sodium cooled fast reactor technology. PFBR is a pool type MoX fuelled reactor designed with 2 primary sodium pumps, 2 secondary loops, 8 single wall integrated once through steam generators, and a rectangular containment. PFBR is presently under advanced stage of construction. Beyond PFBR, it is planned to construct 6 more FBRs of 500 MWe capacity each. Towards this, a systematic road map has been drawn for improved economy and enhanced safety through a number of measures. Road map for necessary R and D and manufacturing technology has been well detailed. The major features incorporated are twin unit concept, plant life increased to 60 years in comparison to 40 years for PFBR, reduction in number of steam generators from 8 to 6, reduction in special steel specific weight requirements, integrated primary sodium purification, enhanced reliability of shutdown systems, enhanced diversity in decay heat removal systems, enhanced in-service inspection, and compact plant layout. Beyond 2025, a series of 1000 MWe capacity metallic fuel with higher breeding potential are planned. R and D activities have been systematically formulated for metallic fuel development of both sodium bonded and mechanical bonded design. The paper addresses the highlights of current operating experience of FBTR and its life extension, construction status of PFBR, and design features of future sodium cooled fast reactors in India. (author)

  15. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

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

  16. Current status on fast reactor program in Kazakhstan

    International Nuclear Information System (INIS)

    Atomic scientific-industrial complex of Republic of Kazakhstan consist of: Uranium mining, production and power industry: - Enterprises of uranium ores geological searching and number of natural mines (using the mining and underground leaching techniques); - Two plants of U3O8 production at Aktau and Stepnogorsk towns; - Metallurgical plant producing uranium fuel pellets for fuel assemblies of RBMK and VVER reactors types; - Energy plant at Aktau (MAEK) is used for production of heat, electricity and desalination of water and based on three energy blocks using natural gas and one nuclear unit with fast breeder reactor BN-350. The fast breeder reactor BN-350 at Aktau was commissioned in November 1972 and finally shut down in April 1999. Three different type of the research reactors and non reactor test facility on the territory of the former Semipalatinsk Nuclear Test Site and one research reactor and sub critical assembly nearly Almaty are exploiting for the investigation in field of reactors nuclear safety and other type of investigations

  17. Operating Experience with the BN-600 Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The report considers the main design features of the BN-600 liquid metal fast reactor. The performance indicators achieved for 32 years of operation are given. The measures taken to enhance BN-600 reactor power unit safety and replace and extend its equipment lifetime and their results allowed the design lifetime of the power unit to be extended up to 40 years (until 31 March 2020) are presented. The considered integrated material, methodological and theoretical investigations justifying the serviceability of the irreplaceable components of the BN-600 reactor facility have shown that the strength conditions have not been violated in any of the critical reactor components after 45 years of operation. The results, both of the actions taken to enhance the BN-600 reactor power unit safety and corrective measures related to the events at the Fukushima nuclear power plant, allow the safety of the power unit exposed to any possible extreme external impact to be improved. (author)

  18. Thermal and neutronic calculation for fast breeder reactor FBR

    International Nuclear Information System (INIS)

    This research included studying of thermal and neutronic calculation for fast breeder nuclear reactor, to putting the optimum design for this reactor. So a Soviet type (BN-350) was chosen, which has its core composed of two enrichment zones, and with blanket that contains depleted uranium. A group of thermal calculation programs was made by using personal computer, to obtain core and blanket reactor dimensions and volume fractions of reaction input material and number and dimensions of fuel rods which were used for neutron calculations. Several core and blanket enrichments were used to study neutron flux behaviour for two reactors different conditions. First when control rods exist in the core reactor and second when the rods are out of the core. Breeding ratio was also studied for different core and blanket enrichment. 30 tabs.; 24 figs.; 34 refs.; 3 apps

  19. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  20. Development of Fast Sodium Reactor Technology in the Russian Federation

    International Nuclear Information System (INIS)

    The paper provides information about the development of the sodium cooled fast reactors in the former USSR (Russian Federation) starting from the 1950s. The evolution of this technology is traced from the small research reactors to large power units. It is shown how power and parameters were changing in reactor plants; how engineering solutions on the layout, reactor core design, main equipment and systems were evolving; and how the most important problem on increasing the fuel burnup was being gradually solved. Mastering of the dense nitride fuel instead of the MOX fuel is mentioned as an important challenge. Given are operational results for the first power units with the BN-350 and BN-600 reactors; the experience obtained is evaluated. Characterized are the challenges to be faced in the new BN-800 and BN-1200 projects, as well as information about the status of these projects. (author)

  1. Fast breeder reactor. The past, the present and the future. (6) History of fast reactor development in Japan - 1

    International Nuclear Information System (INIS)

    History and present state of fast breeder reactor was reviewed in series. As a history of fast reactor development in Japan - 1, this sixth lecture presented the start of FBR development, and construction and operation of the experimental FBR (JOYO). The JOYO began operation in 1977 and now is being operated at 140 MWt after two times of upgraded modification. The JOYO is aimed at (1) advancement of technology through and experiment, (2) conducting irradiation tests on fuels and materials and (3) validation of innovative technology for development of a future FBR. (T. Tanaka)

  2. Structural dynamics in fast reactor accident analysis

    International Nuclear Information System (INIS)

    Analyses and codes are under development combining the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage the stresses, strains, and deformations of important primary components, as well as the overall adequacy of primary and secondary containments. An arbitrary partition of the structural components treated evolves into (1) a core mechanics effort; and (2) a primary system and containment program. The primary system and containment program treats the structural response of components beyond the core, starting with the core barrel. Combined hydrodynamics-solid mechanics codes provide transient stresses and strains and final deformations for components such as the reactor vessel, reactor cover, cover holddown bolts, as well as the pulses for which the primary piping system is to be analyzed. Both, Lagrangian and Eulerian two-dimensional codes are under development, which provide greater accuracy and longer durations for the treatment of HCDA. The codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. Recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of primary piping. Pulses are provided at the vessel-primary piping interfaces of the inlet and outlet nozzles, calculation includes the elbows and pressure drops along the components of the primary piping system. Recent improvements to the primary containment codes include introduction of bending strength in materials, Langrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. Another development involves the combination of a 2-D finite element code for the reactor cover with the hydrodynamic containment code

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

    International Nuclear Information System (INIS)

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

  4. Reducing the Void Effect in a Large Fast Power Reactor

    International Nuclear Information System (INIS)

    The coolant void effect has been recognized as one of the most serious safety problems of large, liquid-metal-cooled fast power reactors. Several proposals have been made to reduce the positive void reactivity effect. However, in all cases appreciable penalties with respect to internal breeding ratio, fissile inventory and, finally, economics have to be paid. All these proposals, like the pan-cake core reactor and the annular core reactor, have in common that neutron leakage out of all core zones is artificially increased. Some more detailed calculations of voiding characteristics of large sodium-cooled fast reactors with cylindrical cores indicate that the most important positive contribution to the void reactivity effect arises in the central core zone. Consequently it appears promising to reduce the void effect considerably by increasing deliberately neutron leakage only out of the central core region and not necessarily out of all core regions. Compared with other geometries discussed this suggestion should result in better breeding and economical performance concurrent with a satisfactory void effect. To support this proposal an investigation was conducted on some significant reactor data (breeding ratios, fissile enrichment, power distribution, etc.) for three geometrically different sodium-cooled fast breeder reactors with a power of 1000 MW(e) each. In the first case a normal cylindrical core geometry (2 core zones, 1 axial and 1 radial blanket) with an H: D ratio of roughly 0.33 was considered. The second reactor was similar to the first, except that the height of the inner core zone was reduced by approximately 50% to enhance neutron leakage in the core centre, the lost core volume being balanced by increasing the outer radius of the second core zone. The third reactor had an annular core with dimensions consistent with those of the first reactor. Results show a remarkable improvement in the void effect of the variable-core-height reactor compared with

  5. Integral Fast Reactor Program. Annual progress report, FY 1993

    International Nuclear Information System (INIS)

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D

  6. Thermal-hydraulic methods in fast reactor safety

    International Nuclear Information System (INIS)

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

  7. Integral Fast Reactor Program. Annual progress report, FY 1993

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D.

  8. Integral Fast Reactor Program annual progress report, FY 1994

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1994. Technical accomplishments are presented in the following areas of the IFR technology development activities: metal fuel performance; pyroprocess development; safety experiments and analyses; core design development; fuel cycle demonstration; and LMR technology R&D.

  9. Integral Fast Reactor Program annual progress report, FY 1991

    International Nuclear Information System (INIS)

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  10. Status of the fast breeder reactor technology in China

    International Nuclear Information System (INIS)

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

  11. The Programme for Fast Reactor Development in the Russian Federation

    International Nuclear Information System (INIS)

    The paper highlights the status and perspectives on the development of nuclear energy based on fast reactor and closed fuel cycle technologies in the Russian Federation. Information is presented on the new Federal Target Programme 'Nuclear Power Technologies of a New Generation for the Period 2010-2015 and the Outlook to 2020'. (author)

  12. Basic cable routing guidelines for a fast reactor plant

    International Nuclear Information System (INIS)

    In this paper the guidelines evolved for cable routing in 500 MWe Prototype Fast Breeder Reactor (PFBR) are presented. Safety related redundant system cables in a nuclear plant shall not become unavailable due to cable fire. This is ensured by proper cable routing in the plant in addition to the other general fire protection measures

  13. Integral Fast Reactor Program annual progress report, FY 1994

    International Nuclear Information System (INIS)

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1994. Technical accomplishments are presented in the following areas of the IFR technology development activities: metal fuel performance; pyroprocess development; safety experiments and analyses; core design development; fuel cycle demonstration; and LMR technology R ampersand D

  14. A decade of progress in fast reactor fuel

    International Nuclear Information System (INIS)

    This paper discusses the enormous strides that have been made in the World's knowledge of the behavior of fast reactor fuels and materials over the past 10 yr. With ∼ 12 sodium-cooled fast reactors operating in various countries and with the willingness of the international community to freely share the knowledge gained, the technological achievements have been staggering. There is no doubt that fast breeders can now be built and operated safely and effectively. Mixed-oxide fuel has been demonstrated to reliably meet its design objectives, and the breeder communities are now shifting emphasis to economic and safety improvements. In the US, metal fuel is being explored as a possible successor to mixed oxide, because metal fuel offers the potential for in-house reprocessing. The European Community, Japan, and the Soviet Union are all pursuing nitride fuel. Continued testing under fully prototypical conditions in a high-quality research reactor such as the FFTF is the key to the optimal selection of a fast reactor fuel system for meeting the energy needs of future generations

  15. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  16. Integral Fast Reactor Program annual progress report, FY 1991

    Energy Technology Data Exchange (ETDEWEB)

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

  17. Integral Fast Reactor Program annual progress report, FY 1991

    Energy Technology Data Exchange (ETDEWEB)

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R D.

