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Sample records for fast-reactor structural materials

  1. Overview of fast reactor structural materials programme in India

    International Nuclear Information System (INIS)

    Rodriguez, P.; Paranjpe, S.R.; Chetal, S.C.; Mannan, S.L.; Ray, S.K.; Seetharaman, V.; Srinivasan, G.

    The fast reactor structural materials activities in India comprise of the programme on the materials for the Fast Breeder Test Reactor (FBTR), the construction of which is nearing completion, and the programme on the candidate materials for the Prototype Fast Breeder Reactor (PFBR) which is now in the design stage. For the materials in use in FBTR, the main thrust has been towards detailed evaluation and documentation of long term (creep) properties of type 316 stainless steel base material in air. For the PFBR the philosophy has been to identify the candidate materials and to evolve a wider scope for the testing and evaluation programmes. The major structural component is identified as variants of type 304 stainless steel and the programmes undertaken include study of low cycle fatigue properties and environmental effects on creep and stress rupture properties. Evaluations of aging embrittlement of type 316 stainless steel base material and weldments are also in progress. The paper lists the testing programmes identified for adoption in the near future. These include creep-fatigue damage studies and fracture mechanics studies on weldments for type 304 stainless steel and testing programme on 2.25 Cr-1 Mo and 9 Cr-1 Mo steels, the identified candidate materials for steam generators. The development efforts also include a comprehensive programme on inelastic analysis procedure. (author)

  2. Overview of European Community (Activity 3) work on materials properties of fast reactor structural materials

    International Nuclear Information System (INIS)

    Wood, D.S.

    The Fast Reactor Coordinating Committee set up in 1974 the Working Group Codes and Standards, and organized its work into four main activities: Manufacturing standards, Structural analysis, Materials and Classification of components. The main purpose of materials activity is to compare and contrast existing national specifications and associated properties relevant to structural materials in fast reactors. Funds are available on a yearly basis for tasks to be carried out through Study Contracts. At present about four Study Contract Reports are prepared each year

  3. Sodium environment effects to structural materials for fast reactors

    International Nuclear Information System (INIS)

    Hasegawa, Masayoshi; Fujimura, Tadato; Kondo, Tatsuo; Okabayashi, Kunio; Matsumoto, Keishi.

    1976-03-01

    Among the material technology for liquid metal-cooling fast breeder reactors, the characteristic points are high temperature, liquid sodium as a heat medium, and high energy-high density neutron energy spectra, accordingly the secular change of materials due to these factors must be taken into the design. The project of material tests in sodium was started from the metallographical studies on corrosion and mass transfer phenomena in sodium environment, and was evolved to the tests and studies on short time strength, creep strength, fatigue strength, and embrittlement in sodium environment. Concerning the corrosion and mass transfer tests, low purity and medium purity material testing loops were employed, and the test of immersion in sodium was carried out. Domestically produced austenitic stainless steel and Cr-Mo steel were tested, and the measurement of weight change, surface inspection, and the observation of cross sectional structure were carried out before and after the immersion. The decrease of thickness due to the leaching of surface metal and the lowering of strength due to the change of composition or structure come into question only in case of very thin walled stainless tubes, and the lowering of heat transfer is negligible. Cr-Mo steel also showed good corrosion resistance in sodium, but the effect of decarbonization on the strength needs some investigation in the production specifications. (Kako, I.)

  4. U.S. fast reactor materials and structures program

    International Nuclear Information System (INIS)

    Harms, W.O.; Purdy, C.M.

    1984-01-01

    The U.S. DOE has sponsored a vigorous breeder reactor materials and structures program for 15 years. Important contributions have resulted from this effort in the areas of design (inelastic rules, verified methods, seismic criteria, mechanical properties data); resolution of licensing issues (technical witnessing, confirmatory testing); construction (fabrication/welding procedures, nondestructive testing techniques); and operation (sodium purification, instrumentation and chemical analysis, radioactivity control, and in-service inspection. The national LMFBR program currently is being restructured. The Materials and Structures Program will focus its efforts in the following areas: (1) removal of anticipated licensing impediments through confirmation of the adequacy of structural design methods and criteria for components containing welds and geometric discontinuities, the generation of mechanical properties for stainless steel castings and weldments, and the evaluation of irradiation effects; (2) qualification of modified 9 Cr-1 Mo steel and tribological coatings for design flexibility; (3) development of improved inelastic design guidelines and procedures; (4) reform of design codes and standards and engineering practices, leading to simpler, less conservative rules and to simplified design analysis methods; and (5) incorporation of information from foreign program

  5. Some aspects of the chemistry of fast reactor fuel, structural material and decontamination

    International Nuclear Information System (INIS)

    Ganesan, V.

    2012-01-01

    The chemistry of materials pertaining to fast reactors is both fascinating and challenging considering the nature of materials involved such as the fuel, coolant, control and shielding materials in addition to the interactions between the structural materials and the fuel/coolant depending on the nature and conditions involved. The different chemical forms of fuel materials, the need to operate up to high burnups with consequent interactions of the fuel with clad materials, the need to close the fuel cycle by recovery of the fuel materials from spent fuels for refabrication and the necessity to manage the waste, throw a host of challenges which make their study scientifically interesting and technologically important. The use of liquid sodium as coolant in fast reactor heat transport systems combined with its inherent chemical reactivity opens up an interesting branch of chemistry involving liquid sodium especially in contact with structural materials during normal operation of the reactor and with fuels in the event of fuel pin failure. The phenomenon of sodium wetting and the associated corrosion of structural materials in contact with it combined with the need to carryout decontamination of such materials make it interesting to examine and evaluate their suitability for reuse without compromising on their structural integrity. Boron being the material of choice for control and shielding applications in fast reactors with varying isotopic enrichment and the technological challenge to produce large quantities of boron carbide makes it unique. Some of these aspects are addressed in this paper. (author)

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

    International Nuclear Information System (INIS)

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

    1978-02-01

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

  7. Results and Prospects of Development of Works on Structural Core Materials for Russian Fast Reactors

    International Nuclear Information System (INIS)

    Nikitina, A.A.; Ageev, V.S.; Leontyeva-Smirnova, M.V.; Mitrofanova, N.M.; Tselishchev, A.V.

    2015-01-01

    The strategy of development of atomic energy in Russia in the first half of XXI century contemplates construction and putting in operation of fast reactors of new generation with different types of coolant: sodium (BN-800, BN-1200, MBIR), lead (BREST-OD-300) and lead-bismuth eutectic (SVBR-100). For assurance of the working capacity of reactors that are under construction and achievement of economically reasonable burn-up of nuclear fuel the structural core materials with necessary level of radiation resistance, heat resistance, corrosion resistance to products of fuel fission, corrosion resistance in coolant and in water must be developed and justified. For sodium cooled reactors the key challenge is creation of radiation resistant and heat resistant cladding materials, which must ensure the achievement of damage doses at least 140 dpa. The solution of this problem is provided by phased use as cladding materials of austenitic steels ChS68 and EK164 (maximum damage doses ~ 92 and ~110-115 dpa, respectively), precipitation-hardening heat resistant ferritic-martensitic steels EK181 and ChS139 (maximum damage dose ~140 dpa) and oxide dispersion strengthened (ODS) steels (maximum damage dose more than 140 dpa). For development of core materials for reactors with lead and lead-bismuth eutectic coolants the most serious challenge is corrosion resistance of materials in coolant. Therefore at present time a very wide range of works on study of corrosion resistance of candidate materials is carrying out. As the basic material for the cladding tubes is considered a ferritic-martensitic steel EP823 with high silicon content. In this report the main results of works on justification of the working capacity of materials of different classes in respect to use it in cores of operating and prospective fast reactors with different types of coolant and prospects of further development of works are presented. (author)

  8. Overview of UK programme on mechanical properties of fast reactor structural materials

    International Nuclear Information System (INIS)

    Wood, D.S.

    The UK programme has been devised to endorse the use of Type 316 steel and its associated weld metal in the primary circuit of a fast reactor. In relation to ferritic steels for the steam generator, most emphasis is being placed on 9%Cr1%Mo steel (thin and thick section), with attention also being given to 2.25%Cr1%Mo steel and a number of high strength 9%Cr steels. Baseline information is being obtained on material 'as received' condition but emphasis is also being given to service conditions which may influence the behaviour such as thermal aging, irradiation and sodium environments. The situation regarding and future work intentions on these steels for UK fast reactor applications is given

  9. Measurements and calculations of integral capture cross-sections of structural materials in fast reactor spectra

    International Nuclear Information System (INIS)

    Seth, S.; Brunson, G.; Gmuer, K.; Jermann, M.; McCombie, C.; Richmond, R.; Schmocker, U.

    1979-01-01

    This paper relates the checking of integral data of steel and iron in fast reactor lattices. The fully-rodded GCFR benchmark lattice of the zero-energy reactor PROTEUS was successively modified by replacing the PuO 2 -UO 2 fuel rods by steel-18/8 or steel-37 (iron) rods. The neutron spectra of the modified lattices in fact have median energies close to that of a typical LMFBR. The replacement of fuel by the structural material of interest was such that in each case the value of k(infinity) was reduced to near-unity. This allowed the measurement of the lattice-k(infinity) by the null-reactivity technique. In addition, the principal reaction rates (namely U238 capture and fission, relative to Pu239 fission) and the neutron spectrum were measured. These directly measured integral data which are particularly sensitive to the steel cross-sections can be used for the checking and systematic adjustment of data sets. The results may also be analysed so as to derive specific values for the integral capture cross-sections of steel and iron. Neutron balance equations were set-up for each of the lattices using the measured k(infinity) and reaction rates

  10. Materials development for fast reactor applications

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-15

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

  11. Experimental determination of neutron capture cross sections of fast reactor structure materials integrated in intermediate energy spectra. Vol. 2: description of experimental structure

    International Nuclear Information System (INIS)

    Tassan, S.

    1978-01-01

    A selection of technical documents is given concerning the experimental determination of the neutron capture cross-sections of fast reactor structural materials (Fe, Cr, Ni...) integrated over the intermediate energy spectra. The experimental structure project and modifications of the reactor RB2 for this experiment, together with criticality and safety calculations, are presented

  12. The behaviour of materials in fast reactors

    International Nuclear Information System (INIS)

    Matthews, J.R.

    1977-01-01

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

  13. Material choices for the commercial fast reactor steam generators

    International Nuclear Information System (INIS)

    Willby, C.; Walters, J.

    1978-01-01

    Experience with fast reactor steam generators has shown them to be critical components in achieving a high availability. This paper presents the designers views on the use of ferritic materials for steam generators and describes the proposed design of the steam generators for the Commercial Fast Reactor (CFR), prototype of which are to be inserted in the Prototype Fast Reactor at Dounreay. (author)

  14. Corrosion of Structural Materials in Liquid Metals Used as Fast Reactor Coolants

    International Nuclear Information System (INIS)

    Balbaud-Célérier, F.; Courouau, J.L.; Martinelli, L.

    2013-01-01

    Conclusions: • Thermodynamic data give the stable state of the system, the compounds susceptible to form but no information on the kinetics of the process; • Need to perform corrosion tests in controlled conditions of temperature, chemistry, hydrodynamics; • Comparison of the materials behaviour: first selection of materials, optimisation of the composition; • Fundamental work on the understanding of the corrosion process to develop corrosion models and predictive laws to guarantee the long term behaviour

  15. Design and selection of materials for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Chetal, S.C.

    2011-01-01

    Sodium cooled fast reactors are currently in operation, under construction or under design by a number of countries. The design of sodium cooled fast reactor is covered by French RCC - MR code and ASME code NH. The codes cover rules as regards to materials, design and construction. These codes do not cover the effect of irradiation and environment. Elevated temperature design criteria in nuclear codes are much stringent in comparison to non nuclear codes. Sodium corrosion is not an issue in selection of materials provided oxygen impurity in sodium is controlled for which excellent reactor operating experience is available. Austenitic stainless steels have remained the choice for the permanent structures of primary sodium system. Stabilized austenitic stainless steel are rejected because of poor operating experience and non inclusion in the design codes. Route for improved creep behaviour lies in compositional modifications in 316 class steel. However, the weldability needs to be ensured. For cold leg component is non creep regime, SS 304 class steel is favoured from overall economics. Enhanced fuel burn up can be realized by the use of 9-12%Cr 1%Mo class steel for the wrapper of MOX fuel design, and cladding and wrapper for metal fuel reactors. Minor compositional modifications of 20% cold worked 15Cr-15Ni class austenitic stainless steel will be a strong candidate for the cladding of MOX fuel design in the short term. Long term objective for the cladding will be to develop oxide dispersion strengthened steel. 9%Cr 1%Mo class steel (Gr 91) is an ideal choice for integrated once through sodium heated steam generators. One needs to incorporate operating experience from reactors and thermal power stations, industrial capability and R and D feedback in preparing the technical specifications for procurement of wrought products and welding consumables to ensure reliable operation of the components and systems over the design life. The paper highlights the design approach

  16. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

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

  17. Use the results of measurements on KBR facility for testing of neutron data of main structural materials for fast reactors

    Science.gov (United States)

    Koscheev, Vladimir; Manturov, Gennady; Pronyaev, Vladimir; Rozhikhin, Evgeny; Semenov, Mikhail; Tsibulya, Anatoly

    2017-09-01

    Several k∞ experiments were performed on the KBR critical facility at the Institute of Physics and Power Engineering (IPPE), Obninsk, Russia during the 1970s and 80s for study of neutron absorption properties of Cr, Mn, Fe, Ni, Zr, and Mo. Calculations of these benchmarks with almost any modern evaluated nuclear data libraries demonstrate bad agreement with the experiment. Neutron capture cross sections of the odd isotopes of Cr, Mn, Fe, and Ni in the ROSFOND-2010 library have been reevaluated and another evaluation of the Zr nuclear data has been adopted. Use of the modified nuclear data for Cr, Mn, Fe, Ni, and Zr leads to significant improvement of the C/E ratio for the KBR assemblies. Also a significant improvement in agreement between calculated and evaluated values for benchmarks with Fe reflectors was observed. C/E results obtained with the modified ROSFOND library for complex benchmark models that are highly sensitive to the cross sections of structural materials are no worse than results obtained with other major evaluated data libraries. Possible improvement in results by decreasing the capture cross section for Zr and Mo at the energies above 1 keV is indicated.

  18. Structural dynamics in fast reactor accident analysis

    International Nuclear Information System (INIS)

    Fistedis, S.H.

    1975-01-01

    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

  19. Materials challenges supporting new sodium fast reactor designs

    International Nuclear Information System (INIS)

    Gelineau, O.; Goff, S. Dubiez-le; Dubuisson, Ph.; Dalle, F.; Blat, M.

    2009-01-01

    Sodium Fast Reactor is considered in France as the most mature technology of the different Generation IV systems. In the short-term the designing work is focused on the identification of the potential tracks to improve competitiveness, safety, efficiency and to reduce cost. In that frame the materials have a key role to play. This paper is focused on the new materials envisaged and on the Research and Development program launched in France by Areva NP, CEA and EDF in order to sustain the innovative design options: ferritic steels as candidates for exchangers, steam generators and possibly sodium circuits, optimization of materials and fabrication processes to improve safety and risk management, extension of material databases to take into account the 60 years life duration including irradiation and ageing effect. (author)

  20. Recent developments for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, N.; Nakamuria, K.; Morishita, M.; Shibamoto, H.; Nagashima, H.; Inoue, K.

    2005-01-01

    For realization of reliable and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute(JNC) and Japan Atomic Power Company(JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts through evaluation of their structural integrity, the research and development of 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)' is recognized as an essential theme. FDS focuses on particular failure modes of FRs such as ratchet deformation and creep fatigue damages due to cyclic thermal loads. To evaluate these modes, three main developments are in progress. One is 'Refinement of Failure Criteria' for particular modes of FRs. Next is development of 'Guidelines for Inelastic Design Analysis' in order to predict elastic plastic and creep behaviors. Furthermore, efforts are being made toward preparing 'Guidelines for Thermal Load Modeling' for FR component design where thermal loads are dominant. These studies were performed under the sponsorship of the Ministry of Economy, Trade and Industry of Japanese government. (authors)

  1. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour

  2. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour.

  3. Statistical properties of material strength for reliability evaluation of components of fast reactors. Austenitic stainless steels

    International Nuclear Information System (INIS)

    Takaya, Shigeru; Sasaki, Naoto; Tomobe, Masato

    2015-03-01

    Many efforts have been made to implement the System Based Code concept of which objective is to optimize margins dispersed in several codes and standards. Failure probability is expected to be a promising quantitative index for optimization of margins, and statistical information for random variables is needed to evaluate failure probability. Material strength like tensile strength is an important random variable, but the statistical information has not been provided enough yet. In this report, statistical properties of material strength such as creep rupture time, steady creep strain rate, yield stress, tensile stress, flow stress, fatigue life and cyclic stress-strain curve, were estimated for SUS304 and 316FR steel, which are typical structural materials for fast reactors. Other austenitic stainless steels like SUS316 were also used for statistical estimation of some material properties such as fatigue life. These materials are registered in the JSME code of design and construction of fast reactors, so test data used for developing the code were used as much as possible in this report. (author)

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

    International Nuclear Information System (INIS)

    2013-01-01

    In recent years, engineering oriented work, rather than basic research and development (R&D), has led to significant progress in improving the economics of innovative fast reactors and associated fuel cycle facilities, while maintaining and even enhancing the safety features of these systems. Optimization of plant size and layout, more compact designs, reduction of the amount of plant materials and the building volumes, higher operating temperatures to attain higher generating efficiencies, improvement of load factor, extended core lifetimes, high fuel burnup, etc. are good examples of achievements to date that have improved the economics of fast neutron systems. The IAEA, through its Technical Working Group on Fast Reactors (TWG-FR) and Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), devotes many of its initiatives to encouraging technical cooperation and promoting common research and technology development projects among Member States with fast reactor and advanced fuel cycle development programmes, with the general aim of catalysing and accelerating technology advances in these fields. In particular the theme of fast reactor deployment, scenarios and economics has been largely debated during the recent IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, held in Paris in March 2013. Several papers presented at this conference discussed the economics of fast reactors from different national and regional perspectives, including business cases, investment scenarios, funding mechanisms and design options that offer significant capital and energy production cost reductions. This Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics addresses Member States’ expressed need for information exchange in the field, with the aim of identifying the main open issues and launching possible initiatives to help and

  5. Proliferation Resistance and Material Type considerations within the Collaborative Project for a European Sodium Fast Reactor

    International Nuclear Information System (INIS)

    Renda, Guido; Alim, Fatih; Cojazzi, Giacomo GM.

    2015-01-01

    The collaborative project for a European Sodium Fast Reactor (CP‑ESFR) is an international project where 25 European partners developed Research & Development solutions and concepts for a European sodium fast reactor. The project was funded by the 7. European Union Framework Programme and covered topics such as the reactor architectures and components, the fuel, the fuel element and the fuel cycle, and the safety concepts. Within sub‑project 3, dedicated to safety, a task addressed proliferation resistance considerations. The Generation IV International Forum (GIF) Proliferation Resistance and Physical Protection (PR and PP) Evaluation Methodology has been selected as the general framework for this work, complemented by punctual aspects of the IAEA‑INPRO Proliferation Resistance methodology and other literature studies - in particular for material type characterization. The activity has been carried out taking the GIF PR and PP Evaluation Methodology and its Addendum as the general guideline for identifying potential nuclear material diversion targets. The targets proliferation attractiveness has been analyzed in terms of the suitability of the targets’ nuclear material as the basis for its use in nuclear explosives. To this aim the PR and PP Fissile Material Type measure was supplemented by other literature studies, whose related metrics have been applied to the nuclear material items present in the considered core alternatives. This paper will firstly summarize the main ESFR design aspects relevant for PR following the structure of the GIF PR and PP White Paper template. An analysis on proliferation targets is then discussed, with emphasis on their characterization from a nuclear material point of view. Finally, a high‑level ESFR PR analysis according to the four main proliferation strategies identified by the GIF PR and PP Evaluation Methodology (concealed diversion, concealed misuse, breakout, clandestine production in clandestine facilities) is

  6. Materials Options of Steam Generator for Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Fu Xiaogang; Long Bin; Han Liqing; Qin Bo; Zhang Jinquan; Wang Shuxing

    2013-01-01

    Overview of the material options of steam generator for sodium-cooled fast reactors, the method to calculate the service life, the thinning of wall thickness and the sodium corrosion rate, the degradation of mechanical properties (thermal aging and decarburization) and the calculation results of theoretical models

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  8. Materials Performance in Sodium-Cooled Fast Reactors: Past, Present, and Future

    International Nuclear Information System (INIS)

    Natesan, K.; Li Meimei

    2013-01-01

    • This paper gives an overview of the requirements, selection, and performance of materials for in-core and out-of-core components in SFRs. • Globally, sodium-cooled fast reactors have been designed, built, and operated in several countries. A substantial database exists for the existing materials on their functional and mechanical performance. • The 60-yr design life of the SFR presents a significant challenge to the development of database, extrapolation/prediction of long-term performance, and high-temperature design methodology for the structural components. • Licensing of SFR requires a valid assessment of the environmental effects (irradiation, thermal aging, and sodium) on materials performance. • Advanced materials such as, ODS alloys for cladding, Gr91 and 92 F/M steels, and austenitic alloys such as NF709 for structures can improve the economy, safety, and flexibility of SFRs. A substantial database is needed for all these materials and global effort is underway to develop the needed information through experimentation and modeling

  9. Overview of 9Cr steels properties for structural application in sodium fast reactors

    International Nuclear Information System (INIS)

    Cabet, Celine; Courouau, Jean-Louis; Dalle, France; Desgranges, Clara; Forest, Laurent; Martinelli, Laure; Sauzay, Maxime

    2015-01-01

    A research and development programme has been launched by CEA, EDF and AREVA for the choice and qualification of material for sodium fast reactor (SFR) structural components. The requirements on steam generator (SG) are demanding, with operating temperatures ranging from 240 deg. C to 530 deg. C in water/steam and in sodium for an extended design life of several decades. The selection of the SG materials is based on many characteristics: fabrication, welding, thermal properties, mechanical strength at low and high temperature, environmental resistance. 9%Cr steels which are relevant candidate alloys for different designs of SGs have been extensively studied in the past decade. The objective of this paper is to review some advances made at CEA on determining properties of the X10CrMoVNb9-1 steel (hereafter named 'grade 91'): welding, modelling of cyclic softening, modelling of long-term creep, compatibility with liquid sodium, corrosion in steam. (authors)

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

    International Nuclear Information System (INIS)

    2013-01-01

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

  11. Integrity of Safety-Related Fast Reactor Structures

    International Nuclear Information System (INIS)

    Rose, R.T.; Tomkins, B.

    1981-01-01

    The LMFBR contains several structural items whose integrity must be safeguarded during the life of the plant. These items include the main core support structures (strongback, diagrid) and the primary tank to which these structures are attached. In order to demonstrate an acceptable level of structural integrity, the chosen design philosophy must be supported by both analytical and experimental evidence. This paper describes the current approaches in the UK to these requirements. Section 2 describes the materials mechanical properties tests performed to date on both fracture toughness and fatigue crack growth in Type 316 austenitic stainless steel plate and weldments. This data illustrates the problems in identifying the relevant materials fracture parameters for use in assessments. Section 3 shows the test programmes in hand to extend the materials programmes to tests on structural features (mainly welded wide plate tests) which incorporate the complexity of weldments in a structural context. This includes experimental evidence on the effects of local weld residual stresses on structural failure. Various routes are open for the integrity assessment of FR structures. These are discussed in Section 4 but in effect they reduce to a fracture mechanics approach using some technique to cope with elastic-plastic fracture. The main problems at present relate to our ability in analysis to cope with residual stresses and the post-initiation region of the fracture resistance curve. Also, there is the problem of initial defect sizing by current NDE techniques. Current conservative analytical assessments give acceptable defect sizes of order a few millimetres in irradiated weldments. Finally, Section 5 discusses the options open in design to cope with safety related structures under normal and abnormal loading conditions. It is clear that several options exist in design to satisfy the demand for high integrity

  12. Consequences of the improvement of fast reactor material behavior under irradiation on fuel element performance

    International Nuclear Information System (INIS)

    Leclere, J.; Dupouy, J.M.; Marcon, J.P.

    1979-01-01

    The most important problems in fast reactor fuel element come from the excessive swelling of the structural materials used. The limitations of irradiation time for a given reactor result from the cladding or hexagonal wrapper deformations. Irradiation creep plays a major role, either in inducing additional deformations, or in providing possible ways of accommodation of bending stresses. Progress has been made in designing swelling resistant and/or low irradiation creep modulus materials. For instance in FRANCE, annealed 316 SS has been eliminated from pin and subassembly, and replaced by cold worked 316; we are now considering introduction of stabilizing elements in 316 SS as a further improvement and studying different alloys (nickel alloys, or ferritic steels). It has to be checked that the improvement of irradiation characteristic is not counterbalanced by losses on other properties (embrittlement for instance). Considering that pushing off or eliminating a limit may lead to the onset of a new one, it is porposed to make a review of the consequences of substantial improvement of structural material behavior

  13. Status of core material development for fast reactor in Japan

    International Nuclear Information System (INIS)

    Ukai, S.; Shibahara, I.; Nagai, S.

    1994-01-01

    In the last two decades, extensive efforts have been devoted to the development of mixed-oxide fuel for LMFBR in Japan. For the fuel of the prototype reactor MONJU, drastic improvement in creep rupture strength and swelling resistance were attained by modification within the compositional specification of the standard Type 316 stainless steel (PNC316). For the fuel of future large-scale reactors, extensive research and development program are under way to realize the long life fuel. The candidate material for demonstration reactor is advanced austenitic stainless steel (PNC1520) which intended to modify the composition beyond the Type 316 stainless steel specification. In order to further improve the swelling resistance, the austenitic stainless steel with higher nickel content (High Ni alloy) and ferritic/martensitic steel (PNC-FMS) are developed. In a prospective cladding material for the long life fuel, the development of oxide dispersion strengthened (ODS) ferritic steel is focused to establish the alloying design and fabrication process toward as high as 250dpa. (author)

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

    International Nuclear Information System (INIS)

    2002-01-01

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

  15. Fast reactor database

    International Nuclear Information System (INIS)

    1996-02-01

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

  16. Design codes for fast reactor steam generators

    International Nuclear Information System (INIS)

    Townley, C.H.A.

    1978-01-01

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

  17. Factors in the selection of broiler tube materials for a civil fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tyzack, C; Chitty, A

    1975-07-01

    This paper briefly considers some of the factors which must be balanced in the selection of a boiler tube material for a Civil Fast Reactor. The merits and possible demerits of low alloy ferritic steels and the austenitic Alloy 800 are compared with respect to waterside corrosion resistance, mechanical properties, fabrication and weldability and possible effects of exposure to the sodium environment under normal and fault conditions. It is pointed out that although there is operational experience of most of the materials in boiler superheater applications there is little or none in evaporative regimes. (author)

  18. Core performance of equilibrium fast reactors for different coolant materials and fuel types

    International Nuclear Information System (INIS)

    Mizutani, Akihiko; Sekimoto, Hiroshi

    1998-01-01

    Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

  19. Package of programs for calculating accidents involving melting of the materials in a fast-reactor vessel

    International Nuclear Information System (INIS)

    Vlasichev, G.N.