  18. Symposium on key questions about the fast breeder reactor

    International Nuclear Information System (INIS)

    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

  19. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  20. Choice of pyroprocess for Integral Fast Reactor fuel

    International Nuclear Information System (INIS)

    A design objective for the Integral Fast Reactor (IFR) is fuel self sufficiency. This can be achieved only by employing chemical reprocessing as part of the fuel cycle. Because the fuel is a metal alloy (U-Pu-Zr), direct production of metal is highly advantageous. This makes a pyrometallurgical process attractive

  1. Fast breeder reactor-block antiseismic design and verification

    International Nuclear Information System (INIS)

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

  2. Experience with the generating plant at fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWth/13.2 MW(e) sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and generating systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct.1985 with Mark-I core (70 % PUC-30 % UC). FBTR heat transport system consists of two primary sodium loops, two secondary sodium loops and one common tertiary steam and water circuit. Heat generated in the reactor core is transported to the tertiary loop by primary and secondary sodium loops. The steam water system mainly consists of a once through steam generator, which produces super heated steam at a pressure of 120 bars and temperature of 480 degC, feed water system and condensate system. The steam produced is supplied to a condensing turbine. The turbine in turn is coupled to an alternator. The steam generator was put in service in Jan.1993 and turbine generator was synchronized to the grid in July 1997. The paper presents operating experience with generating plant consisting of steam water circuit, condensing turbine and its associated systems and the alternator, various modifications carried out to improve system reliability and availability and certain incidents taken place in the generating plant. (author)

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

    International Nuclear Information System (INIS)

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

  4. Vessel core seismic interaction for a fast reactor

    International Nuclear Information System (INIS)

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

  5. Fast Reactor Knowledge Organization System: Implementation and challenges

    International Nuclear Information System (INIS)

    For three decades, several countries had large and vigorous fast breeder reactor development programmes, which had their peaks by 1980. From that time onward, Fast Reactor (FR) development generally began to decline and efforts for FR reactor development essentially disappeared by 1994. This development stagnation continued until 2003. In September 2003, in Resolution GC(47)/RES/10.B, the International Atomic Energy Agency (IAEA) General Conference recognised the vitality of nuclear knowledge. The loss of FR knowledge has been taken seriously and the IAEA took the initiative to coordinate the efforts of the member states in the preservation of knowledge in FRs. In the framework of this initiative, the IAEA intends to create an international inventory combining information from different member states on FRs and organized in the knowledge system in a systematic and structured manner

  6. Simulation on primary coolant system of sodium cooled fast reactor

    International Nuclear Information System (INIS)

    In this paper the thermal-hydraulic characteristics of the primary loop of China Experimental Fast Reactor (CEFR) are calculated and analyzed. A one-dimension, single-phase flow model is used to establish the system control equations. The single channel model is adopted in the reactor core, and a dynamic model of intermediate heat exchanger is built. At the same time, the property of sodium and flow and heat transfer correlations or models of sodium are collected and compiled. The discussion of the sensitivity of different flow and heat transfer correlations is given. The validation of the code developed in this paper shows that the code can be adopted to do some typical transient and accident analysis. The model and code presented in this paper can be used not only in the safety analysis of pool-type sodium cooled fast reactor, but also in the development of CEFR simulation platform. (author)

  7. Alternate fuel cycles for fast breeder reactors

    International Nuclear Information System (INIS)

    In this contribution to the syllabus for Subgroup 5D, a full range of alternate breeder fuel cycle options is developed and explored as to energy supply capability, resource utilizations, performance characteristics and technical features that pertain to proliferation resistance. Breeding performance information is presented for designs based on Pu/U, Pu/Th, 233 U/U, etc. with oxide, carbide or metal fuel; with lesser emphasis, heterogeneous and homogeneous concepts are presented. A potential proliferation resistance advantage of a symbiotic system of a Pu/U core, Th blanket breeder producing 233 U for utilization in dispersed LWR's is identified. LWR support ratios for various reactor and fuel types and the increase in uranium consumption with higher support ratios are identified

  8. Operating experience of BN-600 fast neutron reactor and BN-800 reactor design

    International Nuclear Information System (INIS)

    Full text: Experience gained in Russia (USSR) in R and D work in the area of sodium cooled fast reactors in the period of 1950-1970s has been used in the design of NPP with the BN-600 reactor. Since its start-up in 1980, BN-600 reactor has demonstrated operating characteristics, which are unique for this nuclear technology. Average load factor value for 23 years of operation is near 74%, its values in 2002 and 2003 being respectively 77.35% and 75.7%. Release of inert radioactive gases is within 0.3% of reference value, while average collective dose rate of personnel is about 0.3 man. Sv per year. In the course of operation of NPP with the BN-600 reactor, effectiveness of steam generator protection system was demonstrated in 12 cases of small and large water-into-sodium leaks. Besides, unique experience was gained in confining either radioactive and non-radioactive sodium fires in case of sodium leaks from the circuits. Radioactive sodium leak from the primary auxiliary circuit occurred in December 1994 is a typical example. Total amount of sodium released from the circuit was about 1000 kg, and protection system was capable of confining sodium ignition nucleation site and limiting radioactivity release to the atmosphere by 10 Ci value. This release has had almost zero effect on radiological conditions of the NPP controlled area. BN-800 reactor design is the next stage of development of sodium cooled fast reactor technology. Fourth power unit with the BN-800 reactor is now under construction on Beloyarskaya NPP site. Innovative design approaches have been used in the BN-800 reactor in order to further improve safety of fast reactors with sodium coolant. Among these innovations are as follows: Additional 'passive' safety system using three absorber rods hydraulically suspended by the sodium flow; Passive decay heat removal system using sodium-air heat exchangers; Device for collection and retaining of the core debris in case of its disruption under conditions of

  9. Effect of neutron anisotropic scattering in fast reactor analysis

    International Nuclear Information System (INIS)

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

  10. Fast Neutron Detector for Fusion Reactor KSTAR Using Stilbene Scintillator

    CERN Document Server

    Lee, Seung Kyu; Kim, Gi-Dong; Kim, Yong-Kyun

    2011-01-01

    Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spectra with high radiation from a gamma-ray background. The signals from neutrons and gamma-rays are discriminated by the digital charge pulse shape discrimination (PSD) method, which uses total to partial charge ratio analysis. The signals are digitized by a flash analog-to-digital convertor (FADC). To evaluate the performance of the fabricated stilbene neutron diagnostic system, the efficiency of 10 mm soft-iron magnetic shielding and the detection efficiency of fast neutrons were tested experimentally using a 252Cf neutr...

  11. The generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    The gas cooled fast reactor (GFR) is a helium-cooled fast spectrum reactor operating within a closed fuel cycle. It combines the advantages of fast reactors, in terms of a more sustainable use of uranium resources and waste minimisation, with the wider applicability of high temperature gas reactors, in terms of high efficiency electricity generation and the co-generation of high-quality process heat. Other advantages like the absence of threshold effect due to phase changing, the optical transparency and chemical inertness of the Helium coolant are also acknowledged. Within the European Union, GFR is one of the three fast reactors proposed for development to the demonstration stage within the European Sustainable Nuclear Industry Initiative (ESNII). On a wider global scale, GFR is one of the six systems proposed for further development within the Generation IV International Forum (GIF). In this respect, France, Switzerland, Japan and the European Union (through EURATOM) are signatories to the 'System Arrangement', the instrument through which the international research efforts are coordinated. This paper presents the current status of the development of the GFR system. The status of the GFR programme in each of the signatory countries is summarised including the intended contribution of the newly launched EURATOM 7. Framework Programme project - GoFastR. France has provided the bulk of the effort on conceptual design, safety assessment and fuel development. Switzerland makes significant contributions to the GFR system in the areas of core physics, uncertainty analysis, deterministic safety assessment and fuel development. Historically Japan has been very active in the development of the GFR system. Within the Generation IV GFR system, Japan contributes to the development of fuel and core materials

  12. Application of Candle burnup to small fast reactor

    International Nuclear Information System (INIS)

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

  13. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    Directory of Open Access Journals (Sweden)

    Vladimir Petrochenko

    2012-09-01

    Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

  14. High temperature fast reactor for hydrogen production in Brazil

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. The renaissance of fast sodium reactors 2007 assessment: situation and contributions from the Phenix experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J. [Phenix Plant (France)

    2007-07-01

    The first nuclear reactor to produce electrical current was the fast sodium/potassium reactor EBR-1 in Idaho (Usa). Following this pioneering experience, France, Germany, Great Britain, Usa, Japan, Russia and India launched construction of fast sodium reactors. In the post Chernobyl years, waves of protest against nuclear power grew and swelled, leading to a strong overall slowdown for this reactor type. The SNR-300 project in Germany never started up, and was shut down. In Great Britain, PFR was definitely shut down, operation of MONJU in Japan and BN-800 project in Russia were frozen, FFTF in the United States shut down, and finally the SPX-1 project in France was also stopped. When PHENIX started back up in 2003, there were only three other research reactors operating worldwide: FBTR in India, BOR-60 in Russia and JOYO in Japan, and one power reactor BN-600 in Russia. The Generation-IV initiative was the opportunity for global thinking about reactors for the future, referred to as fourth generation reactors. Six reactor designs were selected, including the fast sodium reactor. However, after several years, most of the countries have officially announced or confirmed that the fast sodium reactor is their priority reference design. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision to build a prototype scheduled for operation in 2020. These and other plans are all sustained in a very practical manner by the ongoing production in the field. PHENIX has been operating since 2003, contributing to the development of future systems and demonstrating the fast reactors ability to burn waste. Following the excellent results obtained by the BN-600, Russia has re-launched the BN-800 project. China is currently in the process of building a 65 MW research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU for divergence in 2008. In India, a 1200 MW (thermal) power reactor is under

  17. The renaissance of fast sodium reactors 2007 assessment: situation and contributions from the Phenix experimental reactor

    International Nuclear Information System (INIS)

    The first nuclear reactor to produce electrical current was the fast sodium/potassium reactor EBR-1 in Idaho (Usa). Following this pioneering experience, France, Germany, Great Britain, Usa, Japan, Russia and India launched construction of fast sodium reactors. In the post Chernobyl years, waves of protest against nuclear power grew and swelled, leading to a strong overall slowdown for this reactor type. The SNR-300 project in Germany never started up, and was shut down. In Great Britain, PFR was definitely shut down, operation of MONJU in Japan and BN-800 project in Russia were frozen, FFTF in the United States shut down, and finally the SPX-1 project in France was also stopped. When PHENIX started back up in 2003, there were only three other research reactors operating worldwide: FBTR in India, BOR-60 in Russia and JOYO in Japan, and one power reactor BN-600 in Russia. The Generation-IV initiative was the opportunity for global thinking about reactors for the future, referred to as fourth generation reactors. Six reactor designs were selected, including the fast sodium reactor. However, after several years, most of the countries have officially announced or confirmed that the fast sodium reactor is their priority reference design. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision to build a prototype scheduled for operation in 2020. These and other plans are all sustained in a very practical manner by the ongoing production in the field. PHENIX has been operating since 2003, contributing to the development of future systems and demonstrating the fast reactors ability to burn waste. Following the excellent results obtained by the BN-600, Russia has re-launched the BN-800 project. China is currently in the process of building a 65 MW research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU for divergence in 2008. In India, a 1200 MW (thermal) power reactor is under

  18. A review of fast reactor programme in India - April 1992

    International Nuclear Information System (INIS)