    1994-01-01

    Methods for calculating one-dimensional nonstationary temperature distribution in a system of physically coupled materials are described. Six computer programs developed for calculating accident processes for fast reactor core melt are described in the article. The methods and computer programs take into account melting, solidification, and, in some cases, vaporization of materials. The programs perform calculations for heterogeneous systems consisting of materials with arbitrary but constant composition and heat transfer conditions at material boundaries. Additional modules provide calculations of specific conditions of heat transfer between materials, the change in these conditions and configuration of the materials as a result of coolant boiling, melting and movement of the fuel and structural materials, temperature dependences of thermophysical properties of the materials, and heat release in the fuel. 11 refs., 3 figs

  20. Fast reactor database. 2006 update

    International Nuclear Information System (INIS)

    2006-12-01

    plants. The recurring themes are the selection and summary of the data associated with the choice of coolant, fuel and structural materials, reduction of the steel weight, simplification of the plant design/layout, other important fast reactor design issues, and how to solve these problems. In the field of fast reactor design and operational data, the last reference document published by the IAEA was the 1996 Fast Reactor Database (IAEA-TECDOC-866). Since its publication, quite a lot has happened: the construction of two new reactors has been launched, and conceptual/design studies were initiated for various fast reactors, e.g. the Japanese JSFR-1500 and the Russian BN-1800 (both cooled by sodium), as well as for a wholly new line of LMFR concepts - modular reactors cooled by sodium and by lead-bismuth alloy, and prototype and demonstration commercial size fast reactors cooled by lead. The data were produced by the IAEA's Technical Working Group on Fast Reactors (TWG-FR). For many of the TWG-FR Member States there is a significant history of fast reactor development, often extending over a period of 40+ years. The new and updated information on LMFR, which are in operation, under construction or development, has been prepared with contributions from China, India, Japan, Republic of Korea and the Russian Federation. The information contained in IAEA-TECDOC-866, produced by France, Germany, Italy, the UK and the USA, was included in the present report with some modification taking into account last events

  1. Proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the integral fast reactor

    International Nuclear Information System (INIS)

    Burris, L.; Steindler, M.; Miller, W.

    1984-01-01

    The pool-type Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps: a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

  2. A proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.; Steindler, M.; Miller, W.

    1984-01-01

    The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps -- a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

  3. Research and development issues for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Ando, Masanori; Morishita, Masaki

    2003-01-01

    For realization of safe and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts and validate their structural integrity, the research and development of 'Fast Reactor Structural Design Standard (FDS)' is recognized as an essential theme. FDS considers general characteristics of FRs and design needs for their rationalization. Three main subjects were settled in research and development issues of FDS. One is rationalization of failure criteria' taking characteristic design conditions into account. Next is development of 'a guideline on inelastic analysis for design' in order to predict elastic plastic and creep behaviours of high temperature components. Furthermore, efforts are being made toward preparing a guideline on thermal loads modeling' for FR component design where thermal loads are dominant. (author)

  4. UKAEA fast reactor project research and development programme on fuel element cladding and sub-assembly wrapper materials

    International Nuclear Information System (INIS)

    Harries, D.R.

    1977-01-01

    Research and development work on fuel element component (cladding, subassembly wrappers, etc.) materials for the U.K. sodium cooled fast reactor programme has been conducted at the United Kingdom Atomic Energy Authority (UKAEA) establishments at Dounreay, Harwell, Risley, and Springfields during the past fifteen years or so. This work has formed an integral part of, and has been co-ordinated by, the UKAEA Fast Reactor Project and has involved close liaison with the Nuclear Power Company (NPC) and the Central Electricity Generating Board (CEGB). The research and development were initially related to the Prototype Fast Reactor (PFR) but the scope has now been extended to cover the first Civil Fast Reactor (CFR1), which has recently been re-designated the Civil Demonstration Fast Reactor (CDFR). The paper outlines the present status of the development of sodium cooled fast reactors in the U.K. and proceeds to summarize the principal PFR and CDFR core and fuel element parameters which have determined the planning and direction of the fuel element materials programme. The current position on the fuel element cladding and wrapper research and development programme is reviewed, and the facilities and future irradiation programme to be carried out in PFR are described

  5. Recent research on structural integrity in support of the european fast reactor

    International Nuclear Information System (INIS)

    Acker, D.; Townley, C.H.A.

    1991-01-01

    For the European Fast Reactor, which is being developed jointly in France, Germany and the U.K., qualified material controlled fabrication processes and validated design rules are essential to achieve cost reduction in components, and at the same time, maintain the required standards of safety and reliability. In the framework of an inter-governmental agreement, the R and D organisations: CEA in France, KfK and INTERATOM in Germany and AEA Technology in the UK, have co-ordinated their work to support the EFR project. In the Structural Integrity area, this co-ordination is undertaken by the AGT 9B group, which: The Purpose of the present paper is to review the progress which has been made up to the end of 1990, as recorded by the 35 AGT9B Milestone Reports which had been issued by this date. Only a brief outline of the main results will be given. Detailed descriptions of the work will be found in the papers presented to SMIRT Conferences in 1987, 1989 and 1991 and referenced in chapter 7. The results fall into five areas - General design guidance - Analysis of structures submitted to repeated loading - Fracture mechanics - Buckling - Design rules for specific components

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

    International Nuclear Information System (INIS)

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-01

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

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

    International Nuclear Information System (INIS)

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

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

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

    International Nuclear Information System (INIS)

    1999-01-01

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

  12. Effect of Reflector Material on the Neutronic Characteristics of the Small Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sung Hwan; Baek, Min Ho; Yoo, Jae Woon; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The sodium-cooled fast reactor (SFR) has been chosen as a candidate for the Gen-IV Nuclear Energy Systems Initiative due to the advantages in utilization of uranium resources and reduction of radioactive wastes. Recently, the uranium blanket concept is omitted for a purpose of the non-proliferation, hence the reflector material plays a more important role in reactor core design. Moreover, especially in the Korean prototype SFR, the initial core should startup with low-enriched uranium ({<=} 20 w/o) for 100 {approx} 150 MWe power. This restriction causes significant difficulties to achieve sufficient excess reactivity. Thus, in this paper, core characteristic studies of various reflector materials (HT9, BeO, MgO, and ZrH{sub 1.6}) are performed to enhance the initial core excess reactivity

  13. Consultancy on 'IAEA initiative to establish a fast reactor knowledge base'. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    At the outset of the meeting, Member States interest in establishing Fast Reactor Knowledge Base was acknowledged by the participants. While the broader objective of the initiative was to develop a Knowledge Base into which the existing Knowledge Preservation Systems will fit, the specific objectives of the meeting were: Make recommendations on FRKP methodology and guidance, Review the proposed structure of the Agency's FRKP Initiative, Make recommendations on the role of the Agency and the Member States implementing the Agency's FRKP Initiative, Develop an approach for the implementation of the structure of the Agency's RFKP Initiative. The meeting concluded covering many aspects of the initiative namely systematic method of data capturing, structuring and functions of FRKP System etc. and placed a strong emphasis on the continues role of IAEA's support and coordination in the data retrieval and knowledge preservation efforts

  14. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  15. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    At present nuclear power accounts for approximately 17% of total electricity generation worldwide. Given continuing population growth and the needs of the third world and developing countries to improve their economic performance and standard of living, energy demand is expected to continue to grow through the 21st century. The proportion of energy supplied as electricity is also expected to continue to increase. Although fossil-fuelled electricity generation is the option preferred by several countries for the short term, there are rising concerns over climatic consequences caused by extended burning of fossil fuels as a result of the demands of a fast expanding world population. In this situation nuclear electricity will become more and more important and the known reserves of uranium would be consumed quite quickly by thermal reactors. It would be possible to sustain a large nuclear programme only by introducing fast reactors. One can conclude that there are strategic reasons for pursuing the development of fast breeder reactors. It will become desirably essential to have this technology available for introduction. The recycling of plutonium into LMFRs would allow 'burning' of the associated extremely long-life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. This additional important mission for the LMFR is gaining worldwide interest. In the framework of disarmament of nuclear weapons and the utilization of the nuclear material or peaceful purposes a role for fast reactors can be also considered. Over the past 29 years, the IAEA has actively encouraged and advocated international co-operation in Fast Breeder Reactor Technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1995, as reported at the 29th Annual

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

    International Nuclear Information System (INIS)

    1996-01-01

    At present nuclear power accounts for approximately 17% of total electricity generation worldwide. Given continuing population growth and the needs of the third world and developing countries to improve their economic performance and standard of living, energy demand is expected to continue to grow through the 21st century. The proportion of energy supplied as electricity is also expected to continue to increase. Although fossil-fuelled electricity generation is the option preferred by several countries for the short term, there are rising concerns over climatic consequences caused by extended burning of fossil fuels as a result of the demands of a fast expanding world population. In this situation nuclear electricity will become more and more important and the known reserves of uranium would be consumed quite quickly by thermal reactors. It would be possible to sustain a large nuclear programme only by introducing fast reactors. One can conclude that there are strategic reasons for pursuing the development of fast breeder reactors. It will become desirably essential to have this technology available for introduction. The recycling of plutonium into LMFRs would allow 'burning' of the associated extremely long-life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. This additional important mission for the LMFR is gaining worldwide interest. In the framework of disarmament of nuclear weapons and the utilization of the nuclear material or peaceful purposes a role for fast reactors can be also considered. Over the past 29 years, the IAEA has actively encouraged and advocated international co-operation in Fast Breeder Reactor Technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1995, as reported at the 29th Annual

  18. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-01-01

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

  19. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

    2013-10-15

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

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

    International Nuclear Information System (INIS)

    2003-01-01

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

  1. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. 2. Replacement of upper core structure

    International Nuclear Information System (INIS)

    Ushiki, Hiroshi; Ito, Hiromichi; Okuda, Eiji; Suzuki, Nobuhiro; Sasaki, Jun; Oota, Katsu; Kawahara, Hirotaka; Takamatsu, Misao; Nagai, Akinori; Okawa, Toshikatsu

    2015-01-01

    In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of MARICO-2 (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS) in 2007. As a part of the restoration work, UCS replacement was begun at March 24, 2014 and was completed at December 17. In-vessel repair (including observation) for sodium-cooled fast reactors (SFRs) is distinct from that for light water reactors and necessitates independent development. Application of developed in-vessel repair techniques to operation and maintenance of SFRs enhanced their safety and integrity. There is little UCS replacement experience in the world and this experience and insights, which were accumulated in the replacement work of in-vessel large structure (UCS) used for more than 30 years, are expected to improve the in-vessel repair techniques in SFRs. (author)

  2. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    Radiation effects on metals or alloys used in fast reactor core components are examined in the papers presented at this conference, the accent being put on swelling and irradiation creep of steels and nickel alloys

  3. SIMIFR: A code to simulate material movement in the Integral Fast Reactor

    International Nuclear Information System (INIS)

    White, A.M.; Orechwa, Yuri.

    1991-01-01

    The SIMIFR code has been written to simulate the movement of material through a process. This code can be used to investigate inventory differences in material balances, assist in process design, and to produce operational scheduling. The particular process that is of concern to the authors is that centered around Argonne National Laboratory's Integral Fast Reactor. This is a process which involves the irradiation of fissile material for power production, and the recycling of the irradiated reactor fuel pins into fresh fuel elements. To adequately simulate this process it is necessary to allow for locations which can contain either discrete items or homogeneous mixtures. It is also necessary to allow for a very flexible process control algorithm. Further, the code must have the capability of transmuting isotopic compositions and computing internally the fraction of material from a process ending up in a given location. The SIMIFR code has been developed to perform all of these tasks. In addition to simulating the process, the code is capable of generating random measurement values and sampling errors for all locations, and of producing a restart deck so that terminated problems may be continued. In this paper the authors first familiarize the reader with the IFR fuel cycle. The different capabilities of the SIMIFR code are described. Finally, the simulation of the IFR fuel cycle using the SIMIFR code is discussed. 4 figs

  4. IAEA Technical Meeting on Innovative Fast Reactor Designs with Enhanced Negative Reactivity Feedback Features. Working Material

    International Nuclear Information System (INIS)

    2012-01-01

    The objective of the TM was to review and discuss the safety characteristics and the performances of the core of innovative fast reactor concepts, as well as to present the ongoing R&D activities in the area of core design and advanced simulation tools and methods for fast reactor core physics analysis. The focus was on fast spectrum cores optimized for actinide utilization and transmutation and, in particular, on core designs with enhanced negative reactivity feedback effects

  5. IAEA Workshop (Training Course) on Codes and Standards for Sodium Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2010-01-01

    The training course consisted of lectures and Q&A sessions. The lectures dealt with the history of the development of Design Codes and Standards for Sodium Cooled Fast Reactors (SFRs) in the respective country, the detailed description of the current design Codes and Standards for SFRs and their application to ongoing Fast Reactor design projects, as well as the ongoing development work and plans for the future in this area. Annex 1 contains the detailed Workshop program

  6. Lead-Cooled Fast Reactor Systems and the Fuels and Materials Challenges

    Directory of Open Access Journals (Sweden)

    T. R. Allen

    2007-01-01

    Full Text Available Anticipated developments in the consumer energy market have led developers of nuclear energy concepts to consider how innovations in energy technology can be adapted to meet consumer needs. Properties of molten lead or lead-bismuth alloy coolants in lead-cooled fast reactor (LFR systems offer potential advantages for reactors with passive safety characteristics, modular deployment, and fuel cycle flexibility. In addition to realizing those engineering objectives, the feasibility of such systems will rest on development or selection of fuels and materials suitable for use with corrosive lead or lead-bismuth. Three proposed LFR systems, with varying levels of concept maturity, are described to illustrate their associated fuels and materials challenges. Nitride fuels are generally favored for LFR use over metal or oxide fuels due to their compatibility with molten lead and lead-bismuth, in addition to their high atomic density and thermal conductivity. Ferritic/martensitic stainless steels, perhaps with silicon and/or oxide-dispersion additions for enhanced coolant compatibility and improved high-temperature strength, might prove sufficient for low-to-moderate-temperature LFRs, but it appears that ceramics or refractory metal alloys will be necessary for higher-temperature LFR systems intended for production of hydrogen energy carriers.

  7. Studies of the reactivity effects of hydrogenous material in a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Ingram, D.; Sweet, D.W.

    1979-01-01

    The reactivity effects of hydrogenous substances, such as the oil used in the primary coolant pumps, which could enter the core of a fast reactor in a hypothetical accident, have been studied in a series of experiments and calculations. Measurements to study the influence of the density of the hydrogen and its location on reactivity were made in two assemblies in the zero power reactor ZEBRA. The first of these was similar in size to PFR and the second was the large BZB assembly of the BIZET programme. The results of this work have been compared with calculations using the FGL5 nuclear data library. Calculations for a 1200 MW (e) CFR have been made using three quantities of material (8, 40 and 160 Kg of hydrocarbon, equivalent to 10, 50 and 200 litres of oil). The calculations have used different geometrical models and hydrocarbon distributions and have explored the influence of core temperature, fuel burn-up and the presence of control rods to estimate the maximum reactivity changes that can be obtained. The results have been analysed in terms of components of the change in neutron balance produced by the material and uncertainties in these have been derived from the ZEBRA work. (author)

  8. Results of investigations in a fast reactor configuration with a strong heterogeneous cell structure

    International Nuclear Information System (INIS)

    Lehmann, E.; Albert, D.; Dietze, K.; Faehrmann, K.; Hansen, W.; Huettel, G.; Wand, H.; Osmera, B.

    1984-11-01

    Investigations of the neutron flux spectrum, of the energy dependent importance function and of sample reactivities were performed in a fast reactor configuration characterized by a marked neutronic microstructure which is produced by the insertion of pellets of polyethylene and cadmium. Contrary to results of calculations with a homogenized composition, values obtained by means of the tree programs CARMEN, YARAB and P1X, developed in Rossendorf, agree well with measurements. Furthermore it could be shown that an adjoint weighting for the determination of cell-averaged values of the importance function is justified. In a configuration with larger heterogeneity induced by the aggregation of four uranium pellets significant spectral differences between uranium and material zones, respectively, show up in activation measurements as well as in sample reactivity determinations. (author)

  9. Gas cooled fast reactor materials: compatibility and reaction kinetics of fuel/matrices couples

    International Nuclear Information System (INIS)

    Lechelle, J.; Aufore, L.; Basini, V.; Belin, R.; Vaudez, S.

    2004-01-01

    Fourth Generation Gas cooled Fast Reactor concept implies a fast neutron spectrum and aims to lead to an iso-generation of minor actinides. Criteria have been defined for these fuels such as: high core filling factor, efficient fuel cooling, low operation temperature, i.e. 400-850 deg C, good fission product retention, burn-ups in the range of 5-8 atom%, Pu content in the range of 15-25%. Materials matching this demand are considered: mixed uranium - plutonium nitrides and carbides as fuels, whereas TiN, TiC, ZrN, ZrC, SiC are investigated as inert matrices. Thermo-chemical compatibility studies have been carried out, mostly for (U,Pu)N/SiC and (U,Pu)N/TiN couples. They have been associated to matching diffusional studies. For the first studies, accidental reactor conditions have been chosen (1600 deg C) so as to select a couple. Results are presented in terms of nature and quantity of resulting phases identified by XRD and SEM for thermodynamical equilibrium experiments. (authors)

  10. IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2011-01-01

    The IAEA, within the framework of its Nuclear Energy Department’s Technical Working Group on Fast Reactors (TWG-FR), assists Member States activities in fast reactors technology development areas by providing an umbrella for information exchange [topical Technical Meetings (TMs), Workshops and large Conferences] and collaborative R&D [Coordinated Research Projects (CRPs)]. The Technical meeting on “Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors” was held from 21 – 22 December 2011 in Vienna, addressing Member States’ expressed needs of information exchange in the field of advanced fast reactor design features, with particular attention to innovative heat exchangers and steam generators. The Objective of the TM is to provide a global forum for in-depth information exchange and discussion on the most advanced concepts of heat exchangers and steam generators for fast reactors. More specifically, the objectives are: · Review of the status of advanced fast reactor development activities with special emphasis on design and performance of heat exchangers and steam generators; · Discuss requirements for innovative heat exchangers and steam generators; · Present results of studies and conceptual designs for innovative heat exchangers and steam generators; · Provide recommendations for international collaboration under the IAEA aegis. The meeting agenda of the meeting is in Annex I

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    2015-01-01

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

  13. Status of liquid metal cooled fast reactor technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-04-01

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

  14. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    1999-04-01

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

  15. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Wheeler, R.C.; Bramman, J.I.

    1988-04-01

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

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

    International Nuclear Information System (INIS)

    2005-01-01

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

  17. Advances on the analysis of fast reactor core and coolant circuit structures

    International Nuclear Information System (INIS)

    Livolant, M.; Imazu, A.; Chang, Y.W.; Eggen, D.T.

    1989-01-01

    For the 10th SMiRT Conference, it has been decided to make general reviews of the accomplishments throughout the conferences. The aim of this paper is to make such a review in the field of fast reactor core and coolant circuit structures, which is now fully treated in division E. That was not true in the past: at the earliest conferences up to the 5th, the division E dealt with accidental studies among which the hypothetical core disruptive accident was the most important. So, to cover the subject from the first SMiRT to now, it has been necessary to search into all the past division in order to recover the studies fitting into the scope of the present division E. This has allowed a table showing the number of presented papers on the various topics at the SMiRT conferences to be set up (table I). Then, some significant topics have been studied in detail, highlighting the main accomplishments, but trying also to point out the shortcomings and the work still to be done, in view of the present state of art

  18. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, Liancheng; Zhang, Bin

    2014-01-01

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  19. Electrochemistry in fast reactor technology

    International Nuclear Information System (INIS)

    Mathews, C.K.

    1987-01-01

    Electrochemistry plays a significant role in the production, characterisation or behaviour of the fuel, the coolant and structural materials used in fast reactor systems. The role of electrochemistry in sodium production, in the fuel cycle, in the development of electrochemical meters used for the on-line monitoring of the various impurities at sub ppm levels and in the recovery of plutonium and uranium are discussed. The advantage of voltammmetric techniques in the analysis of impurities and the application of electrochemical meters have been investigated. (author). 5 figs., 15 refs

  20. Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

    International Nuclear Information System (INIS)

    Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

    2009-01-01

    The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

  1. [Fortieth Annual] Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Working Material

    International Nuclear Information System (INIS)

    2008-01-01

    The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 39th TWG-FR Annual Meeting, including the status of the actions; - Consider meeting arrangements for 2007, 2008 and 2009; - Review the Agency’s ongoing information exchange and co-ordinated research activities in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations; - Discuss future joint activities in view of the Agency’s Programme and Budget Cycle 2008–2009 (and beyond)

  2. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1982-01-01

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

  3. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

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

  4. The fast reactor

    International Nuclear Information System (INIS)

    1980-02-01

    The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  6. Technical Meeting on Impact of Fukushima Event on Current and Future Fast Reactor Designs. Working Material

    International Nuclear Information System (INIS)

    2012-01-01

    The overall purpose of the Technical Meeting was to recognize and analyse the implications of the accident occurred at the Fukushima Dai-ichi Nuclear Power Station on current and future fast neutron systems design and operation. The aim was to provide a global forum for discussing the principal lessons learned from this event, and thus to review safety principles and characteristics of existing and future fast neutron concepts, especially in relation with extreme natural events which potentially may lead to severe accident scenarios. The participants also presented and discussed innovative technical solutions, design features and countermeasures for design extension conditions - including earthquakes, tsunami and other extreme natural hazards - which can enhance the safety level of existing and future fast neutron systems. Furthermore, the meeting gave the opportunity to present advanced methods for the evaluation of the robustness of plants against design extension conditions. Another important goal of this TM was to discuss how to harmonize safety approaches and goals for next generation’s fast reactors. Finally, the meeting was intended to identify areas where further research and development in nuclear safety, technology and engineering in the light of the Fukushima accident are needed. In the frame of the implementation of its Nuclear Safety Action Plan endorsed by all Member States, the IAEA will consider these areas as potential technical topics for new Coordinated Research Projects, to be launched in the near future

  7. Mass tracking and material accounting in the integral fast reactor (IFR)

    International Nuclear Information System (INIS)

    Orechwa, Y.; Adams, C.H.; White, A.M.

    1991-01-01

    This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations

  8. Mass tracking and material accounting in the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Orechwa, Y.; Adams, C.H.; White, A.M.

    1991-01-01

    The Integral Fast Reactor (IFR) is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory (ANL). There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure the compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities at ANL-West, utilizing Experimental Breeder Reactor 2 and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-Tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations. The components of the MTG System include: (1) an Oracle database manager with a Fortran interface, (2) a set of MTG ''Tasks'' which collect, manipulate and report data, (3) a set of MTG ''Terminal Sessions'' which provide some interactive control of the Tasks, and (4) a set of servers which manage the Tasks and which provide the communications link between the MTG System and Operator Control Stations, which control process equipment and monitoring devices within the FCF

  9. Finite element modeling of fluid/thermal/structural interaction for a gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    Bennett, J.G.; Ju, F.D.

    1980-01-01

    Two nonlinear finite element formulations for application to a series of experiments in the Gas-Cooled Fast Reactor (GCFR) development program are described. An efficient beam column element for moderately large deformations is combined with a finite element developed for an engineering description of a convecting fluid. Typical results from both elements are illustrated. A combined application for a problem typical of the GCFR loss-of-coolant experiments is illustrated. These problems are not the usual fluid structural interaction problems in that the inertia coupling is negligible while the thermal coupling is very important

  10. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    In this conference are presented papers dealing with swelling of metals and alloys, (and specially ferritic steels), structural evolution and stability under irradiation, modifications of mechanical properties, consequences on the behaviour of fuel elements and the optimization of materials selection, and irradiation creep [fr

  11. Experimental and numerical analysis of fluid - structure interaction effects in a fast reactor core

    International Nuclear Information System (INIS)

    Martelli, A.; Forni, M.; Melloni, R.; Paoluzzi, R.; Bonacina, G.; Castoldi, A.; Zola, M.

    1990-01-01

    Dynamic experiments in air and water (simulating liquid sodium) were performed by ISMES, on behalf of ENEA, on various core element groups of the Italian PEC fast reactor. Bundles of one, seven and nineteen mock-ups reproducing fuel, reflecting and neutron shield elements in full scale were analysed on shaking tables. Tests concerned both groups of equal elements and mixed configurations which corresponded to real core parts. The effects of PEC core-restraint ring were also studied. Seismic excitations of up to 2.5 g were applied to core diagrid. Test results were analysed by use of the one-dimensional program CORALIE and the two-dimensional program CLASH. The study allowed the fluid effects in the PEC core to be evaluated; it also contributed to validation of the above mentioned programs for their general use for fast reactor core analysis. This paper presents the main features of the experimental and the numerical studies and reports comparisons between calculations and measurements. (author)

  12. The development of fuel pins and material specimens mixed loading irradiation test rig in the experimental fast reactor Joyo. The development of the fuel-material hybrid rig

    International Nuclear Information System (INIS)

    Oyamatsu, Yasuko; Someya, Hiroyuki

    2013-02-01

    In the experimental fast reactor Joyo, there were many tests using the irradiation rigs that it was possible to be set irradiation conditions for each compartment independently. In case of no alternative fuel element to irradiate after unloading the irradiated compartments, the irradiation test was restarted with the dummy compartment which the fuel elements was not mounted. If the material specimens are mounted in this space, it is possible to use the irradiation space effectively. For these reasons, the irradiation rig (hybrid rig) is developed that is consolidated with material specimens compartment and fuel elements compartment. Fuel elements and material specimens differ greatly with heat generation, so that the most important issue in developing of hybrid rig is being able to distribute appropriately the coolant flow which satisfies irradiation conditions. The following is described by this report. (1) It was confirmed that the flow distribution of loading the same irradiation rig with the compartment from which a flow demand differs could be satisfied. (2) It was confirmed that temperature setting range of hybrid rig could be equivalent to that of irradiation condition. (3) By standardizing the coolant entrance structure of the compartment lower part, the prospect which can perform easily recombination of the compartment from which a type differs between irradiation rigs was acquired. (author)

  13. Evaluation of Microstructural and Mechanical Property of Medium-sized HT9 Cladding Forged Material for Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Lee, Kang Soo; Kim, Sung Ho; Lee, Chan Bock

    2012-01-01

    Microstructural and mechanical property were evaluated at the medium-sized HT9 (12Cr-1MoWV) forged steel which was considered as primary candidate for the fuel cladding in sodium-cooled fast reactor (SFR). Material was forged at 1170 degrees C after the induction melting to make round bar as 160 mm diameter, 7000 mm length then the radial distribution of microstructure as well as microhardness was evaluated. The results showed that overall microstructure exhibited as ferrite-martensite structure, where small amount (2-3%) of delta ferrite was formed throughout the specimen and maximum 15% of transformed ferrite was formed at the center, where it gradually decreased toward the radial direction. Sensitivity analysis of the cooling curve and Time-Temperature-Transformation (TTT) diagram revealed that formation of transformed ferrite could be avoided when the diameter was decreased down to 120 mm.