    There is no change in the basic policy for development of nuclear energy in India. Fast Breeder Reactors are required to be available commercially to supply increasing quantities of nuclear energy when the first phase programme of deployment of Pressurised Heavy Water Reactors would be reaching the limit imposed by indigenously available natural uranium. Based on presently proven reserves of economically exploitable uranium one cannot expect to support more than 10 to 15 million kilowatt of installed capacity of PHWRs. The immediate goal of the Fast Reactor Programme therefore, remains completion by 2002-2003 of the first 500 MWe Prototype Fast Breeder Reactor which will become the first reactor in the series of reactors to be built there afterwards. This will enable addition of one 500 MWe reactor each year even if the first phase of programme of PHWR is limited to 6.0 million kilowatt. The capital cost of installed kilowatt for FBRs is expected to be comparable to the capital cost per kilowatt for PHWRS. It is expected to launch the construction of PFBR in the next 2 or 3 years as soon as the over all economic condition shows some improvement. In the meantime, manufacturing development of important NSS components like Steam Generators, Sodium Pumps, Main Vessel and Inner Vessel has been initiated. Detailed designs of Control Rod Drive Mechanism (Primary) has been completed and contacts with the manufacturers are being established to identify the industry which would be entrusted with the responsibility of manufacturing the Control Rod Drive Mechanisms. Manufacturing technology for making cladding tubes of D9 stainless steel has been developed and significant progress has been made towards the production of hexagonal wrapper (i.e. Hex-Cans). Inclined Fuel Transfer Machine for loading and unloading the fuel from the Main Vessel has been designed and manufacturing of the prototype machine has been initiated. It is hoped that these steps will enable timely completion

  19. Investigation of small and modular-sized fast reactor

    International Nuclear Information System (INIS)

    In this paper, feasibility of the multipurpose small fast reactor, which could be used for requirements concerned with various utilization of electricity and energy and flexibility of power supply site, is discussed on the basis of examination of literatures of various small reactors. And also, a possibility of economic improvement by learning effect of fabrication cost is discussed for the modular-sized reactor which is expected to be a base load power supply system with lower initial investment. (1) Multipurpose small reactor (a) The small reactor with 10MWe-150MWe has a potential as a power source for large co-generation, a large island, a middle city, desalination and marine use. (b) Highly passive mechanism, long fuel exchange interval, and minimized maintenance activities are required for the multipurpose small reactor design. The reactor has a high potential for the long fuel exchange interval, since it is relatively easy for FR to obtain a long life core. (c) Current designs of small FRs in Japan and USA (NERI Project) are reviewed to obtain design requirements for the multipurpose small reactor. (2) Modular-sized reactor (a) In order that modular-sized reactor could be competitive to 3200MWe twin plant (two large monolithic reactor) with 200kyenWe, the target capital cost of FOAK is estimated to be 260kyen/yenWe for 800MWe modular, 280kyen/yenWe for 400MWe modular and 290kyen/yenWe for 200MWe by taking account of the leaning effect. (b) As the result of the review on the current designs of modular-sized FRs in Japan and USA (S-PRISM) from the viewpoint of economic improvement, since it only be necessary to make further effort for the target capital cost of FOAK, since the modular-sized FRs requires a large amount of material for shielding, vessels and heat exchangers essentially. (author)

  20. A computer code for analysis of core transient behavior in a Na-cooled metal fuel fast reactor

    International Nuclear Information System (INIS)

    The core transient behavior calculation code 'EXCURS' for a Na-cooled oxide fuel fast reactor was modified for the application to a Na-cooled metal fuel fast reactor (LMR). The results of the core transient behavior calculated with the modified EXCURS were compared with those calculated by ANL for EBR-II and also compared with those by CRIEPI for 1000MWe-LMR. These calculations agreed quite well. The modified EXCURS, therefore, can be used for analysing the core transient behavior of LMR. In a design study of actinide burner reactors (ABR), the analysis of core transient behavior is important from the viewpoint of safety. The ULOF and UTOP analyses for a Na-cooled metal fuel ABR (M-ABR) were carried out using the modified EXCURS. The effect of heat conductivity of fuel and that of feedback reactivity coefficients on the core transient behavior were also evaluated. It is calculated that the maximum temperature of fuel is strongly affected by flowering reactivity coefficient, delayed neutron fraction and heat conductivity of fuel in this order. (author)

  1. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    International Nuclear Information System (INIS)

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

  2. Performance and characteristics of a small-sized oxide-fuelled fast reactor (PRISM) for the burning of excess plutonium

    International Nuclear Information System (INIS)

    The increasing availability of fissile material and the worldwide perception of the need to reduce or control such material make a new approach towards nuclear reactors necessary. The Nuclear Fission Division at ENEA is presently interested in evaluating various options for the burning of excess plutonium, among which the LWR reactors using rock-like inert matrix fuel in a once-through cycle and fast reactors with recycling. This paper deals with the latter option. A fast system, rich in neutrons, can produce plutonium by breeding, or it can burn it (together with Minor Actinides) with a high degree of effectiveness: this flexibility makes the fast system worth considering. In the framework of a co-operation project between GE-USA and ENEA, the PRISM MOD D oxide-fuelled reactor (small size, 840 MW) was studied as a burner. A complete set of studies was carried out covering several fields: neutronics, thermohydraulics, dynamics and safety. A negative void coefficient was achieved thanks to a particular arrangement of the fuel subassemblies (leaky core) and to the enrichment distribution. The central zone, loaded with shielding subassemblies, accounts for a negative void coefficient even in the inner zone. Moreover, this zone could become a suitable place for burning long-lived fission products. As a burner, using a standard fuel, the expected performance is a consumption of about 60-70 kg Pu/TWhe. ULOF and TOP analysis show a begin transient evolution. Only half of the GEMs are sufficient to close the ULOF transient; the reactor power reaches its maximum at 1.35 of the nominal figure during a TOP accident. A method of detecting a failure in a subassembly and identifying the subassembly concerned was developed. It was demonstrated that it is possible to create a sufficient number of tags for them to be recognised as different by the detector system, by using 'cocktails' of only four tag gases. The total gas required to overcome the uncertainties is less than 15 ncm3

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

    International Nuclear Information System (INIS)

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

  4. Status of fast reactor development in India

    International Nuclear Information System (INIS)

    The economic liberalization process has accelerated the industrial growth and this requires considerable energy input. Nuclear energy has to play an important role in supporting the increasing demand of energy. In this year FBTR was operated at 10.2 MWt power in a sustained manner. Several physics and engineering experiments were carried out. The mixed carbide fuel has achieved a peak burnup of 16,000 MW d/t without failure. First batch of irradiation experiments was completed and pins were delivered to RML for PIE. For PFBR, a thorough review of the conceptual design was carried out towards reducing the capital cost, construction time and for improving plant reliability. A 2 loop concept with 2 PSPs, 4 IHXs and 2 secondary loops having 4 integrated SG modules has been finally chosen with an expected savings of 15% in NSSS capital cost, 2-3 years in construction time and 10-12% in capacity factor. Number of materials to be used for major components was reduced to three to facilitate speedy development. Operating temperatures were finalized after optimization studies. Discussion is in progress to finalize fuel handling system. Research and development activities are continuing at ICGAR in the areas of reactor physics, engineering development, core engineering, thermal hydraulics, structural mechanics, metallurgy, post irradiation examination, instrumentation and electronics, chemistry, fuel reprocessing, safety research and health physics etc. (author). 9 figs

  5. Oxygen distribution in fast reactor oxide fuels

    International Nuclear Information System (INIS)

    The chemistry of irradiated fuel pins has been examined with respect to oxygen distribution. Two hypostoichiometric mixed-oxide fuel pins were irradiated to high burn-up under controlled conditions in a materials testing reactor. The pins were carefully characterized initially and were examined in detail by ceramographic and electron-probe techniques. The radial oxygen profile at the end of irradiation was measured. Small specimens were extracted from different points on a radius of the fuel; their oxygen potential was measured in a micro-galvanic cell. The results have shown that, although the pins had initially widely different oxygen contents (O/M values of 1.948 and 1.976), the final O/M values were virtually identical at 1.992 and 1.997. This is contrary to the behaviour predicted from existing fuel chemistry theory. Oxidation above 1.997 O/M ratio in the higher initial-O/M pin is prevented by the oxidation of the clad. Rapid failure of the thoria electrolyte in the high-temperature galvanic cell occurred under the influence of minute quantities of irradiated fuel. Such an effect imposes limits on the large-scale application of galvanic-cell techniques to irradiated materials. (author)

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

    International Nuclear Information System (INIS)

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

  7. A neutronic study on advanced sodium cooled fast reactor cores with thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    Highlights: • SFR burner core configurations are explored and analyzed for effective use of thorium blankets. • Thorium blankets can significantly improve SFR burner core performances. • No recycling or partial recycling of Th blankets with multi-batches is very effective. - Abstract: In this paper, new design concepts of sodium cooled fast reactor (SFR) cores having thorium blanket are suggested for pursuing effective burning of TRU (transuranics) nuclides from LWR spent fuels and their neutronic performances are analyzed. Several core configurations having different arrangements of thorium blankets are explored to improve the core performances and safety-related parameters including sodium void worth which is one of main concerns on safety of SFR cores. Specifically, axial and radial thorium blankets are considered for two type cores. The first one is the typical annular type cores having two different fuel regions where axial thorium blankets are placed in the axially central regions while the second one is the single fuel region cores having central non-fuel region where the axial blanket and radial blankets are considered. Also, the effects of the recycling options and fuel management schemes of the used thorium blanket on the core performances are analyzed. The core performance analyses show that thorium blankets with no recycling option and multi-batch fuel management schemes are very effective to improve the core performances including burnup reactivity swing, sodium void worth and TRU consumption rate

  8. A review of the United Kingdom fast reactor programme

    International Nuclear Information System (INIS)

    Total energy consumption in the UK in 1989 was 340 million tonnes of coal or coal equivalent, made up as follows: coal 31%, petroleum 35%, natural gas 24%, nuclear electricity 8%, hydroelectricity 1% and imported electricity 1%. About half of the nuclear electricity generated came from 14 Advanced Gas-Cooled Reactors (AGRs) and about half from the 24 older gas-cooled Magnox reactors, one Steam-Generating Heavy-Water Reactor (SGHWR) and one fast reactor (the Prototype Fast Reactor, PFR, at Dounreay). The privatization of the Electricity Supply Industry (ESI) in the UK is proceeding. On 9 November 1989, however, it was announced by the Secretary of State for Energy that the privatization plan would be changed and that the CEGB's nuclear stations were to remain in state ownership, through the formation of an additional company, Nuclear Electric. At the same time, the Secretary of State for Scotland announced the formation of a similar state-owned company, Scottish Nuclear. Nuclear Electric was asked, in the interim, to examine priorities in the whole nuclear field with particular reference to the improvement of the economics and performance of existing reactors, to the development of the Sizewell and alternative reactors and to the development of longer-term options such as the fast reactor and fusion. Nuclear Electric has been asked to formulate its new policy by June 1990. The PFR programme will continue to be funded by the UK government until March 1994. AEA Technology is endeavouring to find alternative funding to maintain the operation of the PFR until at least the year 2000. The House of Commons Select Committee on Energy stated in its report that the fast reactor ''is a matter for the British Government to foster as a long-term option for the generation of electricity in this country'', and recommended that in the interim the Government reassesses its position on this new technology in the light of increasing concern about CO2 emissions and the long

  9. Design of fuel fabrication plant of Fast Reactor Fuel Cycle Facility for reload requirement of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    India's economic growth is on a fast growth track. The energy demand is expected to grow rapidly in the coming decades. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear energy is best suited to meet this demand in a sustainable manner without causing undue environmental impact. Fast reactors are expected to be major contributors in sufficing this demand to a great extent. As an effort to achieve the objective, a Prototype Fast Breeder Reactor is being constructed at Kalpakkam. This paper also highlights the design features of FFP, unit operations, scheme of automation, branched layout of glove box train, shielding arrangement on glove boxes, accident consequence analysis etc.

  10. Scenario analysis for transuranic transmutation by using fast reactors

    International Nuclear Information System (INIS)

    Symbiotic fast reactor scenarios with the existing nuclear power systems have been analyzed from the viewpoint of a transuranics transmutation. In this study, a sodium-cooled fast reactor (SFR) and an accelerator driven system (ADS) are considered as representative fast reactor systems. For a comparative analysis of the fuel cycle options, the once-through fuel cycle was at first analyzed based on the current nuclear power plant construction plan and the currently operating nuclear power plants such as the pressurized water reactor (PWR) and the Canada deuterium uranium (CANDU) reactor. After setting up a once-through fuel cycle model, the SFR and ADS scenarios were modeled based on the same nuclear energy demand prediction used for the once-through fuel cycle. Then important fuel cycle parameters such as the amount of the spent fuel and corresponding plutonium, minor actinides and fission products inventories were estimated and compared with those of the once-through fuel cycle. In this fuel cycle model, the Pyro process is assumed for all the spent fuel recycling. In the process all the actinides are recovered and some fraction of the fission product is removed. The deployment fractions of the fast reactor are 25, 10 and 20% for the periods of 2030-2040, 2041-2070 and 2071-2100, respectively. In order to feed the fast reactor systems, it was also assumed that the PWR and CANDU spent fuels are reprocessed from 2025 and the fast reactor spent fuel reprocessing begins in 2035. The fuel cycle calculation was performed by the DYMOND code, which has been used for an analysis of the Generation-IV road map studies. The analysis results of the once-through fuel cycle can be summarized as follows: - The nuclear power demand is expected to grow to 25.2 GWe in the year 2100. - The total spent fuel inventory is expected to be 65000 t in 2100. - The transuranics and fission product inventories are estimated to be 660 and 2390 t, respectively, in 2100. The fast reactor cycle

  11. Fast reactor development strategy targets study in China

    International Nuclear Information System (INIS)

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

  12. Fast reactors will eat nuclear waste from LWR

    International Nuclear Information System (INIS)

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

  13. Physics aspects of metal fuelled fast reactors with thorium blanket

    International Nuclear Information System (INIS)

    Metal fuelled fast breeder reactors (MFBR) with high breeding ratio will play a major role in meeting the high nuclear power growth envisaged in India. In this regard several conceptual reactor designs with alloys of U–Pu–Zr fuel have been suggested for commercial operations. This study focusses on the physics design aspects of a sodium cooled U–Pu–6%Zr fuelled 1000 MWe fast breeder reactor, which can attain a breeding ratio of nearly 1.5. The calculation results on reactor kinetics and safety parameters of the 1000 MWe MFBR are presented. The changes in the breeding ratio by introduction of thorium in the blankets of the MFBR are also investigated. Burnup analyses are carried out to compare the core burnup effects in MOX and metal fuelled FBRs. Since the MOX fuelled 500 MWe prototype fast breeder is getting constructed at IGCAR, for burnup comparisons a MFBR of similar design is considered. The results of this study indicate that the loss of reactivity in the metal core with burnup is less than half that of a MOX core and its breeding ratio remains nearly constant. It is also found that the isotopic composition of plutonium (Pu-vector composition) remains more steady with burnup in a metal core

  14. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

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

  15. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

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

  16. Fast reactor development programme in France during 1995

    International Nuclear Information System (INIS)

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

  17. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

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

  18. Innovations in Equipment Erection of Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is sodium cooled, pool type reactor with generating capacity of 1250 MWt/500 MWe. Reactor assembly consists of large dimensional vessels like Safety vessel (13.54 m diameter, 12.8 m height and weight approximately 155 MT) and Main vessel (12.9 m diameter, 12.94 m height and weight approximately 202 MT including core catcher, core support structure and cooling pipes) and Steam generator (26 m length, 1.5 m diameter, and weight approximately 35 MT). PFBR reactor equipment erection was a challenging task where thin walled vessels had transported and handled with utmost precaution to avoid radial forces on the vessels which could buckle the vessels. There was a real challenge in lifting the vessels without swing, placement of large size and heavy vessel at a distance of 57 m where the crane operator had no line of site to the equipment being erected. To handle such over dimensional reactor components many mock-up tests had been carried out before erection and gained lot of confidence. Lot of care had been taken during lifting, handling and erection of thin walled over dimensional reactor components with innovative methods used for lifting fixtures, guiding arrangements, alignment fixtures and achieved the stringent erection tolerances. This paper discusses the first ever experiences gained during the handling and erection of such thin walled, over dimensional reactor components at PFBR site. (author)

  19. Fast reactor programme. Annual progress report 1982

    International Nuclear Information System (INIS)

    The status of recent fast-capture cross sections for important fission-product nuclides has been reviewed; an intercomparison of evaluations for Eu-isotopes has been made and corrections have been applied to recent reported evaluations of neutron capture cross sections for Pd isotopes. An outline of the evaluation procedure for the nuclides sup(58g)Co and sup(58m)Co is given. The evaluation of the cover-gas nuclides has been completed with additional results for 36Ar and 38Ar. Some results of the latest fuel failure experiments under simulated reduced coolant conditions, the so-called SHOT experiments, are given. The first irradiation experiments with the prototype irradiation facility HFR-TOP 01 are described. Neutron flux calculations have been performed to determine the dimensions of a flux depression plate to achieve a symmetric flux distribution inside the fuel during pre-irradiation. The creep investigations on various heats and welded joints of DIN 1.4948 have been finished; the main findings are reported. A first project on the low-cycle fatigue behaviour of DIN 1.4948 has been completed. A three-dimensional finite element analysis has been performed on compact tension test specimen having a curved crack-front due to crack-tunneling. A code version of the VITESSE computer code has been developed to predict the thermohydraulic behaviour of distorted bundle geometries. Results from the LDA measuring programmes in the different test sections with respect to the secondary flow velocities are reported. Noise measurements in an unblocked 60 deg. reference bundle have been performed. (Auth.)

  20. Physical and technical aspects of lead cooled fast reactors safety

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    2001-07-01

    The safety analysis of lead-cooled fast reactors has been performed for the well-developed concept of BREST-OD-300 reactor. The most severe accidents have been considered. An ultimate design-basis accident has been defined as an event resulting from an external impact and involving a loss of leak-tightness of the lead circuit, loss of forced circulation of lead and loss of heat sink to the secondary circuit, failure of controls and of reactor scram with resultant insertion of total reactivity margin, etc. It was assumed in accident analysis that the protective feature available for accident mitigation was only reactivity feedback on the changes in the temperatures of the reactor core elements and coolant flow rate, and in some cases also actuation of passive protections of threshold action in response to low flow rate and high coolant temperature at the core outlet. It should be noted that the majority of the analyzed accidents could be overcame even without initiation of the above protections. It has been demonstrated that a combination of inherent properties of lead coolant, nitride fuel, physical and design features of fast reactors will ensure natural safety of BREST and are instrumental for avoiding by a deterministic approach the accidents associated with a significant release of radioactivity and requiring evacuation of people in any credible initiating event and a combination of events. (author)

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

    International Nuclear Information System (INIS)

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

  2. Recommendations for a demonstrator of Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

    There is currently a renewed interest in molten salt reactors, due to recent conceptual developments on fast neutron spectrum molten salt reactors (MSFRs) using fluoride salts. This concept, operated in the Thorium fuel cycle, may be started either with 233U, enriched U and/or TRU elements as initial fissile load. It has been recognized as a long term alternative to solid fuelled fast neutron systems with a unique potential (such as large negative temperature and void coefficients, lower fissile inventory, no initial criticality reserve, simplified fuel cycle, wastes reduction…) and is thus one of the reference reactors of the Generation IV International Forum. This paper will focus on recommendations to define a demonstrator representing the key points of the reference MSFR power reactor (3000 MWth, fuel salt volume of 18 m3). The MSFR demonstrator is designed to assess the technological choices of this innovative system (fuel salt, structural materials, fuel heat exchangers…). It seems finally possible to slightly modify such a demonstrator which could then be a self-breeder modular reactor. (author)

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

    International Nuclear Information System (INIS)

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

  4. Core Function Changes from a Breakeven Core to a TRU Burner Core

    International Nuclear Information System (INIS)

    A 600MWe sodium cooled fast reactor named as KALIMER-600 has been developed with a single enrichment fuel. This reactor is a pool-type reactor with a 1,523MW thermal power. The core is loaded with a ternary metallic fuel of 15 w/o TRU enriched TRU-U- 10Zr and it is designed to have breakeven breeding characteristics (CR∼1.0). However, a new demand is how to solve a spent fuel disposal problem because nuclear spent fuel storages shall become full by 2016 year. Therefore, a TRU burner concept which can burn out spent fuel actively is needed instead of a breakeven reactor concept. After all spent fuels from LWRs are burned, another issue may be that a TRU burner can not be operated in a breakeven mode any more. In order to overcome this problem, a new concept, a core function change is proposed in this paper. A reactor will operate as a TRU burner at first and then, will play the role of a breakeven core without any core layout change which does not need TRU supply. Since the nuclear conceptual design of a breakeven core - KALIMER-600 is already finished, TRU burner concepts are based on the KALIMER-600 breakeven core and its safety parameters are asked to be compatible with those of the KALIMER- 600 breakeven core

  5. Technical meeting to 'Preserve fast reactor physics knowledge'. Working material

    International Nuclear Information System (INIS)

    The meeting extended its scope beyond reactor physics to include all the main areas of fast reactor data retrieval and knowledge preservation (FR KP). The participants presented the status of the national FR KP efforts and the progress achieved since the kick-off meeting of the IAEA initiative (meeting hosted by ANL-West in Idaho Falls, Idaho, 2-4 April 2002). Details are given in Section 2. The Scientific Secretary of the Technical Working Group on Fast Reactors (TWG-FR) presented the Agency activities (KNK II data and documentation retrieval and preservation), and recalled the Agency's role in this initiative: - Coordination of the national efforts - Ensuring the collaboration with other International Organizations (mainly OECD/NEA) - Establishing and maintaining the access means to the ultimate goal of the initiative, the 'fast reactor knowledge base'. The integration of specific activities relevant to the FR KP initiative, which are planned within the framework of the TWG-FR, was discussed. It was agreed to implement the following as TWG-FR tasks with clear relevance to FR KP initiative: - Japanese 'Proposal from Monju relevant to Fast Reactor Knowledge Preservation Activity in the framework of the IAEA TWG-FR' - Proposal of a CRP on 'Generalization and Analyses of Operational Experience with Fast Reactor Equipment and Systems' - TM on 'Handling of Sodium Coming from Decommissioned Fast Reactors and from the Shutdown of Experimental Facilities' (if not already covered by the TECDOC being prepared by IAEA's Nuclear Waste Technology Section). While the responsibility for fast reactor knowledge preservation, data retrieval and interpretation, as well as quality assurance will rest with the individual Member States joining the FR KP initiative, the participants confirmed the Agency's role (see above). More specifically, the participants in the meeting recommended that the IAEA - support and coordinate data retrieval and interpretation efforts by the fast reactor

  6. Reactor shutdown system of prototype fast breeder reactor

    International Nuclear Information System (INIS)

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

  7. Sodium fast reactor power monitoring using gamma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  8. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    G Pandikumar; V Gopalakrishnan; P Mohanakrishnan

    2009-05-01

    In a thermal neutron reactor, multiple recycle of U–Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving 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 reactor under construction at Kalpakkam. After about five recycles, the cycle-to-cycle variation in the above parameters is below 1%.