  14. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

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

  15. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

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

  16. Development of materials and manufacturing technologies for Indian fast reactor programme

    International Nuclear Information System (INIS)

    Raj, Baldev; Jayakumar, T.; Bhaduri, A.K.; Mandal, Sumantra

    2010-01-01

    Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India. The development of FBRs necessitates extensive research and development in domains of materials and manufacturing technologies in association with a wide spectrum of disciplines and their inter-twining to meet the challenging technology. The paper highlight the work and the approaches adopted for the successful deployment of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme. Indigenous development of in-core materials viz. Titanium modified austenitic stainless steel (Alloy D9) and its variants, ferritic/martensitic oxide-dispersion strengthened (ODS) steels as well as structural materials viz. 316L(N) stainless steel and modified 9Cr-1Mo have been achieved through synergistic interactions between Indira Gandhi Centre for Atomic Research (IGCAR), education and research institutes and industries. Robust manufacturing technology has been established for forming and joining of various components of 500 MWe Prototype Fast Breeder Reactor (PFBR) through 'science-based technology' approach. To achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plantes. Optimization of grain boundary character distribution in Alloy D9 was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation. Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel sub-assemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required for testing

  17. Development of materials and manufacturing technologies for Indian fast reactor programme

    Energy Technology Data Exchange (ETDEWEB)

    Raj, Baldev; Jayakumar, T.; Bhaduri, A.K.; Mandal, Sumantra [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2010-07-01

    Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India. The development of FBRs necessitates extensive research and development in domains of materials and manufacturing technologies in association with a wide spectrum of disciplines and their inter-twining to meet the challenging technology. The paper highlight the work and the approaches adopted for the successful deployment of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme. Indigenous development of in-core materials viz. Titanium modified austenitic stainless steel (Alloy D9) and its variants, ferritic/martensitic oxide-dispersion strengthened (ODS) steels as well as structural materials viz. 316L(N) stainless steel and modified 9Cr-1Mo have been achieved through synergistic interactions between Indira Gandhi Centre for Atomic Research (IGCAR), education and research institutes and industries. Robust manufacturing technology has been established for forming and joining of various components of 500 MWe Prototype Fast Breeder Reactor (PFBR) through 'science-based technology' approach. To achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plantes. Optimization of grain boundary character distribution in Alloy D9 was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation. Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel sub-assemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required

  18. Study of burned optimization for minor actinides in European Sodium Fast Reactor (ESFR) by use of moderator materials

    International Nuclear Information System (INIS)

    Ramos, R L; Villanueva, A J; Buiront, L

    2012-01-01

    The minor actinides (MA) burn up optimization in the European Sodium Fast Reactor (ESFR) core was studied by adding different moderating materials in the Minor Actinides Bearing Blanket subassemblies (MABB SA) using the ERANOS neutron code package. These SA are of hexagonal shape and are composed of pellets inside of pins. These pellets contain a mixture of uranium dioxide (UO 2 ) and americium dioxide (AmO 2 ). If some of these pins are replaced by other identical ones containing moderating material instead of minor actinides, a shift in the spectrum towards lower energies is expected, which might enhance the burn up performance. The results of this work demonstrated that the use of compounds of hydrogen and magnesium as moderators produces a shift in the neutron spectrum, improving the porcentual minor actinides consumption. ZrH 2 moderator material was found to exhibit the best performances for this propose, followed by MgO and MgAl 2 O 4 , in that order. The use of SiC, BeO, TiC, LiO 2 and ZrC material produced no effect on the shift of the neutron spectrum. For safety reasons, it seems hardly realistic to use hydrogenous compounds in sodium fast reactors. So, compounds with magnesium are selected to be placed into the pins to improve the porcentual minor actinides consumption. The ESFR core is composed by 817 SA, 453 of them are fuel SA, 247 are reflectors SA, 84 are MABB (Minor Actinides Bearing Blankets) SA and 33 are control and shutdown rods. When about half of the total pins in each MABB were substituted by moderator pins with MgO pellets (135 of 271 pins), the porcentual consumption of minor actinides was of 30.85 %, i.e., 227.22 kg of minor actinides were consumed out of 736.65 kg in the initial configuration. In the case where all the pins of the MABB contained pellets of minor actinides, the porcentual consumption of minor actinides was of 21.26 %, i.e., 312.13 kg of minor actinides were consumed of 1467.87 kg in the initial configuration (author)

  19. Review of fast reactor activities in India

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1982-01-01

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

  20. Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

    Science.gov (United States)

    Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

    2015-12-01

    U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

  1. Forty-Fifth Annual Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Working Material

    International Nuclear Information System (INIS)

    2012-01-01

    The objectives of the meeting were to: • Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); • Review the progress since the 44 th TWG-FR Annual Meeting, including the status of the actions; • Consider topical technical meeting arrangements for 2012-2013, as well as review FR-related activities included in the IAEA Programme & Budget (P&B) biennium 2012-2013; • Review the IAEA’s concluded, on-going and planned coordinated research projects in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations and international initiatives (GIF, INPRO, NEA, ESNII, etc.). The 45th Meeting of the TWG-FR reached the following conclusions/recommendations: • The participants expressed satisfaction and appreciation for the large amount of new information on on-going activities carried out by the Member States in the field of FR and ADS exchanged during the meeting; • Also the organizations which have participated to the TWG-FR meeting for the first time expressed their appreciation for the lively discussion and the results and thanked the IAEA for inviting them at the meeting; • The meeting was very useful in particular for collecting inputs and advice in view of the preparation of the IAEA Programme & Budget 2014-2015 (and then 2016-2017) in the area of FR and ADS technology development; • The TWG-FR remains an unique international forum for information exchange in the field of fast neutron systems and for promoting RT&D activities in this area; • Due to the increasing interest in FR and in view of the forthcoming realizations, it would be advisable to increase the involvement of industries, regulators and other R&D organizations; • The annual TWG-FR meeting should focused on exchange of information on national and international programmes, avoiding duplications or overlapping’s with other IAEA initiatives in the field;

  2. Micro-structural study and Rietveld analysis of fast reactor fuels: U–Mo fuels

    International Nuclear Information System (INIS)

    Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K.B.; Kumar, Arun

    2015-01-01

    U–Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U–Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U–Mo alloys as fast reactor fuel. - Highlights: • U–Mo alloys in as-cast as well as in annealed conditions have been studied using Optical Microscope, SEM, XRD. • The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. • The dendritic microstructure of γ-(U,Mo) and B.C.C. ‘Mo’ phase of 33 at.% U–Mo alloy have been analysed. • Rietveld analysis has been done to optimize lattice parameters and calculate phase fractions in annealed alloys. • The Vickers microhardness of U_2Mo phase shows lower hardness than two phase microstructures in annealed alloys.

  3. Micro-structural study and Rietveld analysis of fast reactor fuels: U–Mo fuels

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, S., E-mail: sibasis@barc.gov.in [Radiometallurgy Division, Bhabha Atomic Research Centre, Mumbai, 400085 (India); Choudhuri, G. [Atomic Fuels Division, Bhabha Atomic Research Centre, Mumbai, 400085 (India); Banerjee, J. [Radiometallurgy Division, Bhabha Atomic Research Centre, Mumbai, 400085 (India); Agarwal, Renu [Product Development Division, Bhabha Atomic Research Centre, Mumbai, 400085 (India); Khan, K.B.; Kumar, Arun [Radiometallurgy Division, Bhabha Atomic Research Centre, Mumbai, 400085 (India)

    2015-12-15

    U–Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U–Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U–Mo alloys as fast reactor fuel. - Highlights: • U–Mo alloys in as-cast as well as in annealed conditions have been studied using Optical Microscope, SEM, XRD. • The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. • The dendritic microstructure of γ-(U,Mo) and B.C.C. ‘Mo’ phase of 33 at.% U–Mo alloy have been analysed. • Rietveld analysis has been done to optimize lattice parameters and calculate phase fractions in annealed alloys. • The Vickers microhardness of U{sub 2}Mo phase shows lower hardness than two phase microstructures in annealed alloys.

  4. A review of the United Kingdom fast reactor programme: March 1987

    International Nuclear Information System (INIS)

    Bramman, J.I.; Wheeler, R.C.

    1987-03-01

    The UK fast reactor programme 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 Studies. (U.K.)

  5. Minor Actinides Burnup Enhancement in the European Sodium Fast Reactor through Moderator Material Addition

    International Nuclear Information System (INIS)

    Ramos, R.L.; Buiron, L.

    2013-01-01

    Conclusions: • ZrH 2 was the best moderator material, followed by MgO and MgAl 2 O 4 ; • When the number of moderator pins is increased: – the percentage of minor actinides consumed increases; – the total mass consumed of minor actinides decreases; – the decay heat generated decreases; – the neutron flux in the reactor varies very little. Perspectives: • For future studies it would be possible to evaluate the use of other materials with resonances in the scattering cross section in the fast range that would improve the results obtained with Mg. • It would be necessary to consider how to add moderator material without changing the initial mass of minor actinides. E.g., adding the moderator at the periphery of the minor actinide elements

  6. On the Use of Fine Distributed Moderating Material to Enhance Feedback Coefficients in Fast Reactors

    International Nuclear Information System (INIS)

    Merk, B.

    2012-01-01

    Conclusions and outlook: • Use of moderating material to enhance feedback coefficients in SFR: • Creation of a new degree of freedom for SFR design; • Good opportunities for the compensation of the effects due to transmutation fuels; • No major influence on Am incinieration; • Major problem is the thermal stability; • Stability up to ~1300°C by use of YH

  7. Design study of an IHX support structure for a POOL-TYPE Sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

    2009-01-01

    The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity

  8. Technical meeting to 'Review of national programmes on fast reactors and accelerator driven systems (ADS)'. Working material

    International Nuclear Information System (INIS)

    2002-01-01

    The 35th Annual Meeting of the Technical Working Group on Fast Reactors TWG-FR, previously International Working Group on Fast Reactors (IWG-FR, created in 1967), was hosted by the Forschungszentrum Karlsruhe (FZK) and was attended by TWG-FR members and advisers from the following Member States: Brazil, China, France, Germany, India, Japan, the Republic of Kazakhstan, the Republic of Korea, the Russian Federation, and the United States of America. The objectives of the meeting were: to exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); to review the progress since the 34th TWG-FR Annual Meeting, including the status of the actions; to consider meeting arrangements for 2002 and 2003; to review the Agency's co-ordinated research activities in the field of FRs and ADS, as well as co-ordination of the TWG-FR's activities with other organizations

  9. Technical meeting to 'Review of national programmes on fast reactors and accelerator driven systems (ADS)'. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    The 35th Annual Meeting of the Technical Working Group on Fast Reactors TWG-FR, previously International Working Group on Fast Reactors (IWG-FR, created in 1967), was hosted by the Forschungszentrum Karlsruhe (FZK) and was attended by TWG-FR members and advisers from the following Member States: Brazil, China, France, Germany, India, Japan, the Republic of Kazakhstan, the Republic of Korea, the Russian Federation, and the United States of America. The objectives of the meeting were: to exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); to review the progress since the 34th TWG-FR Annual Meeting, including the status of the actions; to consider meeting arrangements for 2002 and 2003; to review the Agency's co-ordinated research activities in the field of FRs and ADS, as well as co-ordination of the TWG-FR's activities with other organizations.

  10. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Smith, R.D.

    1982-01-01

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

  11. Characterization and management of radioactive sodium and other reactor components as input data for the decommissioning of liquid metal-cooled fast reactors. A compilation of data produced of data produced by members of the IAEA technical working group on fast reactors (TWG-FR) at two consultancies and one technical committee meeting. Working material

    International Nuclear Information System (INIS)

    2002-01-01

    A number of liquid metal cooled fast reactors (LMFRs) are in operation and, some have already been shut down; other reactors will reach the end of their design lifetime in a few years and become candidates for decommissioning. It is unfortunate that little consideration was devoted to decommissioning of reactors at the plant design and construction stage. It is with this focus that the Technical Working Group on Fast Reactors (TWGFR) recommended that the IAEA organize the exchange of information on LMFRs decommissioning technology. It was pointed out that the decommissioning of small sodium-cooled reactors has shown that there are two basic differences between thermal and fast reactors decommissioning: on the one side, the treatment and disposal of radioactive sodium coolant, and on the other side, the management of reactor components, for which the structural materials are activated in depth by fast neutrons. To this end, a Technical Committee Meeting on Sodium Removal and Disposal from LMFRs in Normal Operation and in the framework of Decommissioning (Aix-en-Provence, France, November 1997) and two Consultancies on Decommissioning of the Kazakh BN-350 LMFR (Vienna, Austria, October 1996; Obninsk, Russian Federation, February 1998) were convened by the IAEA. These Meetings brought together a group of experts from France, Russia, Kazakhstan, the UK, and the USA to exchange information on, and to review current technical knowledge and experience in the management of radioactive coolant and reactor components following closing of LMFRs, as well as their design features and operating experience relevant for decommissioning procedures. The report provides general and detailed information on activation characteristics of the primary coolant; treatment and disposal of the spent sodium; removal of the residual sodium deposits and decontamination; the activation characteristics of the reactor components and the management of the latter. The recurring theme is finding

  12. Fast reactors worldwide

    International Nuclear Information System (INIS)

    Hall, R.S.; Vignon, D.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1979-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R D

    1979-07-01

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

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

    International Nuclear Information System (INIS)

    Bramman, J.I.; John, C.T.; Wheeler, R.C.

    1986-01-01

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

  16. Structural assessment of intermediate printed circuit heat exchanger for sodium-cooled fast reactor with supercritical CO2 cycle

    International Nuclear Information System (INIS)

    Lee, Youho; Lee, Jeong Ik

    2014-01-01

    Highlights: • We numerically model PCHE stress arising from pressure, and thermal loadings. • Stress levels are the highest around S-CO 2 channels, due to high pressure of S-CO 2 . • The conventional analytic models for PCHE underestimate actual stress levels. • Plasticity sufficiently lowers stress levels at channel tips. • PCHE for SFR-SCO 2 is anticipated to assure compliance with ASME design standards. - Abstract: Structural integrity of intermediate Printed Circuit Heat Exchanger (PCHE) for Sodium-cooled Fast Reactor (SFR) attached to Supercritical CO 2 (S-CO 2 ) is investigated. ANSYS-Mechanical was used to simulate stress fields of representative PCHE channels, with temperature fields imported from FLUENT simulation. Mechanical stress induced by pressure loading is found to be the primary source of stress. As plasticity sufficiently lowers local stress concentration at PCHE channel tips, PCHE type intermediate heat exchangers made of SS316 are anticipated to reliably assure compliance with design standards prescribed in the ASME standards, thanks to the structure temperature that is below the effective creep inducing point. The actual life time of PCHE for SFR-SCO 2 is likely to be affected by mechanical behavior change of SS316 with reactions with S-CO 2 and fatigue

  17. Kick-off research coordination meeting of the IAEA coordinated research project on analyses of, and lessons learned from the operational experience with fast reactor equipment and systems. Working material

    International Nuclear Information System (INIS)

    2007-01-01

    steam generators (SG) and materials (fuels and structural materials), and include the following specific activities: - Retrieval of the documentation and information relevant to feedback from commissioning, operation, and decommissioning of sodium cooled fast reactors relating to the above mentioned areas (SG, fuels, and structural materials); - Establishing bibliographic catalogues of these documents. The second stage of the CRP will consist in the following activities: - Publishing national synthesis reports for each of the above mentioned areas; - Publishing joint synthesis reports for each of the above mentioned areas ('final CRP reports'). Should the participants, during the implementation of the CRP activities, wish to extend the scope of the CRP to other areas (e.g. sodium technology, in-service inspection and repair ISI and R), and systems and equipments), appropriate proposals will be submitted and discussed among the participants. In principle, the IAEA could consider an extension of the CRP beyond three years and/or a new CRP covering the new topical areas of interest to participating Member States

  18. On the problem of heat and mass exchange between liquid metal surface and structural elements in fast reactors

    International Nuclear Information System (INIS)

    Rineisky, A.A.; Sorokin, A.P.; Yatsenko, M.G.

    1986-01-01

    For the development of means ensuring normal operating conditions of the fast reactor vessel some design procedures for calculating temperature conditions of its structural elements over the liquid metal surface are required. The radiative heat transfer from the liquid metal surface playing an important part at working temperatures (550 deg. C), the effect of experimentally detected fog formation process (not taken into account before) upon the radiative heat exchange has been considered. A simplified heat transfer model based upon separation of thin thermal boundary layers and of the main volume at a constant temperature has been proposed. Calculation relationships for the heat flux from the reactor vessel roof have been obtained by solving a one-dimensional equation of radiation transfer within the boundary layer and a three-dimensional one in the bulk volume at an approximation of moments with Marshak boundary conditions. Evaluations performed have shown a possibility of a considerable decrease of the vessel roof temperature due to fog formation. The observed asymmetry of the temperature distribution in the cover gas is explained in this case greater fog density near the evaporation surface and by a possibility of some radiative energy loss due to evaporation from the droplets surface. (author)

  19. Technical meeting to 'Review of national programmes on fast reactors and accelerator driven systems (ADS)'. Working material

    International Nuclear Information System (INIS)

    2003-01-01

    36th Annual Meeting of the Technical Working Group on Fast Reactors, the IAEA Technical Meeting (TM) on 'Review of National Programmes on Fast Reactors and Accelerator Driven Systems (ADS)', hosted by the Korean Atomic Energy Research Institute (KAERI) was attended by TWG-FR Members and Advisers from the following Member States (MS) and International Organizations: Brazil, France, Germany, India, Japan, the Republic of Kazakhstan, the Republic of Korea, the Russian Federation, the United Kingdom, the United States of America, and the OECD/NEA. The objectives of the meeting were to: 1) exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); 2) review the progress since the 35th TWG-FR Annual Meeting, including the status of the actions; 3) consider meeting arrangements for 2003 and 2004; 4) review the Agency's co-ordinated research activities in the field of FRs and ADS, as well as co-ordination of the TWG-FR's activities with other organizations. The participants made presentations on the status of the respective national programmes on FR and ADS development. A summary of the highlights for the period since the 35th TWG-FR Annual Meeting

  20. Technical meeting (TM) to 'Review of national programmes on fast reactors and accelerator driven systems (ADS)'. Technical Working Group on Fast Reactors (TWG-FR) (37th annual meeting). Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The objectives of the 37th Annual Meeting of the Technical Working Group on Fast Reactors, were to: 1) exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); 2) review the progress since the 36th TWG-FR Annual Meeting, including the status of the actions; 3) consider meeting arrangements for 2004 and 2005; 4) review the Agency's co-ordinated research activities in the field of FRs and ADS, as well as co-ordination of the TWG-FR's activities with other organizations. The participants made presentations on the status of the respective national programmes on FR and ADS development. A summary of the highlights for the period since the 36th TWG-FR Annual Meeting is included in this proceedings. Annex IV contains the Review of National Programs on Fast Reactors and Accelerator Driven Systems (ADS), and the TWG-FR Activity Report for the Period May 2003-April 2004.

  1. On establishing constitutive equations for use in design of high-temperature fast-reactor structures

    International Nuclear Information System (INIS)

    Pugh, C.E.

    1978-01-01

    The presentation describes the approach being used to establish constitutive equations for wide use in the design of fast breeder reactor (FBR) components in the US. The approach combines exploratory experiments, constitutive model studies, studies of computational techniques, and tests of simple structural configurations. Short-time (elastic-plastic) behavior, long-time (creep) behavior, and their interactions are considered, and some of the background to equations now identified for use in current FBR design applications involving three structural alloys is discussed. Comments are also given on current efforts aimed at identifying improved constitutive equations for these alloys and on properties data required for design applications. References are cited which have addressed the status of the process at various times. (Auth.)

  2. Integral Fast Reactor concept

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  3. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-01-01

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

  4. The Integral Fast Reactor

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  5. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

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

    1993-06-01

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

  6. Fast reactor fuel design and development

    International Nuclear Information System (INIS)

    Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

    1977-01-01

    Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

  7. Model for determining stresses in the structure of a fast reactor fuel assembly

    International Nuclear Information System (INIS)

    Kervevan, J.-J.

    1974-01-01

    Deformations in a reactor core are due to two metallurgical phenomena, swelling of the steel under irradiation and irradiation creep when the structure is under stress. The first step is to determine the deformation of each sub-assembly supposed free, subjected to a neutron flux or temperature gradient, and the second is to study the interactions amongst most of the sub-assemblies. Under the influence of the deformations the interaction value will change with time, and this development must be determined. Calculation methods were developed for the purpose. A number of computing codes already exist and it is necessary to complete them, modify them if necessary, create new ones as the case arises and form a coherent whole [fr

  8. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, LianCheng; Zhang, Bin

    2016-01-01

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  9. Forty-Second Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Working Material

    International Nuclear Information System (INIS)

    2009-01-01

    The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 41st TWG-FR Annual Meeting, including the status of the actions; - Consider topical technical meetings meeting arrangements for 2009, 2010, 2011 and beyond; - Review the IAEA’s ongoing information exchange and coordinated research activities in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations and international initiatives; - Discuss future joint activities in view of IAEA’s Programme and Budget Cycles beyond 2010-2011

  10. Meeting of the Technical Working Group on Fast Reactors (TWG-FR) (41st Annual Meeting). Working Material

    International Nuclear Information System (INIS)

    2008-01-01

    The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 40th TWG-FR Annual Meeting, including the status of the actions; - Consider meeting arrangements for 2008, 2009, 2010 and beyond; - Review the IAEA’s ongoing information exchange and coordinated research activities in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations and international initiatives; - Discuss future joint activities in view of IAEA’s Programme and Budget Cycles beyond 2008-2009

  11. Technical committee on review of national programmes on fast reactors and accelerator driven systems (ADS). Working material

    International Nuclear Information System (INIS)

    2001-01-01

    The objectives of the meeting were: to exchange information on the national programmes on fast reactors (FR) and accelerator driven systems (ADS); to review the progress since the previous IWG-FR meeting, including the status of the actions; to consider meeting arrangements for 2001 and 2002; to review the Agency co-ordinated research activities in the field of FR and ADS, as well as so-ordination of the TWG-FR activities with their organisations. This report covers the reports presented on the relevant activities in Brazil, China, France, Germany, India, Italy, Japan, Kazakhstan, Republic of Korea, Russia, Sweden, United Kingdom and USA

  12. Power from plutonium: fast reactor fuel

    International Nuclear Information System (INIS)

    Bishop, J.F.W.

    1981-01-01

    Points of similarity and of difference between fast reactor fuel and fuels for AGR and PWR plants are established. The flow of uranium and plutonium in fast and thermal systems is also mentioned, establishing the role of the fast reactor as a plutonium burner. A historical perspective of fast reactors is given in which the substantial experience accumulated in test and prototype is indicated and it is noted that fast reactors have now entered the commercial phase. The relevance of the data obtained in the test and prototype reactors to the behaviour of commercial fast reactor fuel is considered. The design concepts employed in fuel are reviewed, including sections on core support styles, pin support and pin detail. This is followed by a discussion of current issues under the headings of manufacture, performance and reprocessing. This section includes a consideration of gel fuel, achievable burn-up, irradiation induced distortions and material choices, fuel form, and fuel failure mechanisms. Future development possibilities are also discussed and the Paper concludes with a view on the logic of a UK fast reactor strategy. (U.K.)

  13. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aufiero, M.; Cammi, A.; Fiorina, C. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Leppänen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Ricotti, M.E. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy)

    2013-10-15

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  14. Forty-Sixth Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were to: • Review the current status and the progress since the 45th TWG-FR meeting of FR and ADS technology development activities in IAEA Member States; • Review the activities (past, present and planned) of the IAEA’s project 1.1.5.3, “Support for fast reactor research, technology development and deployment” to ensure that they remain relevant to the needs of Member States; • Provide the experts group with updates to advise the IAEA on FR and ADS activities, including on proposals for relevant studies and reviews; • Serve as a means for exchanging information on national and international FR and ADS programmes; • Review the main achievements and outcomes of the “International Conference on Fast Reactors and Related Fuel Cycle: Safe Technologies and Sustainable Scenarios – FR13”, held on 4 – 7 March 2013 in Paris, France; • Promote the exchange of technical information by proposing topics for, and assisting in the organization of, IAEA Workshops and Technical Meetings for 2014-2015 and further, and • Review the IAEA’s concluded, on-going and planned coordinated research projects (CRPs) in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations and international initiatives (GIF, INPRO, NEA, Euratom, etc.)

  15. Meeting of the Technical Working Group on Fast Reactors (TWG-FR) (39th annual meeting). Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    The 39th Annual Meeting of the Technical Working Group on Fast Reactors (TWG FR) was held from 15-19 May 2006 in Beijing, China, at the invitation of the China Institute of Atomic Energy (CIAEA). The meeting was attended by TWG-FR Members and Advisers from the following Member States (MS): Belgium (observer), Brazil, China, France, Germany, India, Italy, Japan, the Republic of Kazakhstan, the Republic of Korea, the Russian Federation, Sweden (observer), the United Kingdom, and the United States. Belarus, Switzerland, the European Commission, and OECD/NEA were unable to participate. Moreover, Prof. Carlo Rubbia, CERN director general emeritus, participated, upon IAEA invitation, in the meeting as distinguished scientist and IAEA expert. Mr. S.C. Chetal, from India (Indira Gandhi Centre for Atomic Research, IGCAR), was appointed chairman. The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 38th TWG-FR Annual Meeting, including the status of the actions; - Consider meeting arrangements for 2006 and 2007; - Reviewed the Agency's ongoing information exchange and co-ordinated research activities in the technical fields relevant to the TWG-FR (FRs and ADS), as well as co-ordination of the TWG-FR's activities with other organizations; - Discuss future joint activities in view of the Agency's Programme and Budget Cycle 2008-2009 (and beyond)

  16. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    1983-02-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  18. Integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

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

  19. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    1992-01-01

    'burning' of the associated extremely long-life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. This additional important mission for the LMFBR is gaining worldwide interest. In the framework of disarmament of nuclear weapons and the utilization of the nuclear material for peaceful purposes a role for fast reactors can be also considered. Over the past 25 years, the IAEA has actively encouraged and advocated international cooperation in Fast Breeder Reactor Technology. At the present time the Working Group on Fast Reactors is the oldest and one of the most active groups in the Division of Nuclear Power. The present document contains information on the status of fast breeder reactor development and on worldwide activities in this advanced nuclear power technology during 1991, as reported at the 25th jubilee Annual Meeting of the IWGFR in Vienna, 27-30 April 1992. The publication is intended to provide information regarding the current status of LMFBR development in IAEA Member States and CEC

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    'burning' of the associated extremely long-life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. This additional important mission for the LMFBR is gaining worldwide interest. In the framework of disarmament of nuclear weapons and the utilization of the nuclear material for peaceful purposes a role for fast reactors can be also considered. Over the past 25 years, the IAEA has actively encouraged and advocated international cooperation in Fast Breeder Reactor Technology. At the present time the Working Group on Fast Reactors is the oldest and one of the most active groups in the Division of Nuclear Power. The present document contains information on the status of fast breeder reactor development and on worldwide activities in this advanced nuclear power technology during 1991, as reported at the 25th jubilee Annual Meeting of the IWGFR in Vienna, 27-30 April 1992. The publication is intended to provide information regarding the current status of LMFBR development in IAEA Member States and CEC.

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R D [United Kingdom Atomic Energy Authority, Risley (United Kingdom)

    1978-07-01

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

  4. Forty-Fourth Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Working Material

    International Nuclear Information System (INIS)

    2011-01-01

    The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 43rd TWG-FR Annual Meeting, including the status of the actions; - Consider topical technical meeting arrangements for 2012-2013, as well as review FR-related activities included in the IAEA Project&Budget (P&B) biennium 2012-2013; - Review the IAEA’s ongoing information exchange and coordinated research projects in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations and international initiatives (GIF, INPRO, NEA, ESNII, etc.)