  9. Sodium fast reactor power monitoring using gamma spectrometry

    International Nuclear Information System (INIS)

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

  10. A study of passive safety conditions for fast reactor core

    International Nuclear Information System (INIS)

    A study has been made for passive safety conditions of fast reactor cores. Objective of the study is to develop a concept of a core with passive safety as well as a simple safety philosophy. A simple safety philosophy, which is wore easy to explain to the public, is needed to enhance the public acceptance for nuclear reactors. The present paper describes a conceptual plan of the study including the definition of the problem a method of approach and identification of tasks to be solved

  11. Liquid metal collector and separator device in a fast reactor

    International Nuclear Information System (INIS)

    The invention applies to a fast reactor including a rotational symmetrical vessel around a vertical axis, a core immersed in liquid metal filling the vessel as also liquid metal circulating pumps and heat exchangers immersed in this liquid metal, and arranged vertically in the vessel nearly at nearly the same distance from the axis of this one. The present device, described in detail in this patent, allows to simplify and to get a less expensive internal structure for the reactor vessel and to improve its resistance to earthquakes

  12. European ERANOS formulaire for fast reactor core analysis

    International Nuclear Information System (INIS)

    ERANOS code scheme was developed within the European collaboration on fast reactors. It contains all the functions required to calculate a complete set of core, shielding and fuel cycle parameters for LMFR cores. Nuclear data are taken from recent evaluations (JEF2.2) and adjusted on integral experiments (ERALIB1). Calculational scheme uses the ECCO cell code to generate cross section data. Whole core calculations are carried out using the spatial modules BISTRO (Sn) and TGVNARIANT (nodal method). Validation is based on integral and power reactor experiments. Integral experiments are also used for adjustment of nuclear data

  13. OECD Nuclear Energy Agency Activities Related to Fast Reactor Development

    International Nuclear Information System (INIS)

    • Despite impact of Fukushima, there remains a high level of interest in continued development of advanced nuclear systems and fuel cycles: – better use of natural resources; – minimisation of waste and reduction of constraints on deep geological repositories. • Ambitious R&D programmes on-going at national level in many countries, also through international projects: – expected to lead to development of advanced reactors and fuel cycle facilities. • OECD/NEA will continue to support member countries in field of fast reactor development and related advanced fuel cycles: – forum for exchange of information; – collaborative activities

  14. The SCARABEE experimental fast reactor safety programme already completed

    International Nuclear Information System (INIS)

    The SCARABEE in-pile experimental programme comprised a series of tests on unirradiated fuel pins, either single or in seven-pin clusters. The main objective was to obtain information on the mode and consequences of fast reactor fuel pin failure in conditions representative of loss of cooling in a LMFBR. The application of such programmes in full scale reactors leads to the great importance of the interpretation of experimental observations. The interpretation of that programme was carried out jointly by CEA, KFK and UKAEA; this international collaboration led to a sharper focusing on essential features to be modelled in experiments and computer codes and to a valuable convergence of views

  15. Performance of fast reactor irradiated fueled emitters at goal burnup

    International Nuclear Information System (INIS)

    UO2-fueled W emitters were examined that had been irradiated to goal burnups of approximately 4 at.% at emitter surface temperatures to 1820 K in a fast reactor to establish their performance for use in thermionic reactors with power levels from tens of kilowatts to multimegawatts. The examinations provided first-time data on structural integrity, dimensional stability, component compatibility, and fuel and fission product behavior. The data are consistent with similar measurements at approximately 2 at.% burnup with the exception of one emitter which breached the W during irradiation

  16. Criteria for structural verification of fast reactor core elements

    International Nuclear Information System (INIS)

    Structural and functional criteria and relative verifications of PEC reactor fuel element are presented and discussed. Particular attention has been given to differentiate the structural verifications of low neutronic damage zones from those high neutronic damage ones. The structural verification criteria, which had already been presented at the 8th SMIRT Seminar Conference in Paris, have had some modifications during the Safety Report preparation. Finally some necessary activities are indicated for structural criteria validation, in particular for irradiated components, and for converging towards a European fast reactor code. (author). 3 refs, 6 tabs

  17. A review of the fast reactor programme in Japan

    International Nuclear Information System (INIS)

    The FBR programme in Japan has shown a steady progress, Reactor Joyo commenced the 17th duty cycle operation with MK-III core. Monju construction was 63.5% complete as of the end of February 1989, including design, manufacturing and construction at the site. Overall site work is now 89% complete. In 1988 JAPC evaluated the Demonstration Fast Breeder Reactor (DFBR) plant maintainability on both pool design and Loop Design. In 1989 JAPC is expected to start the conceptual design for the demonstration of FBR. (author). Figs, 1 tab

  18. Status of fast reactors and ADS programmes in France in 2001

    International Nuclear Information System (INIS)

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

  19. Analysis of thorium/U-233 lattices and cores in a breeder/burner heavy water reactor

    International Nuclear Information System (INIS)

    Due to the inevitable dwindling of uranium resources, advanced fuel cycles in the current generation of reactors stand to be of great benefit in the future. Heavy water moderated reactors have much potential to make use of thorium, a currently unexploited resource. Core fuelling configurations of a Heavy Water Reactor based on the self-sufficient thorium fuel cycle were simulated using the DRAGON and DONJON reactor physics codes. Three heterogeneously fuelled reactors and one homogeneously fuelled reactor were studied. (author)

  20. Core Design Studies for TRU Transmutation in a Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The objectives of this research project is (1) to develop the conceptual core designs for TRU transmutation covering a large variation in power level and conversion ratio and (2) to perform relevant verification and validation analyses through the analyses of fast critical experimental assemblies. An homogeneous and detailed heterogeneous models of metal fueled critical assemblies, BFS-73-1, BFS-75-1, and BFS-55-1, were produced from this study through a review of the critical experiments. Based on these models, BFS critical assemblies were analyzed by a fast reactor analysis code system (TRANSX/ TWODANT/DIF3D) with different evaluated nuclear data files including ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, JENDL-AC2008 in addition to ENDF/B-VI.6. A study of library difference on computational results by both a conventional diffusion method and a Monte-Carlo transport method has been carried out with those models. In addition to the analysis by the design code, Monte-Carlo high fidelity simulation was carried out to support the diffusion solution, mainly an effect of unit fuel cell heterogeneity. BFS and ZPPR critical assemblies were analyzed by both KAERI and ANL systems and the results of the analyses were reviewed by the other side. This improve the reliability of the results of both institutes. For the effective TRU transmutation, the conceptual core design was performed under core power ranged from 1,500MWt to 4,500MWt and found that there is no appreciable degradation in performance or reactivity coefficients for the core power level up to 1,800 MWe and confirmed the possibility of the large scaled transmutation reactor. Even at each pre-determined power level, performance parameters, reactivity coefficients and its implication on the safety analysis can be different when a target TRU conversion ratio changes. In order to address this aspect of design, a variation study of TRU conversion ratio change was covered. Three ATWS events such as UTOP, ULOF and ULOHS are

  1. Approaches to validation of fast reactor lifetime extension

    International Nuclear Information System (INIS)

    As compared with the other reactors, the main feature of fast neutron sodium-cooled reactor operation is the effect of increased temperatures (up to 550-600 deg. C) and intensive fast neutron flux (up to ∼2·1021 n/cm2·year of energy E>0.1 MeV) on structural materials. Under these conditions, the basic mechanisms damaging fast reactor component material are creep, fatigue and their interaction as well. Under intensive neutron flux, the austenitic steel used as reactor structural material is embrittled. Except thermo-mechanical load, the additional loading factor for fast reactor components is non-uniform material swelling due to the effect of high-dose irradiation. This results in considerable deformation of reactor components that can violate their normal operation in the ultimate case. It is necessary to note that most of main fast reactor components are difficult of access for non-destructive inspection in order to detect defects. Therefore, in case of reactor service lifetime prolongation, it is necessary to take account of availability of process and operation defects in these components. In view of above-mentioned, to validate prolongation of BN-600 reactor service life up to 45 years, procedural and material-study activities were performed to develop procedures and methods of strength and lifetime analysis for structural components with defects under the effect of high temperatures and intensive irradiation, as well as to obtain service characteristics of fast reactor structural materials in view of their degradation under the effect of high temperature during more than 2.105 hours and intensive neutron irradiation. Within procedural tasks, the following main procedures and methods have been developed: 1. Definition of design dimensions and shapes of postulated defect. 2. Formulation of constitutive equations of thermal-viscoelastic-plastic deformation of structural material in view of swelling to analyze stress-strain state (SSS) of structural components

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

    International Nuclear Information System (INIS)

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

  3. Fast reactors in Russia: Status as of 2000 and prospects

    International Nuclear Information System (INIS)

    intellectual) resources. Therefore, great attention is paid in Russia to coordination and integration of international efforts in the development of nuclear technologies. At the UN Millennium Summit on September 6, 2000, the President of the Russian Federation announced the initiative on organization within the framework of an international project with IAEA participation and development of innovative reactor technology and nuclear fuel cycle with natural safety eliminating proliferation of nuclear weapons and providing incineration of plutonium and other long lived radioactive elements. Establishment of a special IAEA group on the innovative nuclear reactors and fuel cycles aimed at the analysis, choosing and development of advanced nuclear technology, is the first step in this direction. Fast neutron reactor technology is most promising from the standpoint of meeting imposed requirements. Undoubtedly, new advanced nuclear technologies will be chosen on the basis of results achieved in the existing technologies. Therefore, it is important to involve experts from other IAEA working groups, including TWGFR, in discussions and examination of advanced options. Three fast reactors are in operation in Russia in 2000: Test reactor BR-10, experimental reactor BOR-60 and prototype reactor BN-600. Current status and basic areas of design studies of fast reactor technology are described. Research related to accelerator driven systems in Russia include experimental studies of accelerator driven system parameters and analytic studies, and computer codes development

  4. Project and characteristics of a 5MW experimental fast reactor

    International Nuclear Information System (INIS)

    Characteristics of a 5 MW experimental fast reactor are reported. The reactor is designed with emphasis on fuel and materials irradiation and uses fuel assemblies of a standard structure. The reference core consist of 37 fuel assemblies, each of which contains 19 pins of metallic Pu/Zr fuel. With a core height of 17.6 cm the core volume is 11.4 liter and the central fast (E >=100 KeV) flux is 0.9 x 1015 n/cm2 sec. In addition to twelve control rod assemblies with a total reactivity worth of 5.5% Δk, 42 assemblies for reactivity compensation are placed in the two rings outside the core. Replacing these assemblies with driver, blanket, or refletor-shield assemblies, large reactivities can be added to make the central assembly position available for test irradiations and to assure high levels of burnup of driver assemblies. (Author)

  5. European Union: Review of fast reactor related activities

    International Nuclear Information System (INIS)

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

  6. Thermoelectric direct energy conversion system for fast reactors

    International Nuclear Information System (INIS)

    A concept of direct energy conversion system for fast reactors has been presented by using FGM thermoelectric (TE) cell elements combined with FGM compliant pads based on the assumption that energy conversion efficiency of 20% could be achieved. The design involves TE energy conversion modules in which 360 TE cell elements are assembled. These energy conversion modules are connected to sodium and water circuits by cesium heat pipes and water heat pipes, respectively. The following design approach has been demonstrated. 1) Approximately 4100 energy conversion modules installed in a 150 MWt fast reactor can affords 27 MW of electricity. 2) An energy conversion building (single floor, 15 m x 15 m) enables to eliminate intermediate heat exchangers, steam generators and sodium-water reaction counter measures. (author)

  7. Operating experience of fast breeder reactors in the USSR

    International Nuclear Information System (INIS)

    The operating experience results of BN-600, BN-350, BOR-60 and BR-10 fast breeder reactors are presented. The fast reactors design and operation experience in the USSR has demonstrated their high operational qualities, safety, reserves of improvement. After 11 years' operation the BN-600 and 18 years' operation the BN-350 these two nuclear plants present a very satisfactory global loading rate of above 65%. The operation flexibility of the nuclear power plants and, in particular, the possibility of operation at 2/3 nominal power (BN-600) and at 4/5 and/or 3/5 nominal power (BN-350) have allowed for these loading rates to be reached in spite of numerous steam generators and pumps replacement. (J.P.N.)

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

    International Nuclear Information System (INIS)

    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

  9. Core Seismic Tests for a Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2007-01-15

    This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

  10. Fast current pulse amplifier for neutron flux monitoring system of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The neutron flux monitoring system (NFMS) for Prototype Fast Breeder Reactor (PFBR) measures the neutron power and the reactivity changes in the core in all the states such as shut down, fuel handling, reactor startup, intermediate and power ranges using high temperature cylindrical fission chambers, four section fission counter and high temperature boron coated counter. Fast Current Pulse Amplifier has been developed to use in NFMS of PFBR that amplifies single/four numbers of input current pulses independently, discriminates and electronically wire - OR them to give differential pulse output along with the Campbell output. The paper describes the design, development of integrated single/Quad channel fast current pulse amplifier based on in-house developed ASIC, Hybrid IC, in built test features, LV and HV supplies. (author)

  11. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  12. Teaching Sodium Fast Reactors in CEA-INSTN

    International Nuclear Information System (INIS)

    Conclusion: Education and Training: - a key element for the future of the development of Sodium Fast Reactors, and more particularly ASTRID project. - a tool to create a new generation of skilled nuclear engineers in the field. - a unique mean to share basic knowledge, operational feedback, safety approaches. The two entities aimed to deliver Training sessions, i.e. Sodium School in Cadarache, and INSTN-Cadarache, are ready: - to conceive and propose tailored sessions, - to collaborate with other foreign Education and Training Entities

  13. A review of the UK fast reactor programme. April 1992

    International Nuclear Information System (INIS)

    Total energy consumption in the UK in 1991 was 351.6 million tones of coal or coal equivalent, an increase of 2.1% on 1990. Nuclear electricity accounted for 19.5% of the total electricity consumption of about 300 TWh. The technical part of the report is principally concerned with progress with the Prototype Fast Reactor (PFR) and its associated fuel reprocessing plant and with some aspects of international cooperation on fast reactors. The total gross electrical generation of PFR for 1991 was 34,767 MWd, equivalent to annual load factor of 41.6%. The principal factor depressing the load factor figure was an ingress of lubricating oil from bearing on primary sodium pump 2 into the primary coolant which led to the station being out of service for six months. Two PFR fuel reprocessing campaigns were undertaken during the year. In the first, 18 subassemblies at burnup levels up to 12%, plus some loose pins from the fuel post-irradiation examination facility, were processed. In the second, a further 7 subassemblies at burnup levels up to 17.3%, plus some more loose pins were dealt with. The cumulative total amount of fuel reprocessed to date is now 17.99 tons of heavy metal, containing 3.17 tonnes of plutonium. The reduction of Government funding to the fast reactor research and development programme since 1989 has led to termination of fuel cycle research and development work. However, valuable information continues to be obtained from operation of the PFR fuel reprocessing plant and its support facilities and from development work on the manufacture of thermal MOX fuel. Information exchanges and cooperative work programmes conducted under the UKAEA's agreements with Japan (PNC and JAERI), the USA (US Department of energy), and the CIS are now coordinated with those of the UKAEA's European Fast Reactor research and development partners

  14. Specialists' meeting on fast reactor cover gas purification

    International Nuclear Information System (INIS)

    The tentative agenda was adopted by the participants without comment and was followed throughout the meeting. The following topics were discussed at the subsequent sessions of the meeting on 'Fast Reactor Cover Gas Purification': National Position Papers; Impurities: Sources and Measurement; Cover Gas Purification Techniques; Sodium Aerosol Trapping; Radiological Considerations. Based on the papers presented and the discussions following, session summaries and conclusions were prepared and are included in this report

  15. Evaluation of the breed/burn fast reactor concept

    International Nuclear Information System (INIS)

    A core design concept and fuel management strategy, designated breed/burn, has been evaluated for heterogeneous fast breeder reactors. In this concept internal blanket assemblies after fissile material is bred in over several incore cycles, are shuffled into a moderated radial blanket and/or central island. The most promising materials combination identified used thorium in the internal blankets (due to the superior performance of epithermal Th-U233 systems) and zirconium hydride (ZrH16) as the moderator

  16. Fast neutron reactor fuel elements and power grid duty cycling

    International Nuclear Information System (INIS)

    The PHENIX power grid cycling operation in 1982-1983 will allow verification of the models and criteria developed in the interim. It will provide indispensible statistical data and will open the way to power grid duty for Super PHENIX beginning in 1986. Although at the present time it is impossible to resolve the question of weekly or daily load variations, it is felt that fast neutron reactor fuel subassemblies should provide satisfactory performance for primary and secondary frequency adjustments

  17. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    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

  18. Knowledge management: knowledge and competence maintaining; problematics; Fast reactor example

    International Nuclear Information System (INIS)

    In this expose are examined two representative aspects, with a first part reserved to the general problematics of the knowledge and competence maintaining as it looks at Framatome, in its activities of pressurized water boiler supplier (PWR) and as provider of nuclear services, and a second part treating the solutions used by the different actors intervening in fast neutrons reactors, that is to say, the Cea, Framatome and EDF. (N.C.)

  19. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    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

  20. A methodology of neutronic-thermodynamics simulation for fast reactor

    International Nuclear Information System (INIS)

    Aiming at a general optimization of the project, controlled fuel depletion and management, this paper develop a neutronic thermodynamics simulator, SIRZ, which besides being sufficiently precise, is also economic. That results in a 75% reduction in CPU time, for a startup calculation, when compared with the same calculation at the CITATION code. The simulation system by perturbation calculations, applied to fast reactors, which produce errors smaller than 1% in all components of the reference state given by the CITATION code was tested. (author)

  1. Role of small lead-cooled fast reactors for international deployment in worldwide sustainable nuclear energy supply

    International Nuclear Information System (INIS)

    Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. Meeting future worldwide projected energy demands during this century (e.g., 1000 to 2000 GWe by 2050) in a sustainable manner while maintaining CO2 emissions at or below today's level will require massive deployments of nuclear reactors in non-fuel cycle states as well as fuel cycle states. The projected energy demands of non-fuel cycle states will not be met solely through the deployment of Light Water Reactors (LWRs) in those states without using up the world's resources of fissile material (e.g., known plus speculative virgin uranium resources = 15 million tonnes). The present U.S. policy is focused upon domestic deployment of large-scale LWRs and sodium-cooled fast spectrum Advanced Burner Reactors (ABRs) working in a symbiotic relationship that burns existing fissile material while destroying the actinides which are generated. Other major nuclear nations are carrying out the development and deployment of SFR breeders as witness the planning for SFR breeder deployments in France, Japan, China, India, and Russia. Small (less that 300 MWe) and medium (300 to 700 MWe) size reactors are better suited to the growing economies and infrastructures of many non-fuel cycle states and developing nations. For those deployments, fast reactor converters which are fissile self-sufficient by creating as much fissile material as they consume are preferred to breeders that create more fissile material than they consume. Thus

  2. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  3. Delayed gamma power measurement for sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Graphical abstract: Display Omitted Research highlights: →20F and 23Ne tagging agents are produced by fast neutron flux. →20F signal has been measured at the SFR Phenix prototype. → A random error of only 3% for an integration time of 2 s could be achieved. →20F and 23Ne power measurement has a reduced temperature influence. → Burn-up impact could be limited by simultaneous 20F and 23Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,α) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  4. Optimization of ultra-long cycle fast reactor core

    International Nuclear Information System (INIS)

    An optimization of an ultra-long cycle fast reactor (UCFR) design with a power rate of 1000 MW (electric), UCFR-1000, has been performed to increase the safety of UCFR. Firstly, geometric optimization has been performed to decrease its peaking factors so that the peak temperatures measured by thermal hydraulic feedback are within the limit of design basis event (DBE). Secondly, fuel composition optimization has been performed by adopting Pressurized Water Reactor (PWR) spent fuel as a blanket material instead of natural uranium. Lastly, a small-size UCFR with a power rate of 100 MWe, UCFR-100, has been proposed for developing a short term deployable nuclear reactor. The major optimization process for UCFR-100 is decreasing maximum neutron flux and fast neutron fluence. The optimized UCFR-1000 has been enlarged radially and shortened axially from the initial UCFR design and this modification makes the burning speed of active core movement slower. It has been confirmed that a full-power operation of 60 years without refueling is feasible for both UCFR-1000 and UCFR-100 core designs by a breed-and-burn strategy. By the design optimization study, the reductions of maximum neutron flux, fast neutron fluence, and axial power peaking have been achieved, which are favorable for the safety of the UCFR. (author)

  5. Application of nitrogen alloyed steels for Indian Fast Reactor programme

    International Nuclear Information System (INIS)

    Towards building fast reactors for fulfilling energy requirements through second stage of nuclear power program planned by Department of Atomic Energy, a 500 MWe Prototype Fast Breeder Reactor (PFBR) is under advanced stage of construction at Kalpakkam, a coastal site. Nitrogen alloyed types 304LN and 316LN austenitic Stainless Steels have been selected for out of core components except for the steam generator primarily due to inclusion in the design codes favourable effect of nitrogen on mechanical strength and sensitization, and excellent weldability. For the once through steam generator design selected from economics and safety, modified 9Cr-1 Mo (Gr 91) has been selected from inclusion in the design codes, adequate mechanical strength, sound industrial experience and carbon transfer considerations. The presentation highlights the application of nitrogen alloyed types 304LN and 316LN SS, as well as modified 9Cr-1Mo steel for PFBR, and the influence of increased nitrogen alloying on mechanical properties on SS 316L for application to future fast reactors. (author)

  6. Radiological and environmental aspects of fast reactor fuel cycle facilities

    International Nuclear Information System (INIS)

    Availability of energy is an important prerequisite for the socioeconomic development of any country. As the sources of fossil fuels are dwindling fast, India will have to look for nuclear power to secure a stable supply of energy. The Indian nuclear power program aims at large scale utilization of its vast thorium resources. The energy potential of uranium increases by 150 times and that of thorium by three times through the fast breeder reactor route compared to thermal reactors. This long term objective of thorium utilization is sought to be achieved through three stages of development. In the first stage a series of PHWRs will be constructed for power generation which will incidentally generate plutonium. The second stage consists of FBRs with plutonium as the fuel and thorium as the blanket. In the third stage, U-233 will replace plutonium as the fuel for FBRS. It is interesting to compare the radiological and environmental safety aspects of fast reactor fuel cycle involving U-Pu and Th-U

  7. Status of fast breeder reactor development in Germany

    International Nuclear Information System (INIS)

    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)

  8. Romanian Contribution to the Development of Lead Cooled Fast Reactors

    International Nuclear Information System (INIS)