  5. 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

  6. The instrumentation of fast reactor

    International Nuclear Information System (INIS)

    Endo, Akira

    2003-03-01

    The author has been engaged in the development of fast reactors over the last 30 years with both an involvement with the early technology development on the experimental breeder reactor Joyo, and latterly continuing this work on the prototype breeder reactor, Monju. In order to pass on this experience to younger engineers this paper is produced to outline this experience in the sincere hope that the information given will be utilised in future educational training material. The paper discusses the wide diversity on the associated instrument technology which the fast breeder reactor requires. The first chapter outlines the fast reactor system, followed by discussions on reactor instrumentation, measurement principles, temperature dependencies, and verification response characteristics from various viewpoints, are discussed in chapters two and three. The important issues of failed fuel location detection, and sodium leak detection from steam generators are discussed in chapters 4 and 5 respectively. Appended to this report is an explanation on the methods of measuring response characteristics on instrumentation systems using error analysis, random signal theory and measuring method of response characteristic by AR (autoregressive) model on which it appears is becoming an indispensable problem for persons involved with this technology in the future. (author)

  7. A numerical design and feasibility study of self-wastage experiment using simulant material in a sodium fast reactor

    International Nuclear Information System (INIS)

    Jang, Sung Hyun; Takata, Takashi; Yamaguchi, Akira

    2016-01-01

    A sodium-water reaction takes place when high-pressured water vapor leaks into sodium through a tiny defect on the surface of the heat transfer tube in a steam generator of the sodium-cooled fast reactor. The sodium-water reaction brings deterioration of the mechanical strength of the heat transfer tube at the initial leakage site. As a result, it damages the crack itself, which may eventually enlarge into a larger opening. This self-enlargement is called 'self-wastage phenomenon.' In this study, a simulant experiment was proposed to reproduce the self-enlargement of a crack and to evaluate the mechanism of the self-wastage. The damage on the surface of the crack was simulated by making the neutralization reaction with hydrochloric acid solution and sodium hydroxide solution. A numerical investigation was carried out to validate the feasibility of the approach and to determine experimental conditions. From the computation results, it is observed that when 5M HCl is injected into 5M of NaOH with 0.05 m/s inlet velocity, the temperature at the surface near the crack increased over 319.26 K. The computational results show that the self-wastage phenomenon is capable of being reproduced by the simulant experiment

  8. A Numerical Design and Feasibility Study of Self-Wastage Experiment Using Simulant Material in a Sodium Fast Reactor

    Directory of Open Access Journals (Sweden)

    Sunghyon Jang

    2016-04-01

    Full Text Available A sodium–water reaction takes place when high-pressured water vapor leaks into sodium through a tiny defect on the surface of the heat transfer tube in a steam generator of the sodium-cooled fast reactor. The sodium–water reaction brings deterioration of the mechanical strength of the heat transfer tube at the initial leakage site. As a result, it damages the crack itself, which may eventually enlarge into a larger opening. This self-enlargement is called “self-wastage phenomenon.” In this study, a simulant experiment was proposed to reproduce the self-enlargement of a crack and to evaluate the mechanism of the self-wastage. The damage on the surface of the crack was simulated by making the neutralization reaction with hydrochloric acid solution and sodium hydroxide solution. A numerical investigation was carried out to validate the feasibility of the approach and to determine experimental conditions. From the computation results, it is observed that when 5M HCl is injected into 5M of NaOH with 0.05 m/s inlet velocity, the temperature at the surface near the crack increased over 319.26 K. The computational results show that the self-wastage phenomenon is capable of being reproduced by the simulant experiment.

  9. A numerical design and feasibility study of self-wastage experiment using simulant material in a sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Sung Hyun; Takata, Takashi [Graduate School of Engineering, Osaka University, Osaka (Japan); Yamaguchi, Akira [Nuclear Professional School, The University of Tokyo, Ibaraki (Japan)

    2016-04-15

    A sodium-water reaction takes place when high-pressured water vapor leaks into sodium through a tiny defect on the surface of the heat transfer tube in a steam generator of the sodium-cooled fast reactor. The sodium-water reaction brings deterioration of the mechanical strength of the heat transfer tube at the initial leakage site. As a result, it damages the crack itself, which may eventually enlarge into a larger opening. This self-enlargement is called 'self-wastage phenomenon.' In this study, a simulant experiment was proposed to reproduce the self-enlargement of a crack and to evaluate the mechanism of the self-wastage. The damage on the surface of the crack was simulated by making the neutralization reaction with hydrochloric acid solution and sodium hydroxide solution. A numerical investigation was carried out to validate the feasibility of the approach and to determine experimental conditions. From the computation results, it is observed that when 5M HCl is injected into 5M of NaOH with 0.05 m/s inlet velocity, the temperature at the surface near the crack increased over 319.26 K. The computational results show that the self-wastage phenomenon is capable of being reproduced by the simulant experiment.

  10. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

    International Nuclear Information System (INIS)

    Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

    2014-01-01

    An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

  11. On-site experimental dynamic analysis for evaluating the soil-structure interaction and the seismic behaviour of the Italian PEC fast reactor building

    International Nuclear Information System (INIS)

    Casirati, M.; Castoldi, A.; Panzeri, P.; Pezzoli, P.; Martelli, A.; Masoni, P.; Brancati, V.

    1988-01-01

    The paper describes the on-site dynamic tests carried out on the PEC fast reactor building, using various excitation methods (two eccentric back-rotating-mass mechanical vibrator, blasting in bore-hole, hydraulic actuators at the building foundations). It points out the purposes of the four tests campaigns performed at various construction stages and reports the main experimental results. These results concern both the design safety margins and the data for the validation of the three-dimensional numerical model of the reactor building, including soil-structure interaction phenomena. (author)

  12. New fast reactor installation concept

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The large size and complexity of fast reactor installations are emphasised and these difficulties will be increased with the advent of fast reactors of higher power. In this connection a new concept of fast reactor installation is described with a view to reducing the size of the installation and enabling most components, including even the primary vessel, to be constructed within the confines of a workshop. Full constructional details are given. (U.K.)

  13. Fast reactors in nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkovskii, O

    1981-02-01

    The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

  14. Knowledge management in fast reactors

    International Nuclear Information System (INIS)

    Kuriakose, K.K.; Satya Murty, S.A.V.; Swaminathan, P.; Raj, Baldev

    2010-01-01

    This paper highlights the work that is being carried out in Knowledge Management of Fast Reactors at Indira Gandhi Centre for Atomic Research (IGCAR) including a few examples of how the knowledge acquired because of various incidents in the initial years has been utilized for the successful operation of Fast Breeder Test Reactor. It also briefly refers to the features of the IAEA initiative on the preservation of Knowledge in the area of Fast Reactors in the form of 'Fast Reactor Knowledge Organization System' (FR-KOS), which is based on a taxonomy for storage and mining of Fast Reactor Knowledge. (author)

  15. Fast reactor research activities in Brazil

    International Nuclear Information System (INIS)

    Menezes, A.

    1998-01-01

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

  16. Dounreay fast reactor

    International Nuclear Information System (INIS)

    Maclennan, R.; Eggar, T.; Skeet, T.

    1992-01-01

    The short debate which followed a private notice question asking for a statement on Government policy on the future of the European fast breeder nuclear research programme is reported verbatim. In response to the question, the Minister for Energy said that the Government had decided in 1988 that the Dounreay prototype fast reactor would close in 1994. That decision had been confirmed. Funding of fast breeder research and development beyond 1993 is not a priority as commercialization is not expected until well into the next century. Dounreay will be supported financially until 1994 and then for its subsequent decommissioning and reprocessing of spent fuel. The debate raised issues such as Britain losing its lead in fast breeder research, loss of jobs and the Government's nuclear policy in general. However, the Government's position was that the research had reached a stage where it could be left and returned to in the future. (UK)

  17. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1985-01-01

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

  18. Uranium and the fast reactor

    International Nuclear Information System (INIS)

    Price, T.

    1982-01-01

    The influence of uranium availability upon the future of the fast reactor is reviewed. The important issues considered are uranium reserves and resources, uranium market prices, fast reactor economics and the political availability of uranium to customers in other countries. (U.K.)

  19. Review of fast reactor activities

    Energy Technology Data Exchange (ETDEWEB)

    Balz, W [Commission of the European Communities, Brussels (Belgium)

    1978-07-01

    The Commission of the European Communities continued its activities on the following lines: activities aimed at preparing for commercialization of fast breeder reactors which are essentially performed in the frame of Fast Reactor Coordinating Committee (FRCC); the execution of its own research program in the Joint Research Center. The report covers activities of the FRCC, of the Safety Working Group (SWG), the Whole Core Accident Code (WAC) subgroup, Containment (CONT) subgroup, Codes and Standards Working Group (CSWG). Research and development activities are concerned with LMFBR safety, subassembly thermal hydraulics, fuel-coolant interactions, post-accident heat removal, dynamic load response, safety related material properties, utilization limits of fast breeder fuels, plutonium and actinide aspects of nuclear fuel cycle.

  20. Review of fast reactor activities

    International Nuclear Information System (INIS)

    Balz, W.

    1978-01-01

    The Commission of the European Communities continued its activities on the following lines: activities aimed at preparing for commercialization of fast breeder reactors which are essentially performed in the frame of Fast Reactor Coordinating Committee (FRCC); the execution of its own research program in the Joint Research Center. The report covers activities of the FRCC, of the Safety Working Group (SWG), the Whole Core Accident Code (WAC) subgroup, Containment (CONT) subgroup, Codes and Standards Working Group (CSWG). Research and development activities are concerned with LMFBR safety, subassembly thermal hydraulics, fuel-coolant interactions, post-accident heat removal, dynamic load response, safety related material properties, utilization limits of fast breeder fuels, plutonium and actinide aspects of nuclear fuel cycle

  1. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1981-06-01

    The accuracy requirements and the status of the evaluated fission-product cross sections for fast reactors are reviewed; the work on calculating the sensitivity of the sodium void effect to fission-product cross sections is described; some results of the intercomparison of adjusted data sets for capture cross sections of fission-products (RCN-2A and CARNAVAL-IV) are discussed; the applicability of the maximum-likelihood method for the analysis of resolved resonance parameters for a large class of fission-product nuclides is demonstrated; the neutron cross sections for corrosion product 64 Ni are evaluated. Some results of post-irradiation examination of a loss-of-cooling experiment are given; the progress in testing the equipment and instrumentation for transient-overpower experiments is reported. The proceedings in the thermochemical investigations on uranium compounds with some fission-products are described. The creep behaviour of a heat of DIN 1.4948 parent metal is investigated with respect to the changes in strain with different test temperatures. Sodium smoke aerosols have been produced and analysed with respect to their aerodynamic behaviour and morphology. The two-phase local boiling experiments have been analysed to find criteria for the occurrence of different boiling regimes with the objection to deduce general dryout correlations

  2. Status of fast reactor control rod development in the United Kingdom

    International Nuclear Information System (INIS)

    Kelly, B.T.

    1984-01-01

    The two large fast reactors constructed in the United Kingdom, that is the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR) differed substantially in their control systems. DFR was controlled by variation of the neutron leakage from the core while PFR uses conventional control rods containing neutron absorbing materials. This paper describes the development of the PFR control systems, the progressive design of the control systems for the prototype Civil Fast Reactor (CFR) and the supporting research and development programmes. (author)

  3. Fast reactors and nonproliferation

    International Nuclear Information System (INIS)

    Orlov, V.V.

    1997-01-01

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

  4. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

    A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

  5. Review of fast reactor activities

    International Nuclear Information System (INIS)

    1982-01-01

    A description of some highlights of the activities performed by the Commission of the European Communities in the field of fast reactors is given. They fall into two categories: coordinating and harmonizing activities and research activities. The former are essentially performed in the frame of the Fast Reactor Coordinating Committee (FRCC), the latter in the Commission's Joint Research Center and to some extent under contract in research centers of the Member States

  6. SACRD: a data base for fast reactor safety computer codes, general description

    International Nuclear Information System (INIS)

    Greene, N.M.; Forsberg, V.M.; Raiford, G.B.; Arwood, J.W.; Simpson, D.B.; Flanagan, G.F.

    1979-01-01

    SACRD is a data base of material properties and other handbook data needed in computer codes used for fast reactor safety studies. Data are available in the thermodynamics, heat transfer, fluid mechanics, structural mechanics, aerosol transport, meteorology, neutronics, and dosimetry areas. Tabular, graphical and parameterized data are provided in many cases. A general description of the SACRD system is presented in the report

  7. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

    Parcy, J.P.

    1982-09-01

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

  8. Overview of U.S. Fast Reactor Technology Program

    International Nuclear Information System (INIS)

    Hill, Robert

    2013-01-01

    • Concept development studies guide R&D tasks by evaluating system impact for broad variety of technology options: – Small-scale facilities for R&D on key technology; – No near-term plan for demonstration reactor. • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction): – Advanced Structural Materials; – Advanced Energy Conversion; – Advanced Modeling and Simulation. • Other R&D is conducted to address known technology challenges: – Safety and Licensing; – Fuels Development; – Undersodium Viewing

  9. Flowing and freezing of molten core materials during unprotected loss of flow accidents in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Maschek, W.; Royl, P.

    1988-09-01

    Flowing and freezing of mobile core materials change the fissile material distribution and core-inventory under hypothetical accident conditions and determine the path to permanent shutdown of the neutronic events and the energetic potentials. The report classifies the bondary conditions for such flowing and freezing processes by going through the different situations under which these processes can occur in the scenario of the unprotected loss of flow (ULOF) accident. The classification is based on ULOF-accident simulations for a homogeneous reactor core concept of a 300 MWe LMFBR (e. g. SNR-300), but many boundary conditions are also characteristic for other core designs. A review of the relevant experiments is then made to correlate the available experimental information with these classified boundary conditions and to look at the resulting flowing and freezing processes. Boundary conditions that have been experimentally shown to be important are assigned high priorities. The data are specifically valued in relation to these boundary conditions of high priorities. The review includes the major experimental programs with published results. The discussion shows that the results from most clean condition tests for melt relocations are valuable for a better understanding of basic phenomena and analytical model development, but are not directly applicable to real accident conditions. The database for relevant boundary conditions from the ULOF scenario is limited and largely included in integral sequence tests from which quantitative information for modelling is difficult to obtain. Needs for additional investigations are identified. The suggestions are mainly restricted to investigations of the early phase of fuel removal. They are given with reference to candidate facilities and include relocations in the subassemblies and in the inter-subassembly gaps. Particular emphasis is put on the leading edge properties and possible driving forces to which more attention

  10. Nuclear Burning Wave Modular Fast Reactor Concept

    International Nuclear Information System (INIS)

    Kodochigov, N.G.; Sukharev, Yu.P.

    2014-01-01

    The necessity to provide nuclear power industry, comparable in a scope with power industry based on a traditional fuel, inspired studies of an open-cycle fast reactor aimed at: - solution of the problem of fuel provision by implementing the highest breeding characteristics of new fissile materials of raw isotopes in a fast reactor and applying accumulated fissile isotopes in the same reactor, independently on a spent fuel reprocessing rate in the external fuel cycle; - application of natural or depleted uranium for makeup fuel, which, with no spent fuel reprocessing, forms the most favorable non-proliferation conditions; - application of inherent properties of the core and reactor for safety provision. The present report, based on previously published papers, gives the theoretical backgrounds of the concept of the reactor with a nuclear burning wave, in which an enriched-fuel core (driver) is replaced by a blanket, and basic conditions for nuclear burning wave initiating and keeping are shown. (author)

  11. Overview of the fast reactors fuels program

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

  12. Fast reactor operation in the United States

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  13. Fast reactor collaboration in Europe

    International Nuclear Information System (INIS)

    Smith, G.E.I.

    1987-01-01

    Fast reactors have been developed in several European countries, the United Kingdom, France, Germany and Italy. A suggestion to collaborate on fast reactor research and development resulted in an Intergovernmental Memorandum of Understanding signed in 1984 by the UK, France, Germany, Italy and Belgium. Holland was expected to join later. This provided for co-operation between electric utilities, reactor design, research and development companies and fuel cycle companies. Three steering committees have so far been set up, the European fast reactor utilities Group, the European research and development and the European fuel cycle steering committees. Progress on these is detailed. The main areas of technology exchange are listed in the Appendix. The possibility exists for a series of three large demonstration plants to be built in Europe and a fuel reprocessing plant to confirm the reactor system. (U.K.)

  14. Review of fast reactor activities

    International Nuclear Information System (INIS)

    Haeussermann, W.; Royen, J.

    1978-01-01

    Since 1971, when the Co-ordinating Group on Gas-Cooled Fast reactors Development was set up, the participating countries have maintained an interest in keeping this option as a back-up solution to the sodium cooled fast reactors. Two different concepts were investigated, one based on coated particle type fuel elements and the other on pin type fuel elements. The coated particles studies have been brought to an end, and resources were concentrated on the further development of the pin type concept. The work done in previous years covered design and safety investigations, heat transfer studies and irradiation experiments in thermal reactors

  15. Fast reactor physics - an overview

    International Nuclear Information System (INIS)

    Lee, S.M.

    2004-01-01

    An introduction to the basic features of fast neutron reactors is made, highlighting the differences from the more conventional thermal neutron reactors. A discussion of important feedback reactivity mechanisms is given. Then an overview is presented of the methods of fast reactor physics, which play an important role in the successful design and operation of fast reactors. The methods are based on three main elements, namely (i) nuclear data bases, (ii) numerical methods and computer codes, and (iii) critical experiments. These elements are reviewed and the present status and future trends are summarized. (author)

  16. The fourth research co-ordination meeting (RCM) on 'Updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactors reactivity effects'. Working material

    International Nuclear Information System (INIS)

    2003-01-01

    The fourth Research Co-ordination Meeting (RCM) of the Co-ordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effect' was held during 19-23 May, 2003 in Obninsk, Russian Federation. The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The first RCM took place in Vienna on 24 - 26 November 1999. The meeting was attended by 19 participants from 7 Member States and one from an international organization (France, Germany, India, Japan, Rep. of Korea, Russian Federation, the United Kingdom, and IAEA). The participants from two Member States (China and the U.S.A.) provided their results and presentation materials even though being absent at the meeting. The results for several relevant reactivity parameters obtained by the participants with their own state-of-the-art basic data and codes, were compared in terms of calculational uncertainty, and their effects on the ULOF transient behavior of the hybrid BN- 600 core were evaluated. Contributions of the participants in the benchmark analyses is shown. This report first addresses the benchmark definitions and specifications given for each Phase and briefly introduces the basic data, computer codes, and methodologies applied to the benchmark analyses by various participants. Then, the results obtained by the participants in terms of calculational uncertainty and their effect on the core transient behavior are intercompared. Finally it addresses some conclusions drawn in the benchmarks

  17. Actinides burnup in a sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Pineda A, R.; Martinez C, E.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    The burnup of actinides in a nuclear reactor is been proposed as part of an advanced nuclear fuel cycle, this process would close the fuel cycle recycling some of the radioactive material produced in the open nuclear fuel cycle. These actinides are found in the spent nuclear fuel from nuclear power reactors at the end of their burnup in the reactor. Previous studies of actinides recycling in thermal reactors show that would be possible reduce the amounts of actinides at least in 50% of the recycled amounts. in this work, the amounts of actinides that can be burned in a fast reactor is calculated, very interesting results surge from the calculations, first, the amounts of actinides generated by the fuel is higher than for thermal fuel and the composition of the actinides vector is different as in fuel for thermal reactor the main isotope is the {sup 237}Np in the fuel for fast reactor the main isotope is the {sup 241}Am, finally it is concluded that the fast reactor, also generates important amounts of waste. (Author)

  18. Improved algorithms for the calculation of resolved resonance cross sections with applications to the structural Doppler effect in fast reactors

    International Nuclear Information System (INIS)

    Hwang, R.N.; Toppel, B.J.; Henryson, H. II.

    1980-10-01

    Motivated by a need for an economical yet rigorous tool which can address the computation of the structural material Doppler effect, an extremely efficient improved RABANL capability has been developed utilizing the fact that the Doppler broadened line shape functions become essentially identical to the natural line shape functions or Lorentzian limits beyond about 100 Doppler widths from the resonance energy, or when the natural width exceeds about 200 Doppler widths. The computational efficiency has been further enhanced by preprocessing or screening a significant number of selected resonances during library preparation into composition and temperature independent smooth background cross sections. The resonances which are suitable for such pre-processing are those which are either very broad or those which are very weak. The former contribute very little to the Doppler effect and their self-shielding effect can readily be averaged into slowly varying background cross section data, while the latter contribute very little to either the Doppler or to self-shielding effects. To illustrate the accuracy and efficiency of the improved RABANL algorithms and resonance screening techniques, calculations have been performed for two systems, the first with a composition typical of the STF converter region and the second typical of an LMFBR core composition. Excellent agreement has been found for RABANL compared to the reference Monte Carlo solution obtained using the code VIM, and improved results have also been obtained for the narrow resonance approximation in the ultra-fine-group option of MC 2 -2

  19. Technical Meeting on Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels. Presentations

    International Nuclear Information System (INIS)

    2011-01-01

    The purpose of this meeting was to enable a rationalization and advancement of the design and manufacturing processes, a better selection of promising fuels, and a reduction of the time and costs currently required for R and D and testing, as well as to contribute to the improvement of the safety features of fuels under all operational states and accidental conditions. An overview of the status and perspective of the design, manufacturing and irradiation behaviour of fast reactors fuels were provided during this meeting. The main objectives are the following: Ensure sharing and dissemination of knowledge and expertise; Discuss specific features and issues of existing fuels; Improve knowledge and data for the design and engineering of fast reactor fuel and core structural materials; Discuss perspectives on advanced fuels; Consider modern technological, design and testing tools enabling reliable performance of fuels in current and planned operational environments; Establish international consensus in the developmental efforts on advanced fast reactor technologies, including collaborative programs and experiments. Contribute to the preparation and outline of the planned IAEA Coordinated Research Project on 'Examination of advanced fast reactor fuel and core structural materials. Each of the 24 presentations made at the meeting have been indexed separately

  20. The Integral Fast Reactor concept

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  1. Fast reactor research in Switzerland

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  2. Design, Manufacturing and Irradiation Behaviour of Fast Reactor Fuel. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2013-04-01

    Fast reactors are vital for ensuring the sustainability of nuclear energy in the long term. They offer vastly more efficient use of uranium resources and the ability to burn actinides, which are otherwise the long-lived component of high level nuclear waste. These reactors require development, qualification, testing and deployment of improved and innovative nuclear fuel and structural materials having very high radiation resistance, corrosion/erosion and other key operational properties. Several IAEA Member States have made efforts to advance the design and manufacture of technologies of fast reactor fuels, as well as to investigate their irradiation behaviour. Due to the acute shortage of fast neutron testing and post-irradiation examination facilities and the insufficient understanding of high dose radiation effects, there is a need for international exchange of knowledge and experience, generation of currently missing basic data, identification of relevant mechanisms of materials degradation and development of appropriate models. Considering the important role of nuclear fuels in fast reactor operation, the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) proposed a Technical Meeting (TM) on 'Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels', which was hosted by the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russian Federation, from 30 May to 3 June 2011. The TM included a technical visit to the fuel production plant MSZ in Elektrostal. The purpose of the meeting was to provide a forum to share knowledge, practical experience and information on the improvement and innovation of fuels for fast reactors through scientific presentations and brainstorming discussions. The meeting brought together 34 specialists from national nuclear agencies, R and D and design institutes, fuel vendors and utilities from 10 countries. The presentations were structured into four sections: R and D Programmes on FR Fuel

  3. 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.

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

    International Nuclear Information System (INIS)

    Ludewig, H.; Powers, D.A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.

    2012-01-01

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

  5. Criteria for the selection of PEC primary circuit structural material

    International Nuclear Information System (INIS)

    Antoni, R.; Brunori, G.; Maesa, S.; Scibona, G.; Tomassetti, G.

    1977-01-01

    The choice of the structural materials is generally a compromise between the project requirements, the characteristics (mechanical and environmental) of the materials and the available technology to construct the various parts of the components. The criteria of selection of structural materials for the primary circuit of fast reactor are reported. The criteria concern both general and utilization aspects

  6. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

  7. Status of national programmes on fast reactors

    International Nuclear Information System (INIS)

    1994-04-01

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

  8. Slow clean-up for fast reactor

    Science.gov (United States)

    Banks, Michael

    2008-05-01

    The year 2300 is so distant that one may be forgiven for thinking of it only in terms of science fiction. But this is the year that workers at the Dounreay power station in Northern Scotland - the UK's only centre for research into "fast" nuclear reactors - term as the "end point" by which time the site will be completely clear of radioactive material. More than 180 facilities - including the iconic dome that housed the Dounreay Fast Reactor (DFR) - were built at at the site since it opened in 1959, with almost 50 having been used to handle radioactive material.

  9. The International conference on fast reactors and related fuel cycles: next generation nuclear systems for sustainable development. Book of abstracts

    International Nuclear Information System (INIS)

    2017-01-01

    The materials of the International Conference on Fast Reactors and Related Fuel Cycles (June 26-29, 2017, Yekaterinburg) are presented. The forum was organized by the IAEA with the assistance of Rosatom State Corporation. The theme of the conference: “The New Generation of Nuclear Systems for Sustainable Development”. About 700 specialists from more than 30 countries took part in the conference. The state and prospects for the development of the direction of fast reactors in countries dealing with this topic were discussed. A wide range of scientific issues covered the concepts of prospective reactors, reactor cores, fuel and fuel cycles, operation and decommissioning, safety, licensing, structural materials, industrial implementation [ru

  10. Stationary Liquid Fuel Fast Reactor

    International Nuclear Information System (INIS)

    Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

    2015-01-01

    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

  11. 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

  12. Introduction of the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Matsuba, Ken-ichi; Kawahara, Hirotaka; Aoyama, Takafumi

    2006-01-01

    The experimental fast reactor JOYO at O-arai Engineering Center of Japan Nuclear Cycle Development Institute is the first liquid metal cooled fast reactor in Japan. This paper describes the plant outline, experiences on the fast reactor technology and test results accumulated through twenty eight years successful operation of JOYO. (author)

  13. Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design

    Energy Technology Data Exchange (ETDEWEB)

    Smith, C

    2010-02-22

    The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

  14. Fast Reactor Knowledge Management at IGCAR, India

    International Nuclear Information System (INIS)

    Kuriakose, K.K.

    2013-01-01

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

  15. Fast reactors: the industrial perspective

    International Nuclear Information System (INIS)

    Vaughan, R.D.

    1986-01-01

    Industrial participation in the development of the fast reactor is reviewed, from the construction of PFR at Dounreay to the initial steps towards collaboration in Europe. The optimum design of the fast reactor has changed considerably from the days when it was needed urgently to forestall a shortage of uranium to today when uranium is abundant and cheap. The evolution of the reactor design over this period is described. Collaboration in Europe is shown to be the only answer to high development costs and the search for a reactor which will compete with thermal reactors in today's environment. The partner countries in this collaboration are all motivated differently, and this is leading to some delays in concluding the necessary agreements. The objective on the industrial front is now to participate in the two or three demonstration fast reactors that will be built in Europe during the remainder of the century leading, it is hoped, to a competitive reactor design by the year 2000. (author)

  16. Fast Reactors and Nuclear Nonproliferation

    International Nuclear Information System (INIS)

    Avrorina, E.N.; Chebeskovb, A.N.