    In Romania the nuclear energy is considered an important component of energy mix and for a country sustainable development. Presently based on PHWR CANDU technology, the research and development activities, part of the national nuclear power programme, provide an increased effort towards generation 4, dedicated to support fast reactor lead technology. In the European framework (EU R&D Framework Programmes, European Sustainable Nuclear Industry Initiative) devoted to the development of GenerationIV technologies, Romania is contributing as a partner in EU R&D projects together with a large number of EU research organizations and in the ALFRED MoU, having ANSALDO Nucleare, ENEA and INR as initial members. ALFRED (Advanced Lead Fast Reactor European Demonstrator) project aims to build a 125MWe lead cooled fast reactor demonstrator, connected to the electrical network, with a target date for operation start-up in 2025. In February 2011 Romanian Government approved the option to host ALFRED demonstrator. Based on the access to European structural funds, existing nuclear experience and EU orientation to build the demonstrators in the new member states, Romania is an important option for siting process. An investigation of the existing national capabilities, identification of additional infrastructure and identification of the professional development needs in order to prepare the siting national support are presented in the paper. The main approaches and needed resources to meet expected requirements for ALFRED implementation are discussed as well. (author)

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

    International Nuclear Information System (INIS)

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

  10. Sensitivity analysis for actinide production and depletion in fast reactors

    International Nuclear Information System (INIS)

    In sensitivity analysis of the actinide production and depletion in fast reactors, a mathematical method of calculating sensitivity coefficients is improved and simplified by combining the time-dependent generalized perturbation technique with the eigenvalue method. Numerical calculations show that the eigenvalue method is well applicable in solving the nuclide chain equation and its adjoint equation and the cylic chains in the decay scheme of the actinides can be interpreted by means of complex eigenvalues. The sensitivity coefficients of actinide production and depletion in a 1000 MWe fast reactor are strongly dependent on the type of Pu fuel used, i.e. Pu fuel from BWR or Pu fuel from the blanket of FBR. The sensitivity coefficients due to variations of capture cross sections, σsub(n,2n) of 238U, lambda sub(β) of 241Pu and lambda sub(α) of 242Cm are especially large. Sensitivity analyses for the 1000 MWe fast reactors show that higher priorily should be given to decay constants of 241Pu and 242Cm, capture cross sections of 237Np, 241Am, 243Am and 242Pu, and fission cross sections of 237Np, 242Pu, 241Am and sup(242m)Am. (author)

  11. Immediate relation of ING to fast breeder reactor programs

    International Nuclear Information System (INIS)

    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)

  12. A review of the Italian fast reactor programme, March 1979

    International Nuclear Information System (INIS)

    Year 1978 in Italy was marked by a standstill in the nuclear energy field. The decisions previously made for the installation of eight 1000 MWe LWR-type reactors could not be acted upon because of the opposition of local authorities and lack of Government power. The construction site at Montalto di Castro (two BWR reactors) was ope ned with difficulty, whereas the decision to install a plant in Mo use equipped with two PWR reactor was postponed. The new presidents of ENEL and CNEN were appointed in January this year and the appointments of the new Boards of Directors are underway. With regard to CNEN, many political bodies are in agreement on an institutional change which would widen field of activity to include new energy sources, solar energy in particular. This will open a big problem: if CNEN will be no more a 'nuclear body, it could be necessary to transfer all the activities connected to the Regulatory Commission to another separate body to be instituted. In this context, the fast reactor programme has continued to develop under the directives of CIPE, and has concentrated its effort on the following three objectives: the PEC-Reactor, the Creys-Malville Power Plant and research and development, and industrial promotion. These objectives are being pursued with the participation of CNEN, ENEL and Italian industry. CNEN has the role of committing and operating the PEC reactor; it is also charged to perform part of the R and D Italian-French programme and to promote industrial development. ENEL participates in the NERSA Company, owner of the Creys-Malville Plant. Italian industry, with its activities of architect-engineering, designing and manufacturing will participate in the construction of the PEC and of the Italian part (33%) of the Creys-Malville Plant. During the last months of 1978 a consortium (COREY) was set up by CNEN and NIRA which has the purpose of integrating and ensuring the smooth running of Italian efforts in the field of long-term research and

  13. Multipurposed small fast reactor SVBR-75/100

    International Nuclear Information System (INIS)

    Currently the nuclear power (NP) development meets significant difficulties in many countries. First of all it relates to complicating and cost rising of nuclear power plants (NPP) due to essential enhancing the safety requirements. The possibility and expediency of developing the NP based on unified small power reactor modules SVBR-75/100 with fast neutron reactors cooled by lead-bismuth eutectic alloy is substantiated for the nearest decades in the paper. Based on those modules the following designs can be realized: renovating of the NPP units which operation term has been exhausted; regional nuclear heat power plants (NHPP) of 100-300 MW power which need near cities' location; large power modular NPPs (∼1000 MW) like US concept PRISM or Japanese concept 4S; nuclear power complexes for sea water desalinating in developing countries which meet nonproliferation requirements, reactors for Pu utilization and minor actinides transmutation. (author)

  14. A review of the fast reactor programme in Japan

    International Nuclear Information System (INIS)

    In Japan the experimental reactor ''Joyo'' has provided abundant experimental data and excellent operational records attaining 40,000 hours operation in total by the end of 1989, since its first criticality in 1977. On the prototype reactor ''Monju'', more than eighty percent of construction work has already been completed on schedule, aiming at the initial criticality by October 1992. As for the demonstration fast breeder reactors (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee. (author). Figs and tabs

  15. Fast Burst Reactors in the United States of America

    International Nuclear Information System (INIS)

    Early in 1953, the bare uranium metal reactor, Lady Godiva, produced self-quenching fission bursts for the first time. Since then, seven progeny of Godiva have been produced in the USA, and at least two more are nascent. Fast burst reactors which have operated in the USA are listed in Table I, which shows time of start-up, nominal fission yield per burst and full width of pulse at half maximum power. The first five employ enriched uranium metal (∼93.5% U285); the three latest models contain an alloy of uranium and molybdenum which permits operation at higher temperatures, hence higher fission densities. Essential design features of these reactors with regard to a variety of pulse irradiation applications are discussed together with mechanical limitations and some advanced design proposals. (author)

  16. A Comparison of Long-Lived, Proliferation Resistant Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-09-01

    Various methods have been proposed to transmute and thus consume the current inventory of trans-uranic waste that exists in spent light-water-reactor fuel. These methods include both critical and sub-critical systems. The neutronics of metallic and nitride fuels loaded with 20-30wt% light-water-reactor plutonium plus minor actinides for use in a lead-bismuth and sodium cooled fast reactor are discussed, with an emphasis on the fuel cycle life and isotopic content. Calculations show that core life can extend beyond 20 years, and the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from 0.5 to 0.9 g/MWd.

  17. Fast reactor calculational route for Pu burning core design

    International Nuclear Information System (INIS)

    This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted. The document includes comments on and explanations of the modelling assumptions in the various calculations. Practical information on the use of the calculational route and the computer systems is also given. (J.P.N.)

  18. Computational intelligent systems for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Nearly 15000 process signals are digitized by physically and functionally distributed embedded systems in Prototype Fast Breeder Reactor (PFBR). Digitized signals are processed and relevant information is displayed through Large video display systems at Control Room. It is necessary that correct and reliable information need to be provided to the plant operator. Computational intelligent systems play a major role in enhancing the safe operation of the Nuclear reactor. The paper explains the features of three such systems, one for on-line validation of neutronic power channel through on-line thermal balance calculation and another for detection of anomalous reactivity addition through on-line reactivity balance computation and third for on-line computation of Reactor power from fluctuations of core thermocouple signals. (author)

  19. Fast flux fluid fuel reactor: A concept for the next generation of nuclear power production

    International Nuclear Information System (INIS)

    burnups. It is possible to continuously remove the fission products and to minimize maintenance requirements. Fluid fuel systems possess favorable in-core transient response via a very high immediate negative temperature coefficient because of the expansion of the liquid fuel. Control rods are not necessary because the loss of reactivity can be compensated for by adding fuel in the on-line circuit. The main challenges posed by fluid fuel systems are possible fluctuations of reactivity caused by density changes, loss of delayed neutrons in the fuel leaving the core for the on-line reprocessing circuit, and corrosion and erosion of the containers. The fast flux choice is dictated by the much better neutronic economy offered by a hard spectrum system. The system is flexible enough to be either a burner, a converter, or a breeder. The fast spectrum is the only one that will allow all the transuranics to be efficiently burned. Moreover, because of the very high operating temperatures (1,000 C or more), refractory metals have to be used for the container. These materials have quite large absorption cross sections in the thermal and epithermal range. Therefore, they can be used only in a hard spectrum system without compromising the neutronic efficiency. Two different types of fast flux fluid fuel reactors are being considered: liquid-metal fluid fuel reactors and molten salt reactors

  20. RF torch discharge combined with conventional burner

    International Nuclear Information System (INIS)

    The design of the combined flame-rf-plasma reactor and experimental examination of this reactor are presented. For the determination of the temperature in different parts of the combined burner plasma the methods of emission spectroscopy were used. The temperatures measured in the conventional burner reach the maximum temperature 1900 K but in the burner with the superimposed rf discharge the neutral gas temperature substantially increased up to 2600 K but also the plasma volume increases substantially. Consequently, the resident time of reactants in the reaction zone increases

  1. The fast reactor with a cold bottom vessel

    International Nuclear Information System (INIS)

    2 kinds of design have been proposed so far for pool-type fast reactors where the vessel contains the full primary circuit with pumps and heat exchangers. First, the vessel suspended to the concrete vault top, a rounded bottom supporting the core (Phenix, Superphenix or the European Fast Reactor Project), and secondly, the half resting vessel, the upper cylindrical wall resting on a ring supported by the basement (BN-600, BN-800 Russian reactors)). A new kind of design is developed in this paper: the cold resting bottom vessel where the flat vessel bottom cooled at less than 90 C. degrees rests on the concrete basement and supports the core dia-grid above a sodium in thermal gradient, by vertical shells and by the vessel wall. This space contains heavy and refractory materials, preventing bottom piercing by corium, hot mix flow of fuel and steel in case of core fusion accident. A concrete vault contains the reactor under ground level. Resting cylindrical vessel provides stability and robustness for core and internal components. A toric collector welded around the dia-grid receives vertical exchangers and pumps, homogenizing temperature. Cooling of structures is not needed. Sodium leaks through the bottom plate are prevented by sodium freezing. Exchangers and pumps are stable, maintained at roof and collector levels. Reduced thermal exchanges provide better efficiency. Arrest exchangers, located around pump columns above the collector, allow a reduced vessel diameter. Simplified structures provide lesser weight and cost. The cold resting bottom may be applied to loop reactors and to lead-cooled reactors

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

    International Nuclear Information System (INIS)

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

  3. Comparative breeding characteristics of fusion and fast reactors.

    Science.gov (United States)

    Fortescue, P

    1977-06-17

    Expressions are developed to allow ready comparison of a hybrid fission-fusion plant and a fast breeder with respect to the number of thermal reactors that their fissile production could support, both for their feed requirements and for the new inventory needs of an expanding industry. These relations are expressed in terms of the neutron multiplication factor obtained in the fusion blanket, and the analogous quantities represented by the conversion ratios of the fast and thermal fission associated with the comparison. Results are presented graphically both for the steady state and for industries of arbitrary growth rate, and include the influence of tritium production requirements. Even a modest blanket neutron multiplication factor could enable the hybrid fusion system greatly to outperform the fast breeder on this simple basis of material balances. PMID:17831749

  4. 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)

    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

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

    International Nuclear Information System (INIS)

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

  6. Some results of studying of spatial kinetics in fast reactors

    International Nuclear Information System (INIS)