    2013-01-01

    Conclusion remarks: 1. Fast reactor start-up with U-Pu fuel: – dependent on thermal reactors, – no needs in U enrichment, – needs in SNF reprocessing, – Pu is a little suitable for NED, – practically impossible gun-type NED, – difficulties for implosion-type NED: necessary tests, advanced technologies, etc. – Pu in blankets is similar to WPu by isotopic composition, – Use of blanket for production isotopes (e.g. 233 U), – Combined reprocessing of SNF: altogether blanket and core, – Blanket elimination: decrease in Pu production – No pure Pu separation. 2. Fast reactor start-up with U fuel: - Needs in both U enrichment and SNF reprocessing, - Independent of thermal reactors, - Good Pu bred in the core let alone blankets, - NED of simple gun-type design, - Increase of needs in SWU, - Increased demands in U supply. 3. Fast reactors for export: - Uranium shortage, - To replace thermal reactors in future, - No blankets (depends on the country, though), - Fuel supply and SNF take back, - International centers for rendering services of NFC. Time has come to remove from FRs and their NFC the label unfairly identifying them as the most dangerous installations of nuclear power from the standpoint of being a proliferation problem

  17. Present status and future perspective of R and D on lead heavy metal-cooled fast reactors

    International Nuclear Information System (INIS)

    Takahashi, Minoru

    2007-01-01

    Since a lead heavy metal (lead-bismuth eutectic) is chemically inert and has higher boiling point compared to a sodium, a lead heavy metal-cooled fast reactor can be inherently safe and has good nuclear characteristics and is so suitable to a medium-small size of the reactor. R and D on corrosion of a lead heavy metal has been carried out in the world and this issue might be solved to choose specific corrosion resistant alloys for structural materials and fuel cans of a lead heavy metal-cooled reactor. This article reviews present status and future perspective on lead heavy metal-cooled fast reactors. (T. Tanaka)

  18. Gas cooled fast reactor research in Europe

    International Nuclear Information System (INIS)

    Stainsby, Richard; Peers, Karen; Mitchell, Colin; Poette, Christian; Mikityuk, Konstantin; Somers, Joe

    2011-01-01

    Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce. GFR research within Europe is performed directly by those states who have signed the 'System Arrangement' document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1983-01-01

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

  20. The integral fast reactor concept

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  1. Dosimetry methods for fuels, cladding and structural materials

    International Nuclear Information System (INIS)

    Roettger, H.

    1980-01-01

    This volume of the proceedings of the symposium on reactor dosimetry covers the following topics: the metallurgy and dosimetry interface, radiation damage correlations of structural materials and damage analyses techniques, dosimetry for fusion materials, light water reactor pressure vessel surveillance in practice and irradiation experiments, fast reactor and reseach reactor characterization

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

    International Nuclear Information System (INIS)

    Hoekstra, E.K.

    1970-02-01

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

  3. Fast reactor core monitoring device

    International Nuclear Information System (INIS)

    Sanda, Toshio; Inoue, Kotaro; Azekura, Kazuo.

    1982-01-01

    Purpose: To enable the rapid and accurate on-line identification of the state of a fast reactor core by effectively utilizing the measured data on the temperature and flow rate of the coolant. Constitution: The spacial power distribution and average assembly power are quickly calculated using an approximate calculating method, the measured values and the calculated values of the inlet and outlet temperature difference, flow rate and coolant physical values of an assembly are combined and are individually obtained, the most definite respective values and their errors are obtained by a least square method utilizing a formula of the relation between these values, and the power distribution and the temperature distribution of a reactor core are estimated in this manner. Accordingly, even when the measuring accuracy and the calculating accuracy are equal as in a fast reactor, the power distribution and the temperature distribution can be accurately estimated on-line at a high speed in a nuclear reactor, information required for the operator is provided, and the reactor can thus be safely and efficiently operated. (Yoshihara, H.)

  4. The 'SURA' fast reactor program

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The Commissariat a l'Energie Atomique's SURA program on fast reactor safety consists of two specific testing programs on fastbreeder reactor safety: the Cabri and Scarabee programs. Both Cabri and Scarabee are examples of multinational research collaboration. The CEA and the Karlsruhe Nuclear Research Center are each covering half of the construction costs. Britain, the US and Japan are also due to participate in these experiments. The aim of the programs is to examine the behaviour of fuel in sodium cooled fast reactors. The Cabri program consists of setting off a reactivity accident in a power reactor core which is cooled with liquid sodium, such an accident occurring after a sharp increase in reactivity or as a result of the pump suddenly breaking down without there at the same time being any fall in the control rods. In 1967 the Commissariat a l'Energie Atomique started its Scarabee research program which is trying to analyse the sort of things that can go wrong with fuel cooling systems and what the consequences can be [fr

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

    International Nuclear Information System (INIS)

    Vilim, R.B.; Park, Y.S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-01-01

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

  6. 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.

  7. International Experience with Fast Reactor Operation & Testing

    International Nuclear Information System (INIS)

    Sackett, John I.; Grandy, C.

    2013-01-01

    Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

  8. Influence of high dose irradiation on core structural and fuel materials in advanced reactors

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA International Working Group on Fast Reactors (IWGFR) periodically organizes meeting to discuss and review important aspects of fast reactor technology. The fifth meeting held in Obninsk, Russian Federation, 16-19 June 1997, was devoted to the influence of high dose irradiation on the mechanical properties of reactor core structural and fuel materials. The proceedings includes the papers submitted at this meeting each with a separate abstract

  9. U.S. Status of Fast Reactor Research and Technology

    International Nuclear Information System (INIS)

    Hill, Robert

    2012-01-01

    Summary: • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction) and safety: 1. System Integration and Concept Development; 2. Safety Technology; 3. Advanced Materials; 4. Ultrasonic Viewing; 5. Advanced Energy Conversion (Supercritical CO 2 Brayton cycle); 6. Reactor Simulation; 7. Nuclear Data; 8. Advanced Fuels. • Fast reactors have flexible capability for actinide management: – A wide variety of fuel cycle options are being considered; • International R&D collaboration pursued in Generation-IV and multilateral arrangements

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

    International Nuclear Information System (INIS)

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

    1986-05-01

    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 10 15 n/cm 2 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) [pt

  11. Integral fast reactor safety features

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  12. Holography for fast reactor inspection

    International Nuclear Information System (INIS)

    Tozer, B.A.

    1980-01-01

    Holography, an optical process whereby an image of the original subject can be reconstructed in three dimensions, is being developed for use as an optical inspection tool. With a potential information storage density of 10 16 bits/m 2 , the ability to reconstruct in 3 dimensions, a depth of field of up to 8 metres, extremely wide angle of view, and potentially diffraction limited resolution, holography should be invaluable for the optical recording of fast reactors during construction, and the inspection of optically accessible regions during operation, or maintenance down-times. The photographic emulsions used for high resolution holography are fine-grained and fog only very slowly when subjected to γ-radiation, so that inspection of highly radio-active regions and components can be effected satisfactorily. Some of the practical limitations affecting holography are described and ways of overcoming them discussed. Some preliminary results are presented. (author)

  13. Integral fast reactor safety features

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  14. Interfacial effects in fast reactors

    International Nuclear Information System (INIS)

    Saidi, M.S.; Driscoll, M.J.

    1979-05-01

    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

  15. Radiation control and safety of fast reactor; Radijaciona kontrola i sigurnost postrojenja sa brzim reaktorom

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M; Antic, D [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1983-07-01

    The fundamental activities for safeguard of radiation control and safety and the necessary staff for them for fast reactor plant are shown. The basic systems for the plant radiation control are counted, especially with regards to poisoning of some fuel materials. The particular characteristics of the plant radiation control determined by the fast reactor are pointed out. (author)

  16. The UK commercial demonstration fast reactor design

    International Nuclear Information System (INIS)

    Holmes, J.A.G.

    1987-01-01

    The paper on the UK Commercial Demonstration Fast Reactor design was presented to the seminar on 'European Commercial Fast Reactor Programme, London 1987. The design is discussed under the topic headings:- primary circuit, intermediate heat exchangers and pumps, fuel and core, refuelling, steam generators, and nuclear island layout. (U.K.)

  17. Aspects of the fast reactors fuel cycle

    International Nuclear Information System (INIS)

    Zouain, D.M.

    1982-06-01

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

  18. Advances by the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  19. Economic Issues of Fast Reactor in China

    International Nuclear Information System (INIS)

    Yang Hongyi

    2013-01-01

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

  20. Proceedings 21. International Conference on Applied Physics of Condensed Matter and of the Scientific Conference Advanced Fast Reactors

    International Nuclear Information System (INIS)

    Vajda, J.; Jamnicky, I.

    2015-01-01

    The 21. International Conference on Applied Physics of Condensed Matter was held on 24-26 June, 2015 on Strbske Pleso, Strba, Slovakia. The Scientific Conference the Advanced Fast Reactors was part of the 21 st International Conference on APCOM 2015. The specialists discussed various aspects of modern problems in: Physical properties and structural aspects of solid materials and their influencing; Advanced fast reactors; Physical properties and structural aspects of solid materials and their influencing; Nuclear science and technology, influence of irradiation on physical properties of materials, radiation detection; Computational physics and theory of physical properties of matter; interdisciplinary physics of condensed matter; Nuclear science and technology, influence of irradiation on physical properties of materials, radiation detection; Optical phenomena in materials, photovoltaics and photonics, new principles in sensors and detection methods. Fifty seven contributions relevant of INIS interest has been inputted to INIS.

  1. The fifth research coordination meeting (RCM) on 'Updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactors reactivity effects'. Working material

    International Nuclear Information System (INIS)

    2004-01-01

    The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The objectives of the fifth RCM were: to review the progress achieved since the 4th RCM; to review and finalize the draft synthesis report on BN-600 MOX Fueled Core Benchmark Analysis (Phase 4); to compare the results of Phase 5 (BFS Benchmark Analysis); to agree on the work scope of Phase 6 (BN-Full MOX Minor Actinide Core Benchmark); to discuss the preparation of the final report. In this context, review and related discussions were made on the following items: summary review of Actions and results since the 4th RCM; finalization of the draft synthesis report on BN-600 full MOX-fueled core benchmark analysis (Phase 4); presentation of individual results for Phase 5 by Member States; preliminary inter-comparison analysis of the results for Phase 5; definition of the benchmark model and work scope to be performed for Phase 6; details of the work scope and future CRP timetable for preparing a final report

  2. Reflector Performance Study in Ultra-long Cycle Fast Reactor

    International Nuclear Information System (INIS)

    Tak, Taewoo; Kong, Chidong; Choe, Jiwon; Lee, Deokjung

    2013-01-01

    There are reflector assemblies outside the fuel region, surrounding the fuel assemblies and axial reflector is located at the bottom of the core to control the neutron leakage fraction which is an important factor in fast reactor system. HT-9 was used as a reflector material as well as a structure material. In this study, alternative reflector materials were proposed and their reflection performance was tested and studied focused on its physics. ODS-MA957 and SiC were chosen from iron based alloy and ceramic respectively. The two materials were tested and compared with HT-9 in UCFR-1000 as a radial and an axial reflector and it was evaluated from the neutronics point of view with comparing the core life and the coolant void reactivity. The calculation and evaluation were performed by McCARD Monte Carlo code. The reflector materials for UCFR-1000 have been investigated in the aspect of neutronics. The reflection effect shows different performance corresponding to reflector material used. Also, the neutron energy spectrum is affected by changing materials which causes spectrum softening but it is not enough to influence the core life. With more reflector material candidates such as lead-based liquid metal, reflection performance and core parameter study will be investigated for next step

  3. Intrinsically secure fast reactors with dense cores

    International Nuclear Information System (INIS)

    Slessarev, Igor

    2007-01-01

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

  4. Fast reactor fuel reprocessing. An Indian perspective

    International Nuclear Information System (INIS)

    Natarajan, R.; Raj, Baldev

    2005-01-01

    The Department of Atomic Energy (DAE) envisioned the introduction of Plutonium fuelled fast reactors as the intermediate stage, between Pressurized Heavy Water Reactors and Thorium-Uranium-233 based reactors for the Indian Nuclear Power Programme. This necessitated the closing of the fast reactor fuel cycle with Plutonium rich fuel. Aiming to develop a Fast Reactor Fuel Reprocessing (FRFR) technology with low out of pile inventory, the DAE, with over four decades of operating experience in Thermal Reactor Fuel Reprocessing (TRFR), had set up at the India Gandhi Center for Atomic Research (IGCAR), Kalpakkam, R and D facilities for fast reactor fuel reprocessing. After two decades of R and D in all the facets, a Pilot Plant for demonstrating FRFR had been set up for reprocessing the FBTR (Fast Breeder Test Reactor) spent mixed carbide fuel. Recently in this plant, mixed carbide fuel with 100 GWd/t burnup fuel with short cooling period had been successfully reprocessed for the first time in the world. All the challenging problems encountered had been successfully overcome. This experience helped in fine tuning the designs of various equipments and processes for the future plants which are under construction and design, namely, the DFRP (Demonstration Fast reactor fuel Reprocessing Plant) and the FRP (Fast reactor fuel Reprocessing Plant). In this paper, a comprehensive review of the experiences in reprocessing the fast reactor fuel of different burnup is presented. Also a brief account of the various developmental activities and strategies for the DFRP and FRP are given. (author)

  5. Irradiation Experiments on Plutonium Fuels for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Frost, B. R.T.; Wait, E. [Atomic Energy Research Establishment Harwell, Berks. (United Kingdom)

    1967-09-15

    An assessment carried out some years ago indicated that cermet fuels might provide the high burn-up and integrity required for fast reactors. An irradiation programme was started at Harwell on (U, Pu)O{sub 2} -SS cermet plates and rods, mainly In thermal neutron fluxes, to gain experience of dimensional stability at temperatures typical of modern sodium-cooled fast reactor designs (600-650 Degree-Sign C). A subsequent assessment showed that cermets carried a large penalty as far as breeding was concerned and (U, Pu)C was chosen by Harwell for long-term study as an alternative, economic, fast reactor fuel. However, the results from the cermet experiments were of sufficient promise to proceed with parallel irradiation programmes on cermets and carbide. The studies of cermets showed that dimensional instability (swelling and cladding rupture) were caused by the pressures exerted on the steel matrix by the fuel particles, and that the initial density of the fuel particles was important in determining the burn-up at which failure occurred. Further, it was shown that cermets provided a useful vehicle for studying the changes occurring in oxide fuel particles with increasing burn-up. The disappearance of initial porosity and its replacement by fission gas bubbles and segregated solid fission products was studied in some detaiL No significant differences were observed between UO{sub 2} and(U,Pu)O{sub 2} particles. The initial studies of (U, Pu)C were concerned with the effect of varying composition and structure on swelling and fission gas release. A tantalum-lined nickel alloy cladding material was used to contain both pellet and powder specimens In an irradiation experiment in the core of the Dounreay fast reactor. This showed that the presence of a metal phase in the fuel led to a high swelling rate, that fission gas release was low up to {approx} 3% bum-up, and that a low density powder accommodated the swelling without excessive straining of the can. A subsequent

  6. Coolant-fuel interaction in Sodium-cooled Fast Reactors: Structural investigations of The Na-An-O (An = U, Np, Pu) systems

    International Nuclear Information System (INIS)

    Smith, A.L.; Raison, P.E.; Bykov, D.M.; Konings, R.J.; Caciuffo, R.; Cheetham, A.K.

    2014-01-01

    Nuclear energy has the potential to provide Europe with a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gases emissions. The interest for Sodium-cooled-Fast-spectrum Reactors (SFRs), when compared to Pressurized Water Reactors (PWRs), lies in their more efficient management of plutonium and other actinides as well as their ability to use almost all of the energy in the natural uranium versus 1% utilized in thermal spectrum systems. The high fuel efficiency of fast reactors could greatly dampen concerns about fuel supply. But these reactors have also several drawbacks when compared to PWRs (i.e sodium fire, Na reaction with O2 and H2O, interaction of sodium with oxide fuels). Their development at an industrial scale needs therefore an exhaustive safety assessment that comprises both experimental work and development of sophisticated modelling tools able to describe the reactor behaviour in normal or incidental conditions

  7. Fast reactor fuel reprocessing in the UK

    International Nuclear Information System (INIS)

    Allardice, R.H.; Williams, J.; Buck, C.

    1977-01-01

    Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the U.K. since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium based fast reactor system and the importance of establishing at an early stage fast reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high burn-up thermal reactor oxide fuel. In consequence, the U.K. has decided to reprocess irradiated fuel from the 250 MW(E) Prototype Fast Reactor as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small scale fully active demonstration plant have been carried out over the past 5 years and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant a parallel development programme has been initiated to provide the basis for the design of a large scale fast reactor fuel reprocessing plant to come into operation in the late 1980s to support the projected U.K. fast reactor installation programme. The paper identifies the important differences between fast reactor and thermal reactor fuel reprocessing technologies and describes some of the development work carried out in these areas for the small scale P.F.R. fuel reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast reactor fuel reprocessing plant is outlined and the current design philosophy is discussed

  8. Some questions and answers concerning fast reactors

    International Nuclear Information System (INIS)

    Marshall, W.

    1980-01-01

    The theme of the lecture is the place of the fast reactor in an evolving nuclear programme. The whole question of plutonium is first considered, ie its method of production and the ways in which it can be used in the fast reactor fuel cycle. Whether fast reactors are necessary is then discussed. Their safety is examined with particular attention to those design features which are most criticised ie high volumetric power density of the core, and the use of liquid sodium as coolant. Attention is then paid to environmental and safeguard aspects. (U.K.)

  9. Euratom contributions in Fast Reactor research programmes

    International Nuclear Information System (INIS)

    Fanghänel, Th.; Somers, J.

    2013-01-01

    The Sustainable Nuclear Initiative: • demonstrate long-term sustainability of nuclear energy; • demonstration reactors of Gen IV: •more efficient use of resources; • closed fuel cycle; • reduced proliferation risks; • enhanced safety features. • Systems pursued in Europe: • Sodium-cooled fast reactor SFR; • Lead-cooled fast reactor LFR; • Gas-cooled fast reactor GFR. Sustainable Nuclear Energy Technology Platform SNE-TP promotes research, development and demonstration of the nuclear fission technologies necessary to achieve the SET-Plan goals

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1980-01-01

    Commercial Fast Reactor (CFR) have concentrated on reducing the capital cost and improving the intrinsic safety of the reactor. Particular attention is being paid to achieving the necessary guarantees of structural reliability for those internal structures which are difficult to inspect and good progress has been made in this area, particularly in providing redundancy in the key reactor structures. International collaboration continues to play a large part in the UK programme and there is an active exchange between the UK and all those countries with operating fast reactors. A particular step forward was the signing of an agreement between the AEA and the USDOE for a programme of work on fuel testing, which is aimed to provide important confirmatory work on the safety of fuel in fast reactors. This programme envisages the destructive testing in the US safety test reactor TREAT at Idaho of fast reactor pins of both UK and US origin which have been irradiated in PFR and then shipped across the Atlantic for testing

  11. The fast reactor and energy supply

    International Nuclear Information System (INIS)

    1979-01-01

    The progress made with fast reactor development in many countries is summarised showing that the aim is to provide to the nation concerned an ability to instal fast reactor power stations at the end of this century or early in the next one. Accepting the importance of fast reactors as a potential independent source of energy, problems concerning economics, industrial capability, technical factors, public acceptibility and in particular plutonium management, are discussed. It is concluded that although fast reactors have reached a comparatively advanced stage of development, a number of factors make it likely that their introduction for electricity generation will be a gradual process. Nevertheless it is necessary to complete demonstration and development phases in good time. (U.K.)

  12. Economic evaluation of fast reactor fuel cycling

    International Nuclear Information System (INIS)

    Hu Ping; Zhao Fuyu; Yan Zhou; Li Chong

    2012-01-01

    Economic calculation and analysis of two kinds of nuclear fuel cycle are conducted by check off method, based on the nuclear fuel cycling process and model for fast reactor power plant, and comparison is carried out for the economy of fast reactor fuel cycle and PWR once-through fuel cycle. Calculated based on the current price level, the economy of PWR one-through fuel cycle is better than that of the fast reactor fuel cycle. However, in the long term considering the rising of the natural uranium's price and the development of the post treatment technology for nuclear fuels, the cost of the fast reactor fuel cycle is expected to match or lower than that of the PWR once-through fuel cycle. (authors)

  13. The energy gap and the fast reactor

    International Nuclear Information System (INIS)

    Hill, J.

    1977-01-01

    The background to the development of fast reactors is summarized. In Britain, the results of the many experiments performed, the operation of the Dounreay Fast Reactor for the past 18 years and the first year's operation of the larger Prototype Fast Reactor have all been very encouraging, in that they demonstrated that the performance corresponded well with predictions, breeding is possible, and the system is exceptionally stable in operation. The next step in fast reactor engineering is to build a full-scale fast reactor power station. There would seem to be little reason to expect more trouble than could reasonably be expected in constructing any large project of this general nature. However, from an engineering point of view continuity of experience is required. If a decision to build a commercial fast reactor were taken today there would be a 14-year gap between strating this and the start of the Prototype Fast Reactor. This is already much too long. From an environmental standpoint we have to demonstrate that we can manufacture and reprocess fast reacctor fuel for a substantial programme in a way that does not lead to pollution of the environment, and that plutonium-containing fuel can be transported in the quantities required in safety and in a way that does not attract terrorists or require a private army to ensure its security. Finally, we have to find a way to allow many countries to obtain the energy they need from fast reactors, without leading to the proliferation of nuclear weapons or weapons capability. (author)

  14. Decommissioning of fast reactors after sodium draining

    International Nuclear Information System (INIS)

    2009-11-01

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

  15. Fast reactors - Dounreay and the future

    International Nuclear Information System (INIS)

    Jordan, G.

    1988-01-01

    In 1960 at Dounreay, the Dounreay Fast Reactor (DFR) supplied the world's first fast reactor grid electricity, and went on to a highly successful career as a test facility, as fuel designs advanced. In the 1960s, the Prototype Fast Reactor (PFR) was designed and built, beginning operation in 1974. The PFR was built to provide a sound technical and experienced base to support the UK's future Fast Reactor development and design. The in-vessel fuel handling facilities have demonstrated the flexibility of the pool design and a considerable body of in-core fuel handling experience is available. A key issue for further Fast Reactor application is the performance of fuel and, because PFR was designed to take full-scale fuel assemblies, the fuel performance experience is directly relevant to commercial designs. The original PFR design irradiation target of 60000 MWd/t U (equivalent to 7.5 % burn-up) has already been exceeded by a factor of more than two and a 15.9 % burn-up sub-assembly has been discharged and reprocessed without difficulty. Soon a 20 % sub-assembly will follow. Also the PFR reprocessing plant has demonstrated the safety and efficiency of this essential adjunct to Fast Reactor operation. The safety and the environmental protection features of both the PFR and its fuel reprocessing plant have been demonstrated over the last 14 years. 2 refs., 3 figs

  16. Advanced Safeguards Approaches for New Fast Reactors

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  17. 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.

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

    International Nuclear Information System (INIS)

    2011-01-01

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

  19. SACRD: a data base for fast reactor safety computer codes, contents and glossary of Version 1 of the system

    International Nuclear Information System (INIS)

    Greene, N.M.; Forsberg, V.M.; Raiford, G.B.; Arwood, J.W.; Flanagan, G.F.

    1979-01-01

    SACRD is a data base of material properties and other handbook data needed in computer codes used for fast reactor safety studies. This document lists the contents of Version 1 and also serves as a glossary of terminology used in the data base. Data are available in the thermodynamics, heat transfer, fluid mechanics, structural mechanics, aerosol transport, meteorology, neutronics and dosimetry areas. Tabular, graphical and parameterized data are provided in many cases

  20. Waste management in IFR [Integral Fast Reactor] fuel cycle

    International Nuclear Information System (INIS)

    Johnson, T.R.; Battles, J.E.

    1991-01-01

    The fuel cycle of the Integral Fast Reactor (IFR) has important potential advantage for the management of high-level wastes. This sodium-cooled, fast reactor will use metal fuels that are reprocessed by pyrochemical methods to recover uranium, plutonium, and the minor actinides from spent core and blanket fuel. More than 99% of all transuranic (TRU) elements will be recovered and returned to the reactor, where they are efficiently burned. The pyrochemical processes being developed to treat the high-level process wastes are capable of producing waste forms with low TRU contents, which should be easier to dispose of. However, the IFR waste forms present new licensing issues because they will contain chloride salts and metal alloys rather than glass or ceramic. These fuel processing and waste treatment methods can also handle TRU-rich materials recovered from light-water reactors and offer the possibility of efficiently and productively consuming these fuel materials in future power reactors

  1. Non-linear programming method in optimization of fast reactors

    International Nuclear Information System (INIS)

    Pavelesku, M.; Dumitresku, Kh.; Adam, S.

    1975-01-01

    Application of the non-linear programming methods on optimization of nuclear materials distribution in fast reactor is discussed. The programming task composition is made on the basis of the reactor calculation dependent on the fuel distribution strategy. As an illustration of this method application the solution of simple example is given. Solution of the non-linear program is done on the basis of the numerical method SUMT. (I.T.)

  2. The zero-power basis of fast reactor dosimetry

    International Nuclear Information System (INIS)

    Sanders, J.E.

    1978-06-01

    Predictions of reaction rates, atomic displacements, and gamma-ray energy deposition in the Prototype Fast Reactor are based on cross-section data and calculation methods validated against the results of zero-power experiments. The paper reviews work in Zebra relevant to this dosimetry, including neutron spectrometry, power mapping, foil activations within core heterogeneities, and measurements with thermoluminescent detectors. Comparisons of experiment and calculation are discussed in relation to the accuracies required to meet materials testing objectives. (author)

  3. The zero-power basis of fast reactor dosimetry

    International Nuclear Information System (INIS)

    Sanders, J.E.

    1978-06-01

    Predictions of reaction rates, atomic displacements, and gamma-ray energy deposition in the Prototype Fast Reactor are based on cross-section data and calculation methods validated against the results of zero-power experiments. The paper reviews work in Zebra relevant to this dosimetry, including neutron spectrometry, power mapping, foil activations within core heterogeneities, and measurements with thermoluminescent detectors. Comparisons of experiment and calculation are discussed in relation to the accuracies required to meet material testing objectives. (author)

  4. 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)

  5. New stainless steels of ferrite-martensite grade and perspectives of their application in thermonuclear facilities and fast reactors

    International Nuclear Information System (INIS)

    Ajtkhozhin, Eh.S.; Maksimkin, O.P.

    2007-01-01

    Review of scientific literature for last 5 years in which results on study of radiation effect on ferrite-martensite steels - construction materials of fast reactors and most probable candidates for first wall and blanket of the thermonuclear facilities ITER and Demo - are presented. Alongside with this a prior experimental data on study of microstructure changing and physical- mechanical properties of ferrite-martensite steel EhP-450 - the material of hexahedral case of spent assembly of BN-350 fast reactor- are cited. Principal attention was paid to considering of radiation effects of structural components content changing and ferrite-martensite steel swelling irradiated at comparatively low values of radiation damage climb rate

  6. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  7. Development of physical conceptions of fast reactors

    International Nuclear Information System (INIS)

    Khomyakov, Yu.S.; Matveev, V.I.; Moiseev, A.V.

    2013-01-01

    • Russian experience in developing fast reactors has proved clearly scientific justification of conceptual physical principles and their technical feasibility. • However, the potential of fast reactors caused by their physical features has not been fully realized. • In order to assure the real possibility of transition to the nuclear power with fast reactors by about 2030 it is necessary to consistently update fast reactor designs for solving the following key problems: - increasing of self-protection level of reactor core; - improvement of technical and economical characteristics; - solution of the problems related to the fuel supply of nuclear power and assimilation of closed nuclear fuel cycle; - disposal of long lived radioactive waste and transmutation of minor actinides. • Russian program (2010-2020) on the development of basic concepts of the new generation reactors implies successive solution of the above problems. • New technical decisions will be demonstrated by development and assimilation of the new reactors: - BN-800 – development of the fuel cycle infrastructure and mastering of the new types of fuel; - BN-1200 reactor – demonstration economical efficiency of fast reactor and new level of safety; - BREST development and demonstration new heavy liquid metal coolant technology and alternative design concept

  8. Corrosion of structural materials for Generation IV systems

    International Nuclear Information System (INIS)

    Balbaud-Celerier, F.; Cabet, C.; Courouau, J.L.; Martinelli, L.; Arnoux, P.

    2009-01-01

    The Generation IV International Forum aims at developing future generation nuclear energy systems. Six systems have been selected for further consideration: sodium-cooled fast reactor (SFR), gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR) and very high temperature reactor (VHTR). CEA, in the frame of a national program, of EC projects and of the GIF, contributes to the structural materials developments and research programs. Particularly, corrosion studies are being performed in the complex environments of the GEN IV systems. As a matter of fact, structural materials encounter very severe conditions regarding corrosion concerns: high temperatures and possibly aggressive chemical environments. Therefore, the multiple environments considered require also a large diversity of materials. On the other hand, the similar levels of working temperatures as well as neutron spectrum imply also similar families of materials for the various systems. In this paper, status of the research performed in CEA on the corrosion behavior of the structural material in the different environments is presented. The materials studied are either metallic materials as austenitic (or Y, La, Ce doped) and ferrito-martensitic steels, Ni base alloys, ODS steels, or ceramics and composites. In all the environments studied, the scientific approach is identical, the objective being in all cases the understanding of the corrosion processes to establish recommendations on the chemistry control of the coolant and to predict the long term behavior of the materials by the development of corrosion models. (author)

  9. European Union: Review of fast reactor related activities

    International Nuclear Information System (INIS)

    Goethem, G. van; Hugon, M.

    1998-01-01

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

  10. Fast reactor system factors affecting reprocessing plant design

    International Nuclear Information System (INIS)

    Allardice, R.H.; Pugh, O.

    1982-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  12. Research activities on fast reactors in Switzerland

    International Nuclear Information System (INIS)

    Brogli, R.; Dones, R.; Hudina, M.; Pelloni, S.

    1996-01-01

    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

  13. Management of European fast reactor R and D

    International Nuclear Information System (INIS)

    Judd, A.M.; Sheriff, N.

    1993-01-01

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

  14. Fast reactor fuel pin behaviour modelling in the UK

    International Nuclear Information System (INIS)

    Matthews, J.R.; Hughes, H.

    1979-01-01

    Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

  15. Fast reactor fuel pin behaviour modelling in the UK

    Energy Technology Data Exchange (ETDEWEB)

    Matthews, J R [UKAEA, Harwell, Didcot, Oxon (United Kingdom); Hughes, H [Springfields Nuclear Power Development Laboratories, Springfields, Salwick, Preston (United Kingdom)

    1979-12-01

    Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

  16. The safety of the fast reactor

    International Nuclear Information System (INIS)

    Matthews, R.R.

    1977-01-01

    Verbatim of an address by R.R. Matthews, Chief Nuclear Health and Safety Officer, UK Central Electricity Generating Board given on January 15th 1977. The object of this address was to give some opinions on the safety issues of fast reactors as seen from an operational point of view. An outline of the basic responsibilities for nuclear safety is first given, and it is emphasized that the Central Electricity Generating Board has a statutory responsibility for the safe operation of its nuclear plant. The Nuclear Installations Act places absolute responsibility on the operator for ensuring that injury to persons and damage to property do not occur, and the new Health and Safety at Work Act does likewise. In addition the Board has a Nuclear Health and Safety Department that has to ensure that adequate provision for safety is made in the design, construction, and operation of nuclear plant, and safety at operational stations is monitored continuously by inspectors. In addition the requirements of the Nuclear Installations Inspectorate, laid down in the site licence conditions, must be satisfied. All these requirements are here discussed in the light of application to commercial fast reactors. It is considered that the hazards to fast reactor operating personnel are small and little different from those of other types of reactor, and in some respects the fast reactor has advantages, particularly in regard to the use of a Na coolant. The possibility of various types of accident is considered. Radioactive effluent discharge is also considered. The fast reactor as an international problem is discussed, including security matters. The extensive experience gained in operation of the experimental and prototype fast reactors at Dounreay is emphasized. (U.K.)

  17. Status of fast reactor activities in Russia

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  18. The dismantling of fast reactors: sodium processing

    International Nuclear Information System (INIS)

    Rodriguez, G.; Berte, M.; Serpante, J.P.

    1999-01-01

    Fast reactors require a coolant that does not slow down neutrons so water can not be used. Metallic sodium has been chosen because of its outstanding neutronic and thermal properties but sodium reacts easily with air and water and this implies that sodium-smeary components can not be considered as usual nuclear wastes. A stage of sodium neutralizing is necessary in the processing of wastes from fast reactors. Metallic sodium is turned into a chemically stable compound: soda, carbonates or sodium salts. This article presents several methods used by Framatome in an industrial way when dismantling sodium-cooled reactors. (A.C.)

  19. Discharges from a fast reactor reprocessing plant

    International Nuclear Information System (INIS)

    Barnes, D.S.

    1987-01-01

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

  20. Knowledge Preservation and Data Collection on Fast Reactors in France

    International Nuclear Information System (INIS)

    Vasile, A.

    2013-01-01

    Documentary packages: • MADONA - the French LMFR R&D Documentary collection; - A trilateral CEA, EDF & AREVA collection, 64 synthesis reports with 5300 relevant documents, on line. • RAPSODIE documentary collection - 18000 documents including 8000 drawings gathering available technical information on RAPSODIE; • PHENIX documentary fund - 140 000 documents. Design, specification, drawings, procedures, technical reports gathering available technical information on the PHENIX plant since 1968; • PHENIX lifetime extension project data base - 2000 reports. General methodology and feedback experience; • SUPERPHENIX Plant Archives - Several thousand of documents covering plant design, operation and decommissioning, since 1977; • SPX Safety report - 150 000 pages; - Plant characteristics & Technical options; - General safety analysis; - Manufacturing and operational rules (normal, incidental and anticipated operating conditions); • EFR 98 design documentation - around 5700 documents produced during the EFR studies; • EFR Synthesis reports - 48 synthesis reports; • ACCORE - 20 200 CEA technical documents on the period 1988-2001, with keyword research system; • The Fast Neutron Reactor System Knowledge Book - approx. 1400 licensing documents; • Sodium Technology data base - around 5500 technical documents on sodium circuits, sodium leak and fires, instrumentation, ISI&R, sodium water reaction in SG…; • CIR - The Reprocessing Information Centre data base gathered 25 000 reports, on reprocessing and analytical chemistry; among them around 600 reports concern Fast Reactor fuel reprocessing technology; • Database - Preservation of data relative to fast critical mock-ups: MASURCA (F), SNEAK(D), ZPPR (US), ZEBRA(UK), RRR/SEG(D). • ARCOPAC - Electronic data base on SUPERPHENIX physical measurements; • BREF Fuel Data Base - SFR fuels, behavior of structural materials; • Fatigue experimental data base for 316LN; • PHENIX expert Data base - Built

  1. The design rationale of the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Wade, D.C.; Hill, R.N.

    1997-01-01

    The Integral Fast Reactor (IFR) concept has been developed over the last ten years to provide technical solutions to perceptual concerns associated with nuclear power. Beyond the traditional advanced reactor objectives of increased safety, improved economy and more efficient fuel utilization, the IFR is designed to simplify waste disposal and increase resistance to proliferation. Only a fast reactor with an efficient recycle technology can provide for total consumption of actinides. The basic physics governing reactor design dictates that, for efficient recycle, the fuel form should be limited in burnup only by radiation damage to fuel cladding. The recycle technology must recover essentially all actinides. In a fast reactor, not all fission products need to be removed from the recycled fuel, and there is no need to produce pure plutonium. Recovery, recycle, and ultimate consumption of all actinides resolves several waste-disposal concerns. The IFR can be configured to achieve safe passive response to any of the traditional postulated reactor accident initiators, and can be configured for a variety of power output levels. Passive heat removal is achieved by use of a large inventory sodium coolant and a physical configuration that emphasizes natural circulation. An IFR can be designed to consume excess fissile material, to produce a surplus, or to maintain inventory. It appears that commercial designs should be economically competitive with other available alternatives. (author)

  2. Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

    International Nuclear Information System (INIS)

    Jammes, Christian; Filliatre, Philippe; Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan

    2015-01-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

  3. Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, Christian; Filliatre, Philippe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan [Division of Applied Nuclear Physics, Uppsala University, SE-75120 Uppsala, (Sweden)

    2015-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

  4. Sodium fast reactor safety and licensing research plan. Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

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

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1984-01-01

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

  6. Sodium Fast Reactor Safety and Licensing Research Plan

    International Nuclear Information System (INIS)

    Denman, M.; Lachance, J.; Sofu, T.; Bari, R.; Flanagon, G.; Wigeland, R.

    2015-01-01

    This paper summarizes potential research priorities for the US Department of Energy (DOE) with the intent of improving the licensability of the sodium cooled fast reactor (SFR). In support of this project, five panels were tasked with identifying potential safety related gaps in the available information, data and models needed to support the licensing of an SFR. The areas examined were sodium technology; accident sequences and initiators; source term characterization, codes and methods; and fuels and materials. It is the intent of this paper to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the applied technology access control designation from old documents. The second cross-cutting gap is the need for a robust knowledge management and preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with applied technology and knowledge management. (author)

  7. Sodium fast reactor safety and licensing research plan - Volume I

    International Nuclear Information System (INIS)

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-01-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  8. Nuclear data and multigroup methods in fast reactor calculations

    International Nuclear Information System (INIS)

    Gur, Y.

    1975-03-01

    The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)

  9. Back-to-back technical meetings (TMs): 'TM on the coordinated project (CRP) analyses of and lessons learned from the operational experience with fast reactor equipment and systems' and 'TM to coordinate the Agency's fast reactor knowledge preservation international project in Russia'. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    Since the early 1960's, several countries have undertaken important fast breeder reactor development programs. Fast test reactors were constructed and successfully operated in a number of countries, including Rapsodie (France), KNK-II (Germany), FBTR (India), JOYO (Japan), DFR (UK), BR-10, BOR-60 (Russia), and EBR-II, Fermi, FFTF (USA). This was followed by commercial size prototypes (Phenix, Superphenix (France), SNR-300 (Germany), MONJU (Japan), PFR (UK), BN-350 (Kazakhstan), BN-600 (Russia)], either just under construction, coming on line, or experiencing long term operation. However, from the 1980s onward, and mostly for economical and political reasons, fast reactor development in general began to decline. By 1994, in the USA, the Clinch River Breeder Reactor (CRBR) had been cancelled, and the two fast reactor test facilities, FFTF and EBR-II had been shutdown - EBRII permanently, and FFTF, until recently, in standby condition, but now also facing permanent closure. Thus, in the U.S., effort essentially disappeared for fast breeder reactor development. Similarly, programs in other nations were terminated or substantially reduced. In France, Superphenix was shut down at the end of 1998; SNR-300 in Germany was completed but not taken into operation, and KNK-II was permanently shut down in 1991 (after 17 years of operation) and is scheduled to be dismantled by 2004. In the UK, PFR was shut down in 1994, and in Kazakhstan, BN-350 was shut down in 1998. As the interest and activity in the fast breeder reactor diminished, the retirement of many of the developers and acknowledged experts of this technology reached its peak, between 1990 and 2000. The effort and investment required to replace these skills also diminished in parallel. In addition, the facilities (e.g., hot cells, fuel fabrication and inspection lines, seismic test rigs) required to develop and maintain the fast reactor program are drifting into a degraded state or are being shut down. This leads to the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-20

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

  11. Liquid metal fast reactor transient design

    International Nuclear Information System (INIS)

    Horak, C.; Purvis, E. III

    2000-01-01

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

  12. Thermophysical properties of fast reactor fuel

    International Nuclear Information System (INIS)

    Fink, J.K.

    1984-01-01

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

  13. Neutron shielding studies on an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    Merk, Bruno; Konheiser, Jörg

    2014-01-01

    Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

  14. Implications of Fast Reactor Transuranic Conversion Ratio

    International Nuclear Information System (INIS)

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

    2010-01-01

    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, 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 among the higher actinides, so the neutron emission varies much stronger with CR, about three orders of magnitude.

  15. Study on plant concept for gas cooled fast reactor

    International Nuclear Information System (INIS)

    Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

    2003-05-01

    In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

  16. Strategies for minority actinides transmutation in fast reactors

    International Nuclear Information System (INIS)

    Perez-Martin, S.; Martin-Fuertes, F.; Alvarez-Velarde, F.

    2010-01-01

    Presentation of the strategies that can be followed in fast reactors designed for the fourth generation to reduce the inventory of minority actinides generated in current light water reactors, as the actinides generation in fast reactor.

  17. Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors

    International Nuclear Information System (INIS)

    2002-06-01

    All prototype, demonstration and commercial liquid metal cooled fast reactors (LMFRs) have used liquid sodium as a coolant. Sodium cooled systems, operating at low pressure, are characterised by very large thermal margins relative to the coolant boiling temperature and a very low structural material corrosion rate. In spite of the negligible thermal energy stored in the liquid sodium available for release in case of leakage, there is some safety concern because of its chemical reactivity with respect to air and water. Lead, lead-bismuth or other alloys of lead, appear to eliminate these concerns because the chemical reactivity of these coolants with respect to air and water is very low. Some experts believe that conceptually, these systems could be attractive if high corrosion activity inherent in lead, long term materials compatibility and other problems will be resolved. Extensive research and development work is required to meet this goal. Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Federation in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. The need to exchange information on alternative fast reactor coolants was a major consideration in the recommendation by the Technical Working Group on Fast Reactors (TWGFRs) to collect, review and document the information on lead and lead-bismuth alloy coolants: technology, thermohydraulics, physical and chemical properties, as well as to make an assessment and comparison with respective sodium characteristics

  18. The safety of fast reactors

    International Nuclear Information System (INIS)

    Justin, F.

    1976-01-01

    A response is made to the main questions that a man in the street may arise concerning fast breeder reactors, in particular: the advantages of this line, dangerous materials contained in fast breeder reactors, containment shells protecting the environment from radiations, main studies now in progress [fr

  19. Properties of structural materials in liquid metal environment. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Borgstedt, H U [ed.

    1991-12-15

    The International Working Group on Fast Reactors (IWGFR) Specialists Meeting on Properties of Structural Materials in Liquid Metal Environment was held during June 18 to June 20, 1991, at the Nuclear Research Centre (Kernforschungszentrum) in Karlsruhe, Germany. The Specialists Meeting was divided into five technical sessions which addressed topics as follows: Creep-Rupture Behaviour of Structural Materials in Liquid Metal Environment; Behaviour of Materials in Liquid Metal Environments under Off-Normal Conditions;Fatigue and Creep-Fatigue of Structural Materials in Liquid Metal Environment; Crack Propagation in Liquid Sodium; and Conclusions and recommendations. Individual papers have been cataloged separately.

  20. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'

    International Nuclear Information System (INIS)

    Ito, Hideaki; Ashida, Takashi; Takamatsu, Misao

    2013-01-01

    Regarding the recovery of fuel exchange capability of 'Joyo', the replacement of Upper Core Structure (UCS) and the retrieval of the sample part of Material Testing Rig With Temperature Control (MARICO-2) are being planned for the fiscal years 2015 and 2016. In the recovery operation, sample part was planned to be removed through the hole created by removing USC because the size of existing rotation plug through-hole is smaller than the size of curved sample part. The procedure outline for this recovery operation was: (1) jack-up UCS, (2) pull out UCS and store in a cask, (3) retrieve sample part, and (4) install a new UCS. In this report, the status of UCS replacement, retrieval of sample part, as well as search and retrieval of loose parts are described. Regarding the replacement of UCS, the topics covered are: (1) removal of adhered sodium, (2) interference of UCS and small rotation plug, (3) the weight of cask for UCS storage, as well as (4) UCS jack-up jig and its functional test. Regarding the retrieval of sample part, the topics are: (1) gripping method selection, (2) pull-up method selection, and (3) safety measures and emergency correspondence. (S.K.)

  1. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Picker, C.; Ainsworth, K.F.

    1996-01-01

    The general position with regard to nuclear power and fast reactors in UK during 1995 is described. The status of fast reactor studies made in UK is outlined and a description and statement regarding the conclusions of the programme of studies associated with the closure of the Prototype Fast Reactor is included. (author)

  2. A review of the UK fast reactor programme

    Energy Technology Data Exchange (ETDEWEB)

    Picker, C [AEA Technolgy plc, Risley, Warrington, Cheshire (United Kingdom); Ainsworth, K F [British Nuclear Fuels plc, Sellafield, Cumbria (United Kingdom)

    1996-07-01

    The general position with regard to nuclear power and fast reactors in UK during 1995 is described. The status of fast reactor studies made in UK is outlined and a description and statement regarding the conclusions of the programme of studies associated with the closure of the Prototype Fast Reactor is included. (author)

  3. The fast reactor and electricity supply, a utility view

    International Nuclear Information System (INIS)

    Wright, J.K.; Hall, R.S.; Kemmish, W.B.; Thorne, R.T.

    1982-01-01

    The significance of the fast reactor is discussed from the viewpoint of the Central Electricity Generating Board. The need for the fast reactor and a possible timescale for its introduction are examined. It is emphasised that demonstration of the commercial and environmental acceptability of the fuel cycle will be needed before any commitment can be made to fast reactors. (U.K.)

  4. What is the future for fast reactor technology?

    International Nuclear Information System (INIS)

    Kraev, Kamen

    2017-01-01

    NucNet spoke to Vladimir Kriventsev, team leader for fast reactor technology development at the International Atomic Energy Agency (IAEA), about the possibilities and challenges of technology development in the fast reactor sector. Today, the field of fast reactors is vibrant and full of fascinating developments, some which will have an impact in the nearer term and others in the longer term.

  5. What is the future for fast reactor technology?

    Energy Technology Data Exchange (ETDEWEB)

    Kraev, Kamen [NucNet, Brussels (Belgium). The Independent Global Nuclear News Agency

    2017-08-15

    NucNet spoke to Vladimir Kriventsev, team leader for fast reactor technology development at the International Atomic Energy Agency (IAEA), about the possibilities and challenges of technology development in the fast reactor sector. Today, the field of fast reactors is vibrant and full of fascinating developments, some which will have an impact in the nearer term and others in the longer term.

  6. Sensitivity and optimization studies on plutonium vector variations for a plutonium burning fast reactor

    International Nuclear Information System (INIS)

    Hunter, Stuart N.

    2000-01-01

    Sensitivity studies were carried out on a 600 MW(e) Pu burning fast reactor, to determine the effects of changing Pu vector and the core design changes needed to adapt to a varying Pu vector. The applicability to Pu burner cores of models developed for breeder reactors was examined. The high flexibility of a fast reactor core for Pu burning was demonstrated by an optimization study to show the feasibility of using a single reactor design with Pu vectors varying from highly enriched (military) Pu to degraded Pu produced by multiple recycling. With fuel limited to MOX (∼45% Pu) and a single sub-assembly geometry for all grades of Pu, effective compensation for changes in Pu vector was achieved by replacing fuel with diluent material. The most suitable diluent had two components-absorber ( 10 B 4 C) and a moderator or neutron-transparent material (ZrH was most effective)-this gave an additional degree of freedom for optimizing safety-related core parameters. Where pin power ratings were high, hollow pellets introducing void as diluent were effective. Calculations demonstrated a possibility of flux distortions and anomalous rating distributions; these were a consequence of significant moderation of the flux in combination with the interaction between the core and the above/below core structures in the absence of breeder blankets. (author)

  7. Parameter analysis calculation on characteristics of portable FAST reactor

    International Nuclear Information System (INIS)

    Otsubo, Akira; Kowata, Yasuki

    1998-06-01

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

  8. The development of fast reactors in France

    International Nuclear Information System (INIS)

    Vautrey, L.

    1982-01-01

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

  9. Status of fast reactor activities in Brazil

    International Nuclear Information System (INIS)

    Menezes, Artur

    1996-01-01

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

  10. Nuclear data for advanced fast reactors

    International Nuclear Information System (INIS)

    Rabotnov, N.S.

    2001-01-01

    Interest revives to fast reactors as the only proven technology obviously able of satisfying human energy needs for the next millennium by using full energy content of both natural uranium resources and of vast stocks of depleted uranium. This interest stimulates revision and improvement of fast reactor ND. Progress in reactor calculations accuracy due to better codes and much faster computers also increases relative importance of the input data uncertainties, especially in case of small reactivity margin and fuels of equilibrium compositions. The main objects of corresponding R and D efforts should be minor actinides and heavy liquid metal coolant. Data error bands and covariance information also gain importance as necessary components of neutron physics calculations. (author)

  11. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  12. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

    Shoai Tehrani, Bianka; Da Costa, Pascal

    2013-01-01

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

  13. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1989-01-01

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

  14. A glossary of terms for fast reactors

    International Nuclear Information System (INIS)

    Wheeler, R.C.

    1979-04-01

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

  15. Non-electric Applications of Fast Reactors

    International Nuclear Information System (INIS)

    Safa, Henri; Borgard, Jean-Marc

    2013-01-01

    Conclusions: → Most of industrial applications (80%) require low temperature heat below 540°C; → Fast Reactors are technically suitable to provide industrial steam at temperatures not accessible by standard LWRs; → As an illustrative example, the application at an oil refinery site has been studied showing the economic benefits; → Nuclear Cogeneration enhances the overall energy efficiency of the power plant; • Nuclear Cogeneration allows massive cut in CO 2 emissions

  16. Fast reactor development programme in France

    Energy Technology Data Exchange (ETDEWEB)

    Le Rigoleur, C [Direction des Reacteurs Nucleaires, CEA Centre d` Etudes de Cadarache, Saint-Paul-lez-Durance (France)

    1998-04-01

    First the general situation regarding production of electricity in France is briefly described. Then in the field of Fast Reactors, the main events of 1996 are presented. At the end of February 1996, the PHENIX reactor was ready for operation. After review meetings, the Safety Authority has requested safety improvements and technical demonstrations, before it examines the possibility of authorizing a new start-up of PHENIX. The year 1996 was devoted to this work. In 1996, SUPERPHENIX was characterized by excellent operation throughout the year. The reactor was restarted at the end of 1995 after a number of minor incidents. The reactor power was increased by successive steps: 30% Pn up to February 6, followed by 50% Pn up to May then 60% up to October and 90% Pn during the last months. A programmed shutdown period occurred during May, June and mid-July 1996. The reactor has been shutdown at the end of 1996 for the decenial control of the steam generators. The status of the CAPRA project, aimed at demonstrating the feasibility of a fast reactor to burn plutonium at as high a rate as possible and the status of the European Fast Reactor are presented as well as their evolution. Finally the R and D in support of the operation of PHENIX and SUPERPHENIX, in support of the ````knowledge-acquisition```` programme, and CAPRA and EFR programmes is presented, as well as the present status of the stage 2 dismantling of the RAPSODIE experimental fast reactor. (author). 4 refs, figs, 2 tabs.

  17. Current status of fast reactor physics

    International Nuclear Information System (INIS)

    Hummel, H.H.

    1979-01-01

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

  18. Nodal method for fast reactor analysis

    International Nuclear Information System (INIS)

    Shober, R.A.

    1979-01-01

    In this paper, a nodal method applicable to fast reactor diffusion theory analysis has been developed. This method has been shown to be accurate and efficient in comparison to highly optimized finite difference techniques. The use of an analytic solution to the diffusion equation as a means of determining accurate coupling relationships between nodes has been shown to be highly accurate and efficient in specific two-group applications, as well as in the current multigroup method

  19. Fission energy: The integral fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1989-01-01

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

  20. Some aspects of fast reactor economics

    International Nuclear Information System (INIS)

    Kazachkovskij, O.D.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1993-01-01

    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

  2. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Tanaka, Nobuo

    2017-01-01

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

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

    International Nuclear Information System (INIS)

    Hoekstra, E.K.

    1969-12-01

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

  5. A Simplified Supercritical Fast Reactor with Thorium Fuel

    OpenAIRE

    Peng Zhang; Kan Wang; Ganglin Yu

    2014-01-01

    Super-Critical water-cooled Fast Reactor (SCFR) is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure ...

  6. A review of the European collaborative programme on fast reactors

    International Nuclear Information System (INIS)

    1989-01-01

    Considerable progress has been made in the European Fast Reactor R and D Collaboration during 1988. In parallel with the technical achievements, 1988 has been notable for the determination and willingness to adapt and improve the various structures and procedures in the organization of the European Collaboration. Three important agreements on EFR were signed in Bonn, Federal Republic of Germany on 16 February 1989: Industrial agreement, as a basis for planning, design, future construction and marketing of EFR; R and D agreement on the content and extent of R and D support for EFR design; SERENA-FASTEC agreement on know-how pool and royalties. (author). 1 tab

  7. Technical feasibility study of 60 MWe fast reactor concept: RAPID

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Ueda, Nobuyuki; Uotani, Masaki

    1993-01-01

    A study has been performed on the passive safety features and technical feasibility of an inherently safe 60 MWe fast reactor concept RAPID to meet various power requirements in Japan. The system dynamic analyses on the UTOP and ULOF transients revealed that the enhanced reactivity feedback derived from an annular core configuration and the integrated fuel assembly provides a high margin of self-protection. Structural integrity of the integrated fuel assembly has also been confirmed. The following innovative key technologies have been demonstrated; Lithium Injection Modules (LIM) for ultimate shutdown, Lithium Expansion Modulus (LEM) for inherent reactivity feedback and Void Leading Channel (VLC) for the sodium void worth reduction. (author)

  8. Preliminary Design of Compressor Impeller for innovative Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jekyoung; Cho, Seongkuk; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Cha, Jae Eun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For nuclear power plant application, applying S-CO{sub 2} Brayton cycle to Sodium cooled Fast Reactors and Small Modular Reactors are currently considered and active research is being performed by various research institutions and universities. As a part of research activities on the SCO{sub 2} Brayton cycle development for a nuclear power system, KAIST joint research team is currently working on an innovative Sodium cooled Fast Reactor (iSFR) development which utilizes S-CO{sub 2} Brayton cycle as its power conversion system. Various research subjects including reactor physics, thermo-hydraulics, material, cycle analysis and system integration are being considered as research issues currently. However, technical issues rising from dramatic change of thermodynamic property of CO{sub 2} near the critical point still remain as problems to be solved. As a result, 3D impeller model generation based on 1D mean stream line analysis results was successfully performed for non-airfoil blades. Since 3D model generation module works successfully, KAIST{sub T}MD can support 3D CFD analysis for internal flow structure in the designed impeller. Compressor loss mechanisms are complex phenomena and these are difficulties to be modeled while considering each loss mechanism separately.