    The paper presents an analysis of the solution to the spatial non-stationary equation of neutron transport the example of a fast reactor. To solve the spatial kinetics equation in three-dimensional geometry in diffusion approximation a calculated code TIME was created. In a direct problem the neutron flux density and its derivatives (for example, reactor power) are determinate at each time step. Solution of a direct problem allows to investigate a deformation of a steady neutron field form which is one of the main reasons of spatial effects occurrence in a reactor. Such deformations affect spatial distribution of delayed-neutrons that is reflected in behavior of reactivity and in related with it results of experiments. The paper also considers some questions of reducing of spatial effects through a choice of a point of reactivity input or detector location. In the inverse problem the reactor reactivity is calculated using the known dependence of reactor power on time. The paper describes some problems of solving the inverse problem taking into account spatial effects. (author)

  7. The future role of fast reactors and the EFR project

    International Nuclear Information System (INIS)

    After thirty years of development, the European fast reactor technology is available for industrial applications. It is still necessary to complete the demonstration and some improvements must be achieved but there is no reason to discard it when it offers one of the few possibilities, perhaps the best one, to solve the problems which remain at the back end of the fuel cycle. One could prefer to rely on the direct disposal of spent fuels coming from the light water reactors, but nobody can guarantee that such a policy would be acceptable in most of the countries using nuclear energy for electricity production. Intermediate storages, without any clear definition of the future, would neither be easy to justify. Light water reactors themselves can be used for recycling plutonium. It is already an industrial reality. Technically speaking, there is no precise limit in the number of cycles nor on the list of isotopes which can be loaded in such reactors. But it is clear that they are not well suited for burning degraded plutonium and minor actinides. Thus, a policy relying only on light water reactors, letting aside the possible difficulties with uranium supply in the future, would not be easily consistent with the goal of reducing long lived wastes. (orig.)

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

    Indian Academy of Sciences (India)

    Baldev Raj; S L Mannan; P R Vasudeva Rao; M D Mathew

    2002-10-01

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of fissile material than in thermal reactors, with a matching increase in burn-up. The design of the fuel is an important aspect which has to be optimised for efficient, economic and safe production of power. FBR components operate under hostile and demanding environment of high neutron flux, liquid sodium coolant and elevated temperatures. Resistance to void swelling, irradiation creep, and irradiation embrittlement are therefore major considerations in the choice of materials for the core components. Structural and steam generator materials should have good resistance to creep, low cycle fatigue, creep-fatigue interaction and sodium corrosion. The development of carbide fuel and structural materials for the Fast Breeder Test Reactor at Kalpakkam was a great technological challenge. At the Indira Gandhi Centre for Atomic Research (IGCAR), advanced research facilities have been established, and extensive studies have been carried out in the areas of fuel and materials development. This has laid the foundation for the design and development of a 500 MWe Prototype Fast Breeder Reactor. Highlights of some of these studies are discussed in this paper in the context of our mission to develop and deploy FBR technology for the energy security of India in the 21st century.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  11. Liquid Metal Fast Breeder Reactor program. Volume III. Environmental statement

    International Nuclear Information System (INIS)

    The various alternative technologies, nuclear as well as nonnuclear, that might be utilized in conjunction with or instead of the LMFBR to satisfy the Nation's future electric power requirements are examined. The options considered include the further implementation of various types of nuclear power reactors such as the already existing light water reactor and high temperature gas-cooled reactor, as well as the development of alternative breeder reactors such as the gas-cooled fast reactor, light water breeder reactor and molten salt breeder reactor. The development of another potential nuclear energy system, controlled thermonuclear fusion, is also addressed. The possibilities of increased emphasis on the use of conventional fossil fuels, namely coal, oil and natural gas, and the development of unconventional fossil fuels such as oil shale and domestic tar sands are discussed, followed by consideration of the further development of additional nonnuclear energy sources such as hydroelectric power systems, geothermal energy, solar energy, and other potential sources of power. Each option is examined as to the extent of its energy resource base, the research and development program that would be required (if any) to bring the option into commercial use, the environmental implications of its utilization and the costs and benefits associated with its use, in order to assess its capability for satisfying projected energy requirements. The use of improved energy conversion and storage devices such as gas turbines, fuel cells and magnetohydrodynamics is discussed. An examination of the various elements of a potential national effort in energy conservation to assess their capabilities for reducing projected energy demands and thereby replacing partially or entirely the need for additional power sources such as the LMFBR is presented. (U.S.)

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

    International Nuclear Information System (INIS)

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232Th and 233U 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 239Pu and 233U. • 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 232Th 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 235U and nuclear waste management issues. A mixture of thorium and uranium (232Th + 233U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239Pu. Initial feed of 233U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233U a theoretical evaluation has been used to derive the data for the source of 233U. 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 233U as an alternate to 235U and an alternate use for light water reactor used fuel as a blanket for

  13. Review of fast reactor activities at OECD (NEA), March 1979

    International Nuclear Information System (INIS)

    continue to rise. These facts argue strongly not only for the continued development and commercial demonstration of the breeder, but also for the supporting services required to ensure its success. This report contains analysis of the following issues: gas cooled reactors; safety of LMFBRs, fast reactor physics; fast reactor nuclear data, and activities of the NEA Data Bank

  14. RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao; Cliff Davis

    2008-06-01

    An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heat from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet

  15. Conceptual design of laser fusion reactor KOYO-fast - Concepts of reactor system and laser driver

    Energy Technology Data Exchange (ETDEWEB)

    Kozaki, Y.; Miyanaga, N.; Norimatsu, T.; Nakatsuka, M.; Jitsuno, T.; Fujita, H.; Kawanaka, J.; Tsubakimoto, K.; Fujimoto, Y.; Lu, J. [Osaka Univ., Institute of Laser Engineering (Japan); Soman, Y. [Mitsubishi Heavy Industry (Japan); Hayashi, T. [Japan Atomic Energy Research Institute (Japan); Furukawa, H. [Institute for Laser Technology (Japan); Yoshida, K.; Nakano, H.; Kubomura, H.; Kawashima, T.; Nishimae, J.; Suzuki, Y.; Tsuchiya, N.; Kanabe, T.; Matsuoka, S.; Ikegawa, T.; Owadano, Y.; Ueda, K.; Tomabechi, K. [IFE Forum (Japan)

    2006-06-15

    We have carried out the design studies of KOYO-Fast laser fusion power plant, using fast ignition cone targets, diode-pumped solid state lasers, and LiPb liquid wall chambers. Using fast ignition targets, we could design a middle sized 300 MWe reactor module, with 200 MJ fusion pulse energy and 4 Hz rep-rates, and 1200 MWe modular power plants with 4 reactor modules and a 16 Hz laser driver. The liquid wall chambers with free surface cascade flows are proposed for cooling surface quickly enough to a 4 Hz pulse operation. We examined the potential of Yb-YAG ceramic lasers operated at 150 - 225 K for both implosion and heating laser systems required for a 16-Hz repetition and 8 % total efficiency. (authors)

  16. Fast neutron reactions and fast neutron flux in the NRX reactor

    International Nuclear Information System (INIS)

    This report deals with fast neutron reactions and fast neutron flux in NRX. By fast neutrons it is meant those neutrons having an energy from ∼ 1 Mev to ∼ 25 Mev. The report is divided into three parts. In the first part measurements of (n,2n) cross sections in different irradiation positions of NRX are described. It is shown that from these experimental data, any (n,2n) cross section in NRX can be estimated. In the second part, measurements of the fast neutron flux in different irradiation positions of NRX are described. In the third part the values of the fast neutron flux found in Part II are used to estimate a variety of (n,p) and (n,a) cross sections in NRX. Fast neutron reactions in a reactor have been studied by many different workers. However, this report is not intended to give an exhaustive bibliography on fast neutron research; it will only refer the reader to a few publications where a great deal of information on fast neutron reactions as well as references to earlier work can be found, There is an excellent chapter on fast neutron research in the book of Hughes on Pile Neutron Research (Hughes 1953). A tabulation of the published measurements of threshold reactions for fission neutrons has appeared in Nucleonics (Rochlin 1959), Mellish et al. (1958) have discussed at length flux and cross section measurements with fast neutrons. The reader is also referred to recent work on this subject published in the Canadian Journal of Physics (Roy at el. 1958; Eastwood and Roy 1959; Roy 1959). (author)

  17. Preliminary Heating Rate Evaluation of Control Assembly in Fast Reactor

    International Nuclear Information System (INIS)

    The diameter of B4C absorber pin can expand 1.85 times as the number of absorber pins is reduced from 61 to 19. However, a lager pin diameter leads higher absorber centerline temperature and cladding mid-wall temperature which are major criteria on thermal hydraulic design. The most important thing is to consider the incident neutron energy and inelastic scattering to prevent underestimating the heating rate when the heating rate is evaluated at control assembly in the fast reactor. It is appropriate to apply only absorption reaction in a thermal reactor. In fast reactor, however, the heating rate at all rod-in condition in this study is about 40% bigger when the both effects are taken into account. The heating rate at the control assembly that consists of 19 absorber pins is about 5.3% bigger than the case of 61 pins. But thermal hydraulic integrities in the both designs can satisfy the design requirements by modifying flow rate

  18. Minor actinides transmutation strategies in sodium fast reactors

    International Nuclear Information System (INIS)

    In minor actinides transmutation strategies for fast spectrum reactors, different possibilities regarding the core loading are considered. We study both homogeneous patterns (HOM) with various minor actinides (MA) content values and heterogeneous schemes (HET) with higher percentages of MA (Np, Am and Cm) at the periphery of reactor. We analyze the capability of transmutation of each design and the reactivity coefficients such as the Doppler constant, void worth and the fraction of delayed neutrons. The EVOLCODE2 code is the computational tool used in this study. It is based on MCNPX and ORIGEN/ACAB codes and allows carrying out burn-up calculations to get the isotopic evolution of fuel composition. Among the three strategies studied (HOM 2.5 %, HOM 4% and HET 20 %) for a possible design of a Sodium Cooled Fast Breeder Reactor, the one with better transmutation results is the HOM 4%, which shows higher absolute and relative values (12 Kg-MA/TWe, 29% respectively). Concerning transmutation in blankets with 20% MA content, results show a very little or no transmutation values when considering Np, Am and Cm together, though a positive small value for Np and Am is obtained

  19. Fast reactors bulk sodium coolant disposal NOAH process application

    International Nuclear Information System (INIS)

    Within the frame of the fast reactors decommissioning, the becoming of contaminated sodium coolant from primary, secondary and auxiliary circuits is an important aspect. The 'NOAH' sodium disposal process, developed by the French Atomic Energy Commission (CEA), is presented as the only process, for destroying large quantities of contaminated sodium, that has attained industrial status. The principles and technical options of the process are described and main advantages such as safety , operating simplicity and compactness of the plant are put forward. The process has been industrially validated in 1993/1994 by successfully reacting the 37 metric tons of primary contaminated sodium from the French Rapsodie experimental reactor. The main outstanding aspects and experience gained from this so called 'DESORA' operation (DEstruction of SOdium from RApsodie) are recalled. Another industrial application concerns the current project for destroying more than 1500 metric tons of contaminated sodium from the British PFR (Prototype Fast Reactor) in Scotland. Although the design is in the continuity of DESORA, it has taken into account the specific requirements of PFR application and the experience feed back from Rapsodie. The main technical options and performances of the PFR sodium reaction unit are presented while mentioning the design evolution. (author)

  20. Fabrication of metallic fuel for fast breeder reactor

    International Nuclear Information System (INIS)

    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