  9. An assessment of methods of calculating sodium voiding reactivity in plutonium fuelled fast reactors

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Simmons, W.N.; Stevenson, J.M.

    1979-01-01

    After a survey of the requirements an assessment of the accuracy of calculations of the sodium void effect using UK methods and data is made on the basis of the following work. First, the analysis of small and large sodium voids in the MOZART and Zebra 13 small (300 MW(E)) fast reactor mock-ups and the BIZET large fast reactor mock-ups, all of conventional design. The analysis was carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code, whole reactor diffusion theory calculations of the neutron flux and perturbation theory methods. Exact perturbation theory was used in many cases, otherwise first order perturbation theory calculations were adjusted to give results equivalent to exact perturbation theory. Second, theoretical studies of some effects, including, the effects of extrapolating to fuel operating temperatures, fuel cycle and burn-up effects, and the heterogeneity effects of large fuelled subassemblies in pin geometry. Third, theoretical studies of approximations in the calculational methods including, the importance in the whole reactor calculation of the energy group structure and the spatial mesh, the importance of reactor material boundaries in the calculation of resonance shielding effects, and the use of neutron fluxes calculated using neutron diffusion theory rather than transport theory. (U.K.)

  10. Fabrication of Fast Reactor Fuel Pins for Test Irradiations

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-09-15

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

  11. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. Recovery of MARICO-2 sample part

    International Nuclear Information System (INIS)

    Ashida, Takashi; Ito, Hideaki

    2015-01-01

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. The following items are introduced here: (1) summary of restoration work and overall process of restoration work, (2) recovery operation of MARICO-2 sample part, (3) exchange of the upper core structure that was conducted this year, and (4) results of recovery of MARIKO-2 sample part. (A.O.)

  12. An ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Anish, E-mail: anish@igcar.gov.in; Rajkumar, K.V.; Sharma, Govind K.; Dhayalan, R.; Jayakumar, T.

    2015-02-15

    Highlights: • We demonstrate a novel ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast breeder reactor. • The methodology comprises of the inspection of shell weld immersed in sodium from the outside surface of the main vessel using ultrasonic guided wave. • The formation and propagation of guided wave modes are validated by finite element simulation of the inspection methodology. • A defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably using the developed methodology. - Abstract: The paper presents a novel ultrasonic methodology developed for in-service inspection (ISI) of shell weld of core support structure of main vessel of 500 MWe prototype fast breeder reactor (PFBR). The methodology comprises of the inspection of shell weld immersed in sodium from the outsider surface of the main vessel using a normal beam longitudinal wave ultrasonic transducer. Because of the presence of curvature in the knuckle region of the main vessel, the normal beam longitudinal wave enters the support shell plate at an angle and forms the guided waves by mode conversion and multiple reflections from the boundaries of the shell plate. Hence, this methodology can be used to detect defects in the shell weld of the core support structure. The successful demonstration of the methodology on a mock-up sector made of stainless steel indicated that an artificial defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably.

  13. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 1. MARICO-2 subassembly retrieval work

    International Nuclear Information System (INIS)

    Naito, Hiroyuki; Ashida, Takashi; Ito, Hideaki

    2014-01-01

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. This paper introduces the progress of restoration work and the future work plan, with a focus on the outline of overall restoration work, the method / problems / measures for MARICO-2 sample part recovery operations, and fabrication of sample part recovery device. (A.O.)

  14. Application of MCNP code in shielding calculation of minitype fast reactor

    International Nuclear Information System (INIS)

    He Keyu; Han Weishi

    2008-01-01

    An accurate shielding calculation model has been set up for the minitype sodium-cooled fast reactor (MFR) based on MCNP code and particular calculation of its primary shielding parameters has been carried out. The results indicate that the photon and neutron flux density of MFR has rapidly fallen to a low-level. The material for the shielding layer outside of main container is primarily of carbon steel, which can be design as a shielding structure satisfying the safety code. The sodium activation in primary circuit is extremely limited and it is simple to shield from. Both the output of helium in reflector and burn up of boron-10 in control rod are very small. These materials can be used for several cycle lives. (authors)

  15. GENIUS & the Swedish Fast Reactor programme

    International Nuclear Information System (INIS)

    Wallenius, Janne

    2012-01-01

    Concluding remarks: Sweden’s growing fast reactor programme focuses on LFR technology, but we also participate in ASTRID. • An innovative facility for UN fabrication, an LBE thermal hydraulics loop and a lead corrosion facility are operational. • A plutonium fuel fabrication lab is is under installation (this week!) • The government is assessing the construction of ELECTRA-FCC, a centre for Gen IV-system R&D, at a tentative cost of ~ 140±20 M€. • Location: Oskarshamn (adjacent to intermediate repository) • Date of criticality: 2023 (best case) • Swedish participation in IAEA TWG-FR should intensify

  16. Integral fast reactor shows its mettle

    International Nuclear Information System (INIS)

    Chang, Ya.; Lajnberri, M.; Barris, L.; Uoters, L.

    1988-01-01

    The main aspects of the problem of developing a closed fuel cycle at a NPP built in the so-called integrated version when a fast reactor and the plant for spent fuel regeneration and fuel element production are located in the same site (IFR project), are considered. The technologies of U-Pu-Zr alloy fuel reprocessing and production based on high-temperature metallurgical process and the method of casting under pressure are described. The demonstration of practical feasibility of the fuel cycle on the basis of the IFR reactor is planned for 1990

  17. Integral fast reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  18. Integral Fast Reactor fuel pin processor

    International Nuclear Information System (INIS)

    Levinskas, D.

    1993-01-01

    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

  19. The Argentine-Brazilian fast reactor programme

    International Nuclear Information System (INIS)

    Gho, C.J.; Mauricio, A.

    1989-01-01

    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

  20. Integral Fast Reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  1. Liquid metal cooled experimental fast reactor simulator

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  2. The Integral Fast Reactor (IFR) concept

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  3. Investigation of high temperature phase stability, thermal properties and evaluation of metallurgical compatibility with 304L stainless steel, of indigenously developed ferroboron alternate shielding material for fast reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Rai, Arun Kumar [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, 603102 (India); Raju, S., E-mail: sraju@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, 603102 (India); Jeya Ganesh, B. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, 603102 (India); Panneerselvam, G. [Chemistry Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, 603102 (India); Vijayalakshmi, M.; Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, 603102 (India); Raj, Baldev [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, 603102 (India); Chemistry Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, 603102 (India)

    2011-08-15

    Highlights: > High temperature phase stability and metallurgical compatibility of indigenous Ferro boron alloy (17 wt.%) with 304L SS have been evaluated using calorimetry, metallography and X-ray diffraction. > Fe-B is stable upto 1227 deg. C without any phase transformation. > In combination with SS 304L, it develops an eutectic reaction at 1198 deg. C. > High temperature metallurgical interaction between Ferroboron and 304L SS clad has been simulated for temperatures 550-800 deg. C for reaction times up to 5000 h, using diffusion couple. > Clad penetration depth with time follows a paraobolic rate law. - Abstract: Towards the cause of serving economic power production through fast reactors, it is necessary to bring in functionally more efficient and innovative design options, which also includes exploration of cheaper material alternatives, wherever possible. In this regard, the feasibility of using a commercial grade ferroboron alloy as potential alternate shielding material in the outer subassemblies of future Indian fast reactors has been recently investigated from shielding physics point of view. The present study explores in detail the high temperature thermal stability and the metallurgical compatibility of Fe-15.4B-0.3C-0.89Si-0.17Al-0.006S-0.004P-0.003O (wt.%) alloy with SS 304L material. In addition, the high temperature specific heat and lattice thermal expansion characteristics of this alloy have also been investigated as a part of the present comprehensive characterisation program. The Fe-15 wt.%B alloy is constituted of principally of two boride phases, namely tetragonal Fe{sub 2}B and orthorhombic FeB phases, which in addition to boron also contains some amount of C and Si dissolved in solid solution form. This Fe-B alloy undergoes a series of phase transformation as a function of increasing temperature; the major ones among them are the dissolution of Fe{sub 2}B-lower boride in the matrix through a eutectic type reaction, which results in the

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

    International Nuclear Information System (INIS)

    Pechenkin, V.A.

    2012-01-01

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

  5. Fast reactors will eat nuclear waste from LWR

    International Nuclear Information System (INIS)

    Tokiwai, Moriyasu

    1999-01-01

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

  6. Fast reactors will eat nuclear waste from LWR

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

  7. 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.

  8. The integral fast reactor - an overview

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.; Hannum, W.H.

    1997-01-01

    The Integral Fast Reactor (IFR) is a system that consists of a fast-spectrum nuclear reactor that uses metallic fuel and liquid-metal (sodium) cooling, coupled with technology for high-temperature electrochemical recycling, and with processes for preparing wastes for disposition. The concept is based on decades of experience with fast reactors, adapted to priorities that have evolved markedly from those of the early days of nuclear power. It has four essential, distinguishing features: efficient use of natural resources, inherent safety characteristics, reduced burdens of nuclear waste, and unique proliferation resistance. These fundamental characteristics offer benefits in economics and environmental protection. The fuel cycle never involves separated plutonium, immediately simplifying the safeguarding task. Initiated in 1984 in response to proliferation concerns identified in the International Nuclear Fuel Cycle Evaluation (INFCE, 1980), the project has made substantial technical progress, with new potential applications coming to light as nuclear weapons stockpiles are reduced and concerns about waste disposal increase. A breakthrough technology, the IFR has the characteristics necessary for the next nuclear age. (author)

  9. Seismic appraisal test of control rod drive mechanism of China experiment fast reactor

    International Nuclear Information System (INIS)

    Song Qing; Yang Hongyi; Jing Yueqing; Wen Jing; Liu Guijuan; Sun Lei

    2008-01-01

    The structure of the control rod drive mechanism in pool type sodium-cooled fast reactor is the characterized by long, thin, and geometric nonlinearity, and the seismic load is multiple activation. The anti-seismic evaluation is always paid great attention by the countries developing the technology worldwide. This article introduces the seismic appraisal test of the control rod drive mechanism of China Experimental Fast Reactor (CEFR) performed on a seismic platform which is vertical shaft style and multiple activation. The result of the test shows the structural integrity and the function of the control rod drive mechanism could meet the design requirements of the earthquake intensity. (authors)

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

    International Nuclear Information System (INIS)

    Dombey, N.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Duthie, J.C.; Dostal, M.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

    An evaluation is made on a large MOX fuel fast reactor's capability of burning 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)

  14. The integral fast reactor fuels reprocessing laboratory at Argonne National Laboratory, Illinois

    International Nuclear Information System (INIS)

    Wolson, R.D.; Tomczuk, Z.; Fischer, D.F.; Slawecki, M.A.; Miller, W.E.

    1986-09-01

    The processing of Integral Fast Reactor (IFR) metal fuel utilizes pyrochemical fuel reprocessing steps. These steps include separation of the fission products from uranium and plutonium by electrorefining in a fused salt, subsequent concentration of uranium and plutonium for reuse, removal, concentration, and packaging of the waste material. Approximately two years ago a facility became operational at Argonne National Laboratory-Illinois to establish the chemical feasibility of proposed reprocessing and consolidation processes. Sensitivity of the pyroprocessing melts to air oxidation necessitated operation in atmosphere-controlled enclosures. The Integral Fast Reactor Fuels Reprocessing Laboratory is described

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  16. Conceptual design of module fast reactor of ultimate safety cooled by lead-bismuth alloy

    International Nuclear Information System (INIS)

    Myasnikov, V.O.; Stekolnikov, V.V.; Stepanov, V.S.; Gorshkov, V.T.; Kulikov, M.L.; Shulyndin, V.A.; Gromov, B.F.; Kalashnikov, A.G.; Pashkin, Yu.G.

    1993-01-01

    During past time all basic problems arisen during working-out of NPP with lead-bismuth coolant were solved: physics and thermal physics of the cores, heat transfer and hydrodynamics, corrosion resistance of the structural materials and coolant technology, radiation and nuclear safety, investigations of emergency situations, development of fuel elements and absorbing elements of the reactor, equipment of the primary circuit and other circuits. A powerful experimental base equpped by unique rigs is made. A series of ship and test NPP has been constructed whereat repair of the plants and reactor refuelling are developed. Highly-skilled groups of investigators, designers and operation personnel capable of performing the development of the reactor plant with MFR within short terms have been formed. In this case MFR with lead-bismuth coolant may become the initial step in development of large-scale nuclear power engineering with fast reactors cooled by liquid lead

  17. Fuel cycle facility control system for the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    Benedict, R.W.; Tate, D.A.

    1993-01-01

    As part of the Integral Fast Reactor (IFR) Fuel Demonstration, a new distributed control system designed, implemented and installed. The Fuel processes are a combination of chemical and machining processes operated remotely. To meet this special requirement, the new control system provides complete sequential logic control motion and positioning control and continuous PID loop control. Also, a centralized computer system provides near-real time nuclear material tracking, product quality control data archiving and a centralized reporting function. The control system was configured to use programmable logic controllers, small logic controllers, personal computers with touch screens, engineering work stations and interconnecting networks. By following a structured software development method the operator interface was standardized. The system has been installed and is presently being tested for operations

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1981-01-01

    steam cycle, a re-evaluation of various boiler design styles have been undertaken. The types being studied cover straight tube, U and J tube, hockey stick and helical coil designs. A final choice of steam generator for CDFR has not yet been made. Key features in design selection are overall integrity to ensure high availability, ease of maintenance and repair following minor tube leaks, resistance to escalation of minor leaks to more major sodium/water reaction incidents. inspectability, relevance of previous experience and capital cost. Operation of the PFR was interrupted by a series of small leaks in the tube-to-tube plate welds in the three evaporator units. The key areas have now been shot-peened to remove tensile stresses in the surface material. the units have been repaired and operation at high power should be restarted during March 1981. The performance of the core and primary circuits has continued to be excellent. A major step forward was taken when the first 16 sub-assemblies were reprocessed in the PFR reprocessing plant. During the year a system economic study has been completed to assist in providing a sound technical basis for a fast reactor inquiry. This study was based on a set of programme assumptions and economic ground rules developed and agreed in 1978 within the UK nuclear industry and updated in 1979. Broadly, it was concluded that the introduction of the fast reactor as soon and as rapidly as feasible also leads, even on conservative uranium price assumptions. to economic benefits, the benefits being greatest when electrical demand and raw fuel prices are at their highest. The benefits are little affected by 25% increases in costs, except for an increase in fast reactor capital cost when uranium stays cheap; in this case the benefit virtually disappears if it is assumed that all other capital costs stay constant. Use of the fast reactor also produces a dramatic fall (almost three-fold) in the need uranium imports and, if market forces are

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-05-01

    steam cycle, a re-evaluation of various boiler design styles have been undertaken. The types being studied cover straight tube, U and J tube, hockey stick and helical coil designs. A final choice of steam generator for CDFR has not yet been made. Key features in design selection are overall integrity to ensure high availability, ease of maintenance and repair following minor tube leaks, resistance to escalation of minor leaks to more major sodium/water reaction incidents. inspectability, relevance of previous experience and capital cost. Operation of the PFR was interrupted by a series of small leaks in the tube-to-tube plate welds in the three evaporator units. The key areas have now been shot-peened to remove tensile stresses in the surface material. the units have been repaired and operation at high power should be restarted during March 1981. The performance of the core and primary circuits has continued to be excellent. A major step forward was taken when the first 16 sub-assemblies were reprocessed in the PFR reprocessing plant. During the year a system economic study has been completed to assist in providing a sound technical basis for a fast reactor inquiry. This study was based on a set of programme assumptions and economic ground rules developed and agreed in 1978 within the UK nuclear industry and updated in 1979. Broadly, it was concluded that the introduction of the fast reactor as soon and as rapidly as feasible also leads, even on conservative uranium price assumptions. to economic benefits, the benefits being greatest when electrical demand and raw fuel prices are at their highest. The benefits are little affected by 25% increases in costs, except for an increase in fast reactor capital cost when uranium stays cheap; in this case the benefit virtually disappears if it is assumed that all other capital costs stay constant. Use of the fast reactor also produces a dramatic fall (almost three-fold) in the need uranium imports and, if market forces are

  20. High flux-fluence measurements in fast reactors

    International Nuclear Information System (INIS)

    Lippincott, E.P.; Ulseth, J.A.

    1977-01-01

    Characterization of irradiation environments for fuels and materials tests in fast reactors requires determination of the neutron flux integrated over times as long as several years. An accurate integration requires, therefore, passive dosimetry monitors with long half-life or stable products which can be conveniently measured. In addition, burn-up, burn-in, and burn-out effects must be considered in high flux situations and use of minimum quantities of dosimeter materials is often desirable. These conditions force the use of dosimeter and dosimeter container designs, measured products, and techniques that are different from those that are used in critical facilities and other well-characterized benchmark fields. Recent measurements in EBR-II indicate that high-accuracy results can be attained and that tie-backs to benchmark field technique calibrations can be accomplished

  1. Development of guidelines for inelastic analysis in design of fast reactor components

    International Nuclear Information System (INIS)

    Nakamura, Kyotada; Kasahara, Naoto; Morishita, Masaki; Shibamoto, Hiroshi; Inoue, Kazuhiko; Nakayama, Yasunari

    2008-01-01

    The interim guidelines for the application of inelastic analysis to design of fast reactor components were developed. These guidelines are referred from 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)'. The basic policies of the guidelines are more rational predictions compared with elastic analysis approach and a guarantee of conservative results for design conditions. The guidelines recommend two kinds of constitutive equations to estimate strains conservatively. They also provide the methods for modeling load histories and estimating fatigue and creep damage based on the results of inelastic analysis. The guidelines were applied to typical design examples and their results were summarized as exemplars to support users

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  3. Future plans for the design and construction of fast reactor power stations in Italy

    International Nuclear Information System (INIS)

    Castelli, G.; Ghilardotti, G.

    1978-01-01

    Studies related to fast reactor technology have been pursued in Italy for a long time and this country is now deeply engaged in the demonstration and marketing phases, in accordance with the outlines of the Italian national energy plan. In the paper the following topics are examined: current possibilities for introducing fast reactors in Italy; the main social and political constraints concerning their introduction; the necessary industrial and organizational structures (in the broadest meaning) existing or foreseen; the national programme pertaining to activities towards achieving this goal. (author)

  4. Fast reactor knowledge preservation system: Taxonomy and basic requirements

    International Nuclear Information System (INIS)

    2008-01-01

    The IAEA has taken the initiative to coordinate efforts of Member States in the preservation of knowledge in the area of fast reactors. In the framework of this initiative, the IAEA intends to create an international database compiling information from different Member States on fast reactors through a web portal. Other activities related to this initiative are being designed to accumulate and exchange information on the fast reactor area, to facilitate access to this information by users in different countries and to assist Member States in preserving the experience gained in their countries. The purpose of this publication is to develop a taxonomy of the Fast Reactor Knowledge Preservation System (FRKPS) that will facilitate the preservation of the world's fast reactor knowledge base, to identify basic requirements of this taxonomy on the basis of the experience gained in the fast reactor area, as well as results of previous IAEA activities on fast reactor knowledge preservation. The need for such taxonomy arises from the fact that during the past 15 years there has been stagnation in the development of fast reactors in the industrialized countries that were involved, earlier, in intensive development of this area. All studies on fast reactors have been stopped in countries such as Germany, Italy, the United Kingdom and the United States of America and the only work being carried out is related to the decommissioning of fast reactors. Many specialists who were involved in the studies and development work in this area in these countries have already retired or are close to retirement. In countries such as France, Japan and the Russian Federation that are still actively pursuing the evolution of fast reactor technology, the situation is aggravated by the lack of young scientists and engineers moving into this branch of nuclear power

  5. Fast reactor versions: elements of choice

    International Nuclear Information System (INIS)

    Tassart, J.; Zerbib, J.C.

    1984-01-01

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

  6. Analysis of fast reactor steam generator performance

    International Nuclear Information System (INIS)

    Hulme, G.; Curzon, A.F.

    1992-01-01

    A computer model for the prediction of flow and temperature fields within a fast reactor steam generator unit is described. The model combines a commercially available computational fluid dynamics (CFD) solver (PHOENICS) with a steam-tube calculation and provides solutions for the fully coupled flow and temperature fields on both the shell side and the tube side. The model includes the inlet and outlet headers and the bottom end stagnant zone. It also accounts for the effects of support grids and edge-gaps. Two and three dimensional and transient calculations have been performed for both straight tube and J-tube units. Examples of the application of the model are presented. (7 figures) (Author)

  7. Actinide behavior in the integral fast reactor

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1993-05-01

    Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory's site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core

  8. Improvement of covariance data for fast reactors

    International Nuclear Information System (INIS)

    Shibata, Keiichi; Hasegawa, Akira

    2000-02-01

    We estimated covariances of the JENDL-3.2 data on the nuclides and reactions needed to analyze fast-reactor cores for the past three years, and produced covariance files. The present work was undertaken to re-examine the covariance files and to make some improvements. The covariances improved are the ones for the inelastic scattering cross section of 16 O, the total cross section of 23 Na, the fission cross section of 235 U, the capture cross section of 238 U, and the resolved resonance parameters for 238 U. Moreover, the covariances of 233 U data were newly estimated by the present work. The covariances obtained were compiled in the ENDF-6 format. (author)

  9. Safe Management Of Fast Reactors: Towards Sustainability

    International Nuclear Information System (INIS)

    Dreimanis, Andrejs

    2015-01-01

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

  10. Fourteenth annual meeting of the International Working Group on Fast Reactors. Summary report. Part II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1981-05-01

    This report includes description of the state-of-the art in the field of fast reactor technology, research and development, in France, Belgium, India, Italy, USSR, USA, UK, Switzerland, and European Union. The emphasis in the majority of the reports is on the FBR safety issues, sodium cooling system, fuel elements development, reactor materials testing, risk assessment.

  11. Toward a sustainable energy supply with reduced environmental burden. Development of metal fuel fast reactor cycle

    International Nuclear Information System (INIS)

    Koyama, Tadafumi; Kobayashi, Hiroaki; Kinoshita, Kensuke

    2009-01-01

    CRIEPI has been studying the metal fuel fast reactor cycle as an outstanding alternative for the future energy sources. In this paper, development of the metal fuel cycle is reviewed in the view point of technological feasibility and material balance. Preliminary estimation of reduction of the waste burden due to introduction of the metal fuel cycle technology is also reported. (author)

  12. Recent Development of Radioanalytical Methods at the IBR-2 Pulsed Fast Reactor

    International Nuclear Information System (INIS)

    Nazarov, V.M.; Peresedov, V.F.

    1994-01-01

    Experience in the application of radioanalytical methods, including NAA, at the IBR-2 pulsed fast reactor is reviewed. Details of the instruments dedicated to neutron activation analysis and radiography studies are reported. Applications of resonance neutrons to environmental monitoring and to the investigation of high-purity materials, are examplified. 15 refs. 9 figs., 9 tabs

  13. Fourteenth annual meeting of the International Working Group on Fast Reactors. Summary report. Part II

    International Nuclear Information System (INIS)

    1981-05-01

    This report includes description of the state-of-the art in the field of fast reactor technology, research and development, in France, Belgium, India, Italy, USSR, USA, UK, Switzerland, and European Union. The emphasis in the majority of the reports is on the FBR safety issues, sodium cooling system, fuel elements development, reactor materials testing, risk assessment

  14. Fast Reactor Knowledge Organization System (FR-KOS)

    International Nuclear Information System (INIS)

    Costantini, Livio

    2013-01-01

    FR-KOS Web-portal general concepts: FR-KOS web-portal is an integrated and unique solution for knowledge preservation, retrieval and management in fast reactor (FR) knowledge domain; The exploitation of the FR-KOS provides an effective and flexible tool to access FR knowledge repositories. The dynamic and interactive process of searching information is guided by structures that are based on the FR Taxonomy; FR-KOS can be extended to other types of reactor and technologies, such as PWR, WWER, BWR, CANDU, etc. In addition can also be utilized in other information management process such as the preservation of nuclear accidents knowledg;. FR-KOS addresses needs in non-structured knowledge preservation for both embarking countries and countries with established nuclear power; FR-KOS repositories contain more that 50.000 samples of technical and scientific publications, including full text documents in different languages (English, German, French and eventually in Russian)

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

    International Nuclear Information System (INIS)

    2017-01-01

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

  16. Status of Fast Reactor Research and Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-04-01

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

  17. Status of Fast Reactor Research and Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-01

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

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

    International Nuclear Information System (INIS)

    Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

    1985-01-01

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

  19. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    2012-01-01

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

  20. The problems of thermohydraulics of prospective fast reactor concepts

    International Nuclear Information System (INIS)

    Sedov, A.A.

    2000-01-01

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

  1. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    2013-01-01

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

  2. A review of fast reactor activities in Italy

    International Nuclear Information System (INIS)

    Tavoni, R.

    1996-01-01

    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)

  3. Overview of major HZDR developments for fast reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Merk, Bruno, E-mail: b.merk@hzdr.de; Glivici-Cotruţă, V.; Duerigen, S.; Rohde, U.; Kliem, S.

    2013-12-15

    The upgrading of the DYN3D code for the application for fast reactors is described. After extension and validation, a diverse code with the possibility for steady state and transient core analysis on the basis of coupled thermal hydraulics/neutronics calculations is available. The work on the use of fine distributed moderating material in SFR cores is discussed with the target on enhancing the feedback effects in SFR cores without influencing the operational. Newly developed analytical solutions without separation of space and time for the analysis of ADS experiments are shown with good agreement for the YALINA experiment. The analytical solutions are a very promising tool for the development of a new method for the analysis of ADS experiments.

  4. Characterisation of the activities associated with fast reactor cladding hulls

    International Nuclear Information System (INIS)

    Biddle, P.; Rees, J.H.

    1987-01-01

    Four batches of hulls from the reprocessing of fuel from the Prototype Fast Reactor at Dounreay have been examined visually and photographically. The appearance of the hulls (shiny or black) depended on the original location in the fuel pin. The total fissile activities measured corresponded to less than 0.1% or the original fuel. The α-emitter, fission product and activation product activities have been determined from dissolution studies. Profile diagrams have been constructed for the distribution of the principal activities along lengths of seventeen hulls. The evidence obtained suggested that much of the activity on the hulls arose through deposition from the dissolver liquors. The deposits included insoluble particulate material which was thought to be responsible for the wide variations in activity across short lengths of some of the hulls. (orig.)

  5. An assessment of methods of calculating sodium-voiding reactivity in plutonium-fuelled fast reactors

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Simmons, W.N.; Stevenson, J.M.

    1980-01-01

    After a survey of the requirements an assessment of the accuracy of calculations of the sodium-void effect using UK methods and data is made on the basis of the following work: (a) The analysis of small and large sodium voids in the MOZART and Zebra 13 small (300 MW(e)) fast reactor mock-ups and the BIZET large fast reactor mock-ups, all of conventional design. The analysis was carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code, whole reactor diffusion theory calculations of the neutron flux and perturbation theory methods. Exact perturbation theory was used in many cases, otherwise first-order perturbation theory calculations were adjusted to give results equivalent to exact perturbation theory. (b) Theoretical studies of some effects, including the following: (i) The effects of extrapolating to fuel operating temperature; (ii) Fuel-cycle and burnup effects, including the gradual replacement through a fuel cycle of control-rod absorption by fission product absorption, the loss of fissile material and the change in fuel nuclide relative composition; (iii) The heterogeneity effects of large fuelled subassemblies in pin geometry. (c) Theoretical studies of approximations in the calculational methods, including the following: (i) The importance in the whole reactor calculation of the energy group structure and the spatial mesh, including comparisons of calculations in two (RZ) and three-dimensional geometry; (ii) The importance of reactor material boundaries in the calculation of resonance shielding effects; (iii) The use of neutron fluxes calculated using neutron diffusion theory rather than transport theory. (author)

  6. Logistical and economic obstacles to a fast reactor programme

    International Nuclear Information System (INIS)

    Sweet, C.

    1982-01-01

    Fast reactor studies used to place great emphasis on its role as a breeder of plutonium, thereby raising the prospect of a nuclear power system free from the constraint of uranium supplies. Today, not only the timing of fast reactor introduction has slipped (by two or three decades) but the perspective which was central to energy policy has changed dramatically. This article first examines the fast reactor as a system and looks at the interaction of four key variables in its logistics. It then looks at the rise in real costs, especially capital costs. Given the parameters that determine the plutonium balance and the economics of the fast reactor system, the author questions whether there is a sound basis for its introduction, and concludes by suggesting that the most pressing requirement is a study of the opportunity costs of fast reactor R and D expenditures. (author)

  7. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Picker, C.; Ainsworth, K.F.

    1998-01-01

    The general position with regard to nuclear power and fast reactors in the UK during 1996 is described. The main UK Government-funded fast reactor research and development programme was concluded in 1993, to be replaced by a smaller programme which is mainly funded and managed by British Nuclear Fuels plc. The main focus of this programme sustains the UK participation in the European Fast Reactor (EFR) collaboration and the broader international links built-up over the previous decades. The status of fast reactor studies made in the UK in 1996 is outlined and, with respect to the Prototype Fast Reactor at Dounreay, a report of progress with the closure studies, fuel reprocessing and decommissioning activities is provided. (author)

  8. Design, construction and operating experience of demonstration LMFBRs. The application of core and fuel performance experience in British reactors to commercial fast reactor design

    International Nuclear Information System (INIS)

    Bagley, K.Q.

    1978-01-01

    The Prototype Fast Reactor (PFR) sub-assembly design is described with particular emphasis on the choice of factors that are important in determining satisfactory performance. Reasons for the adoption of specific clad and fuel design details are given in their historical context, and irradiation experience - mostly from the Dounreay Fast Reactor (DFR) - in support of the choices is described. The implications of factors that are now better understood than when the PFR fuel was designed, notably neutron-induced void swelling and irradiation creep, are then considered. It is shown that the 'free-standing' core design used in PFR, in which the sub-assembly is unsupported above the level of the lower axial breeder, relies on the availability of low-swelling, preferably irradiation-creep-resistant alloys as sub-assembly structural materials in order to achieve the prescribed burn-up target. The advantages of a 'restrained core', which makes use of irradiation creep to redress the effects of material swelling, are noted briefly, and the application of this concept to the Commercial Demonstration Fast Reactor (CDFR) core design is described. Probable future trends in pin and sub-assembly design are reviewed and the scope of associated irradiation testing programmes defined. Arrangements for monitoring and evaluating fuel performance, both in reactor and post-irradiation, are outlined and the provisions for endorsement of CDFR pin, sub-assembly and core design details in PFR are indicated. (author)

  9. Methodology for substantiation of the fast reactor fuel element serviceability

    International Nuclear Information System (INIS)

    Tsykanov, V.A.; Maershin, A.A.

    1988-01-01

    Methodological aspects of fast reactor fuel element serviceability substantiation are presented. The choice of the experimental program and strategies of its realization to solve the problem set in short time, taking into account available experimental means, are substantiated. Factors determining fuel element serviceability depending on parameters and operational conditions are considered. The methodological approach recommending separate studing of the factors, which points to the possibility of data acquisition, required for the development of calculational models and substantiation of fuel element serviceability in pilot and experimental reactors, is described. It is shown that the special-purpose data are more useful for the substantiation of fuel element serviceability and analytical method development than unsubstantial and expensive complex tests of fuel elements and fuel assemblies, which should be conducted only at final stages for the improvement of the structure on the whole

  10. Description of the prototype fast reactor at Dounreay

    International Nuclear Information System (INIS)

    Jensen, S.E.; Oelgaard, P.L.

    1995-12-01

    The Prototype Fast Reactor (PFR) at Dounreay, UK, started operation in 1975 and was closed down in 1994. The present report contains a description of the PFR nuclear power plant, based on information available in literature and on information supplied during a visit to the plant. The report covers a description of the site and plant arrangement, the buildings and structures, the reactor core and other vessel internals, the control system, the main cooling system, the decay heat removal system, the emergency core cooling system, the containment system, the steam and power conversion system, the fuel handling system, plant safety features, the control and instrumentation systems and the sodium purification systems. (au) 16 refs

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

    Energy Technology Data Exchange (ETDEWEB)

    Marchetti, Marco

    2017-07-01

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

  12. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

  13. Nonproliferation and safeguards aspects of the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Hannum, W.H.; Wade, D.C.; McFarlane, H.F.; Hill, R.N.

    1997-01-01

    The National Academy of Sciences (NAS) has declared that the large and growing stocks of plutonium from weapons dismantlement in the U.S. and the former Soviet Union FSU are a ''clear and present danger'' to peace and security. Moreover, the opinion of some experts that plutonium of any isotopic blend is a proliferation threat has been well publicized, heightening the concern that plutonium produced in the civilian fuel cycle is itself a proliferation threat. Assuring that separated plutonium, from dismantled warheads as well as from civilian power programs, is under effective control has (again) become a high priority of U.S. diplomacy. One pole of the debate on how to manage this material is to declare it to be a waste, and to search for some way to dispose of it safely, securely, and permanently. The other pole is to view it as an energy resource and to safeguard it against diversion, putting it into active use in the civilian power program. The ultimate choice cannot be separated from the long-term strategy for use of peaceful nuclear power. Continued use of a once-through fuel cycle will lead to an ever-increasing quantity of excess plutonium-requiring safeguarding. Alternatively, recycling the world's stocks of plutonium in fast reactors, contrary to common misconception, will cap the world supply of plutonium and hold it in working inventories for generating power. Transition from the current-generation light water cooled reactors (LWRs) to a future fast-reactor-based nuclear energy supply under international safeguards would, henceforth, limit world plutonium inventories to the amount necessary and useful for power generation, with no further excess production. (author)

  14. Activities of the AZTLAN team on the OECD/Nea benchmark on fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lopez S, R.; Gomez T, A.; Puente E, F. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E.; Arriaga R, L., E-mail: armando.gomez@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, 07738 Ciudad de Mexico (Mexico)

    2017-09-15

    In the present paper, the activities of the AZTLAN Platform Fast Reactor Group on the OECD/Nea Benchmark will be described. The main objective of these activities is to test the group staff and capabilities as well as the domestic code reliability by putting them into test in this exercise with different institutions from around the world. Six different core configurations were treated; these are described in two different versions of the Benchmark document. The main tools used by the group were the Finnish stochastic Monte Carlo code Serpent for full core calculations and macroscopic Cross Sections (X S) generation, and the domestic deterministic code AZNHEX for full core calculations. Different calculations were performed, such as full core calculations under nominal conditions, with control rods fully and partially inserted and with the sodium voided in the active zone as well as different reactivity shift values due to various conditions of radial and axial expansion of the fuel elements and structural material. The results obtained in the full core calculations and most of the reactivity shift calculations obtained by our group were indeed comparable to the ones obtained by different institutions when using similar methodologies. Given these favorable results it can be said that the main objective was met and the group showed their capabilities, as well as its possibility to collaborate with other institutes, placing Mexico in a good position in fast reactor analysis. Future work will continue with the calculations not yet treated and with the new core specifications on the new versions of the Benchmark document. (Author)

  15. Shape optimization of a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Schmitt, D.; Allaire, G.; Pantz, O.; Pozin, N.

    2013-01-01

    Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth. Usual optimization methods for core conception are based on a parametric description of a given core design. New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints. First studies show that these methods could be applied to sodium cooled core conception. In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a given realistic core layout. Its characteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas. (authors)

  16. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.

    1983-04-01

    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)

  17. Investigation of molten salt fast reactor

    International Nuclear Information System (INIS)

    Kubota, Kenichi; Konomura, Mamoru

    2002-01-01

    On survey research for practicability strategy of fast reactor (FR) (phase 1), to extract future practicability image candidates of FR from wide options, in addition to their survey and investigation objects of not only solid fuel reactors of conventional research object but also molten salt reactor as a flowing fuel reactor, investigation on concept of molten salt FR plant was carried out. As a part of the first step of the survey research for practicability strategy, a basic concept on plant centered at nuclear reactor facility using chloride molten salt reactor capable of carrying out U-Pu cycle was examined, to perform a base construction to evaluate economical potential for a practical FBR. As a result, a result could be obtained that because of inferior fuel inventory and heat transmission to those in Na cooling reactor in present knowledge, mass of reactor vessel and intermediate heat exchanger were to widely increased to expect reduction of power generation unit price even on considering cheapness of its fuel cycle cost. Therefore, at present step further investigation on concept design of the chloride molten salt reactor plant system is too early in time, and it is at a condition where basic and elementary researches aiming at upgrading of economical efficiency such as wide reduction of fuel inventory, a measure expectable for remarkable rationalization effect of reprocessing system integrating a reactor to a processing facility, and so on. (G.K.)

  18. Fabrication of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    Mishra, Sudhir; Kumar, Arun; Kutty, T.R.G.; Kamath, H.S.

    2011-01-01

    Mixed oxide (MOX) (U,Pu)O 2 , 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, PuO 2 ), where PuO 2 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 UO 2 /PuO 2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

  19. A resting bottom sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Costes, D.

    2012-01-01

    This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

  20. The integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1990-01-01

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

  1. Use of fast reactors for actinide transmutation

    International Nuclear Information System (INIS)

    1993-03-01

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

  2. Risk Management for Sodium Fast Reactors.

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  3. Fast-reactor fuel reprocessing in the United Kingdom

    International Nuclear Information System (INIS)

    Allardice, R.H.; Buck, C.; Williams, J.

    1977-01-01

    Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the United Kingdom since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium-based fast-reactor system, and the importance of establishing at an early stage fast-reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high-burnup thermal-reactor oxide fuel. The United Kingdom therefore decided to reprocess irradiated fuel from the 250MW(e) Prototype Fast Reactor (PFR) as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small-scale fully active demonstration plant has been carried out since 1972, and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste-management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant, a parallel development programme has been initiated to provide the basis for the design of a large-scale fast-reactor fuel-reprocessing plant to come into operation in the late 1980s to support the projected UK fast-reactor installation programme. The paper identifies the important differences between fast-reactor and thermal-reactor fuel-reprocessing technologies and describes some of the development work carried out in these areas for the small-scale PFR fuel-reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast-reactor fuel-reprocessing plant is outlined and the current design philosophy discussed. (author)

  4. Construction schedule management of China Experimental Fast Reactor

    International Nuclear Information System (INIS)

    Wang Yue

    2012-01-01

    China Experimental Fast Reactor (CEFR) in the first Fast Reactor in China, which is one of large project of the National High Technology Research and Development Program ('863' Program). On 21 st July 2011, CEFR had succeeded to connect to power grid, the target of construction had come true. To a large item, schedule management is one of the most important management, this paper a overall discussion about CEFR item. It has proved that the management of CEFR project is scientific, normative and high-efficiency, it will be valuable for lager Fast Reactor item and designers in interrelated field. (author)

  5. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

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

    International Nuclear Information System (INIS)

    Gregg, R.; Hesketh, K.

    2013-01-01

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

  7. Plant experience of experimental fast reactor 'Joyo'

    International Nuclear Information System (INIS)

    1982-01-01

    The experimental fast reactor ''JOYO'' installed in Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan completed its operation using the first core (called MK-I core) in December, 1981, and the works to transfer to MK-2 core have been performed since January, 1982. In this report, the experiences obtained through the construction, test and operation of ''JOYO'' over 12 years from the start of erection in 1970 to the termination of operation in 1981 are described. The contents of the report are divided into design, construction, the outline of facilities, testing, operating and maintenance experiences, and the topics on MK-I operation. As for the construction, the design changes performed before the start of manufacture or construction and the improvement and trouble restoring works implemented at the start of overall functional tests are reported. As for testing, overall functional tests, criticality test, low power test and power increasing test are described in detail. The number of test items of overall functional testing reached 266. The rated output operation of the reactor at 75 MW was performed six times in 1980 and 1981 until the termination of operation. No fuel failure was detected in MK-I operation, and the stable operation performance of the FBR was proved through MK-I operation. The topics on the MK-I operation includes natural circulation test, the measurement of total leakage rate for the containment vessel, and wear-marks which are the trace of wear due to the contact of fuel pins with the wires wound around the adjacent fuel pins, found in the post irradiation examination of fuel. (Wakatsuki, Y.)

  8. Actinide burning in the integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1993-01-01

    During the past few years, Argonne National Laboratory has been developing the integral fast reactor (IFR), an advanced liquid-metal reactor concept. In the IFR, the inherent properties of liquid-metal cooling are combined with a new metallic fuel and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics, and waste management. A key feature of the IFR concept is its unique pyroprocessing. Pyroprocessing has the potential to radically improve long-term waste management strategies by exploiting the following attributes: 1. Minor actinides accompany plutonium product stream; therefore, actinide recycling occurs naturally. Actinides, the primary source of long-term radiological toxicity, are removed from the waste stream and returned to the reactor for in situ burning, generating useful energy. 2. High-level waste volume from pyroprocessing call be reduced substantially as compared with direct disposal of spent fuel. 3. Decay heat loading in the repository can be reduced by a large factor, especially for the long-term burden. 4. Low-level waste generation is minimal. 5. Troublesome fission products, such as 99 Tc, 129 I, and 14 C, are contained and immobilized. Singly or in combination, the foregoing attributes provide important improvements in long-term waste management in terms of the ease in meeting technical performance requirements (perhaps even the feasibility of demonstrating that technical performance requirements can be met) and perhaps also in ultimate public acceptance. Actinide recycling, if successfully developed, could well help the current repository program by providing an opportunity to enhance capacity utilization and by deferring the need for future repositories. It also represents a viable technical backup option in the event unforeseen difficulties arise in the repository licensing process

  9. Radioisotopes in the primary circuit of a fast reactor

    International Nuclear Information System (INIS)

    Berlin, M.; Cauvin, M.

    1976-01-01

    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

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

    International Nuclear Information System (INIS)

    Stephens, M.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  12. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    International Nuclear Information System (INIS)

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05)

  13. First (Kick-Off) Research Coordination Meeting (RCM) of the IAEA Coordinated Research Project (CRP) on Sodium Properties and Safe Operation of Experimental Facilities in Support of the Development and Deployment of Sodium-Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    General information on recent IAEA activities in the field of fast reactors, including on-going and planned FR-related CRPs, as well as a review of the NAPRO CRP content, objectives, participants, planning and expected outcomes/deliverables, was provided. It was reminded that the NAPRO CRP is being organized in 3 main Work Packages (WP): • WP 1: collection, expert assessment and dissemination of consistent Na property data such as: surface tension, saturation vapor pressure and emissivity, thermo-dynamic behavior of ternary oxides in Na, and solubility - diffusivity of metallic impurities, etc. • WP 2: development of guidelines and best practices for Na facility design and operation, including fill and drain, purification, out-gassing prior to filling, Na storage, component handling, drying of Na piping after repair, etc. • WP 3: development of guidelines and best practices for Na facility safety, including prevention and mitigation of Na leaks, prevention and detection of Na fires, assessment of Na impact in the environment after accidental release, hydrogen hazards in cleaning facilities, etc

  14. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  15. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-01-01

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

  16. Proposed fuel cycle for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.; Walters, L.C.

    1985-01-01

    One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor

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

    International Nuclear Information System (INIS)

    Chang, Yoon I.; Lo Pinto, Pierre; Konomura, Mamoru

    2006-01-01

    The current commercial reactors based on light water technology provide 17% of the electricity worldwide owing to their reliability, safety and competitive economics. In the near term, next generation reactors are expected to be evolutionary type, taking benefits of extensive LWR experience feedbacks and further improved economics and safety provisions. For the long term, however, sustainable energy production will be required due to continuous increase of the human activities, environmental concerns such as greenhouse effect and the need of alternatives to fossil fuels as long term energy resources. Therefore, future generation commercial reactors should meet some criteria of sustainability that the current generation cannot fully satisfy. In addition to the current objectives of economics and safety, waste management, resource extension and public acceptance become other major objectives among the sustainability criteria. From this perspective, two questions can be raised: what reactor type can meet the sustainability criteria, and how to proceed to an effective deployment in harmony with the high reliability and availability of the current nuclear reactor fleet. There seems to be an international consensus that the fast spectrum reactor, notably the sodium-cooled system is most promising to meet all of the long term sustainability criteria. As for the latter, we propose a small modular fast reactor project could become a base to prepare the industrial infrastructure. The paper has the following contents: - Introduction; - SMFR project; - Core design; - Supercritical CO 2 Brayton cycle; - Near-term reference plant; - Advanced designs; - Conclusions. To summarize, the sodium-cooled fast reactor is currently recognized as the technology of choice for the long term nuclear energy expansion, but some research and development are required to optimize and validate advanced design solutions. A small modular fast reactor can satisfy some existing near-term market niche

  18. Status of fast reactor technology in China

    International Nuclear Information System (INIS)

    Xu Mi

    1992-01-01

    The paper has introduced briefly the recent news about the Chinese nuclear programme on PWR and FBR. Concerning the FFR design, some issues under consideration have been presented, including the matches between thermo-parameters of primary sodium and of steam, the arrangement of control and safety rods which correspond to first and second shut-down systems, the structure of inner vessel and the axial length of subassembly. With regard to the R and D of FBR technology, some results on sodium technology and on the cladding materials have been given in the paper. Finally, some progress and troubles on site selection for this reactor have also been outlined. (author)

  19. Control rod homogenization in heterogeneous sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Andersson, Mikael

    2016-01-01

    The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

  20. Design of unique pins for irradiation of higher actinides in a fast reactor

    International Nuclear Information System (INIS)

    Basmajian, J.A.; Birney, K.R.; Weber, E.T.; Adair, H.L.; Quinby, T.C.; Raman, S.; Butler, J.K.; Bateman, B.C.; Swanson, K.M.

    1982-03-01

    The actinides produced by transmutation reactions in nuclear reactor fuels are a significant factor in nuclear fuel burnup, transportation and reprocessing. Irradiation testing is a primary source of data of this type. A segmented pin design was developed which provides for incorporation of multiple specimens of actinide oxides for irradiation in the UK's Prototype Fast Reactor (PFR) at Dounreay Scotland. Results from irradiation of these pins will extend the basic neutronic and material irradiation behavior data for key actinide isotopes

  1. The WINCON programme - validation of fast reactor primary containment codes

    International Nuclear Information System (INIS)

    Sidoli, J.E.A.; Kendall, K.C.

    1988-01-01

    In the United Kingdom safety studies for the Commercial Demonstration Fast Reactor (CDFR) include an assessment of the capability of the primary containment in providing an adequate containment for defence against the hazards resulting from a hypothetical Whole Core Accident (WCA). The assessment is based on calculational estimates using computer codes supported by measured evidence from small-scale experiments. The hydrodynamic containment code SEURBNUK-EURDYN is capable of representing a prescribed energy release, the sodium coolant and cover gas, and the main containment and safety related internal structures. Containment loadings estimated using SEURBNUK-EURDYN are used in the structural dynamic code EURDYN-03 for the prediction of the containment response. The experiments serve two purposes, they demonstrate the response of the CDFR containment to accident loadings and provide data for the validation of the codes. This paper summarises the recently completed WINfrith CONtainment (WINCON) experiments that studied the response of specific features of current CDFR design options to WCA loadings. The codes have been applied to some of the experiments and a satisfactory prediction of the global response of the model containment is obtained. This provides confidence in the use of the codes in reactor assessments. (author)

  2. Comparison of actinides and fission products recycling scheme with the normal plutonium recycling scheme in fast reactors

    Directory of Open Access Journals (Sweden)

    Salahuddin Asif

    2013-01-01

    Full Text Available Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor. For this purpose, the Karlsruhe version of isotope generation and depletion code, KORIGEN, has been modified accordingly. An entirely novel fission product yields library for fast reactors has been created which has replaced the old KORIGEN fission products library. For the purposes of this study, the standard 26 groups data set, KFKINR, developed at Forschungszentrum Karlsruhe, Germany, has been extended by the addition of the cross-sections of 13 important actinides and 68 most important fission products. It has been confirmed that these 68 fission products constitute about 95% of the total fission products yield and about 99.5% of the total absorption due to fission products in fast reactors. The amount of fissile material required to guarantee the criticality of the reactor during recycling schemes has also been investigated. Cumulative high active waste per ton of initial heavy metal is also calculated. Results show that the recycling of actinides and fission products in fast reactors through the atomics international reduction oxidation process results in a reduction of the potential hazard of radioactive waste.

  3. Development of design procedures for fast reactors in the United Kingdom

    International Nuclear Information System (INIS)

    Rose, R.T.; Tomkins, B.; Townley, C.H.A.

    1989-01-01

    A considerable amount of research has been carried out in the United Kingdom during the past two decades to quantify the factors which control the integrity of structural components. The work which has been aimed at understanding the performance of structures at high temperature, is particularly relevant to the Fast Reactor. At the same time, because of the need to demonstrate the tolerance to defects in the low temperature as well as the high temperature components, defect assessment criteria are also of great importance. Emphasis is now being given to the development of design procedures specifically for Fast Reactors, making use of the research so far completed. The United Kingdom proposals are being integrated with those from France, Federal Republic of Germany and Italy as part of the European collaborative venture. The paper outlines the major developments which are currently in hand, and brings up to date the review of United Kingdom activities presented at Tokyo in 1986. (author)

  4. Closing nuclear fuel cycle with fast reactors: problems and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V. [Bochvar Institute - VNIINM, Moscow (Russian Federation)

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  5. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  7. Heterogeneous Recycle of Transuranics Fuels in Fast Reactors

    International Nuclear Information System (INIS)

    Hoffman, Edward; Taiwo, Temitope; Hill, Robert

    2008-01-01

    A preliminary physics evaluation of the impacts of heterogeneous recycle using Pu+Np driver and minor actinide target fuel assemblies in fast reactor cores has been performed by comparing results to those obtained for a reference homogeneous recycle core using driver assemblies containing grouped transuranic (TRU) fuel. Parametric studies are performed on the reference heterogeneous recycle core to evaluate the impacts of variations in the pre- and post-separation cooling times, target material type (uranium and non-uranium based), target amount and location, and other parameters on the system performance. This study focused on startup, single-pass cores for the purpose of quantifying impacts and also included comparisons to the option of simply storing the LWR spent nuclear fuel over a 50-year period. An evaluation of homogeneous recycle cores with elevated minor actinide contents is presented to illustrate the impact of using progressively higher TRU content on the core and transmutation performance, as a means of starting with known fuel technology with the aim of ultimately employing grouped TRU fuel in such cores. Reactivity coefficients and safety parameters are presented to indicate that the cores evaluated appear workable from a safety perspective, though more detailed safety and systems evaluations are required. (authors)

  8. Design factors affecting dynamic behaviour of fast reactor cores. UK review paper

    Energy Technology Data Exchange (ETDEWEB)

    Brindley, K W [National Nuclear Corporation Ltd., Risley, Warrington (United Kingdom); Perks, M A [United Kingdom Atomic Energy Authority, Risley, Warrington (United Kingdom)

    1982-01-01

    This paper summarises the consideration that has been given in the UK to the following factors that affect the dynamic behaviour of fast reactor cores: fuel design - Pu/u homogeneity, fuel expansion, fuel-clad gaps, uranium fraction. Structural response - CR supports, diagrid, sub-assembly bowing sodium expansion coefficients - low void cores including heterogenous cores. Calculational methods and models are outlined and some experimental results are discussed. (author)

  9. Microstructural investigations of fast reactor irradiated austenitic and ferritic-martensitic stainless steel fuel cladding

    International Nuclear Information System (INIS)

    Agueev, V.S.; Medvedeva, E.A.; Mitrofanova, N.M.; Romanueev, V.V.; Tselishev, A.V.

    1992-01-01

    Electron microscopy has been used to characterize the microstructural changes induced in advanced fast reactor fuel claddings fabricated from Cr16Ni15Mo3NbB and Cr16Ni15Mo2Mn2TiVB austenitic stainless steels in the cold worked condition and Cr13Mo2NbVB ferritic -martensitic steel following irradiation in the BOR-60, BN-350 and BN-600 fast reactors. The data are compared with the results obtained from a typical austenitic commercial cladding material, Cr16Ni15Mo3Nb, in the cold worked condition. The results reveal a beneficial effect of boron and other alloying elements in reducing void swelling in 16Cr-15Ni type austenitic steels. The high resistance of ferritic-martensitic steels to void swelling has been confirmed in the Cr13Mo2NbVB steel. (author)

  10. A Simplified Supercritical Fast Reactor with Thorium Fuel

    Directory of Open Access Journals (Sweden)

    Peng Zhang

    2014-01-01

    Full Text Available Super-Critical water-cooled Fast Reactor (SCFR is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR.

  11. 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.

  12. Analysis of the seismic response of a fast reactor core

    International Nuclear Information System (INIS)

    Martelli, A.; Maresca, G.

    1984-01-01

    This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to the PEC core elements top and applying the optimized iterative procedure between the vessel linear calculation and the non-linear ones limited to the core, which had been described in a previous report. It is demonstrated that the convergence of this procedure is very fast, similar to what obtained in the calculations of the cited report, carried out with preliminary data, and it is shown that the cited methods allow a reliable evaluation of the excitation time histories for the experimental tests in support of the seismic verification of the shutdown system and the core of a fast reactor, as well as relevant data for the experimental, structural and functional, verification of the core elements in the case of seismic loads

  13. Uranium self-shielding in fast reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Kadiroglu, O.K.; Driscoll, M.J.

    1976-03-01

    The effects of heterogeneity on resonance self-shielding are examined with particular emphasis on the blanket region of the fast breeder reactor and on its dominant reaction--capture in /sup 238/U. The results, however, apply equally well to scattering resonances, to other isotopes (fertile, fissile and structural species) and to other environments, so long as the underlying assumptions of narrow resonance theory apply. The heterogeneous resonance integral is first cast into a modified homogeneous form involving the ratio of coolant-to-fuel fluxes. A generalized correlation (useful in its own right in many other applications) is developed for this ratio, using both integral transport and collision probability theory to infer the form of correlation, and then relying upon Monte Carlo calculations to establish absolute values of the correlation coefficients. It is shown that a simple linear prescription can be developed for the flux ratio as a function of only fuel optical thickness and the fraction of the slowing-down source generated by the coolant. This in turn permitted derivation of a new equivalence theorem relating the heterogeneous self-shielding factor to the homogeneous self-shielding factor at a modified value of the background scattering cross section per absorber nucleus. A simple version of this relation is developed and used to show that heterogeneity has a negligible effect on the calculated blanket breeding ratio in fast reactors.

  14. Potential of multi-purpose liquid metallic fuelled fast reactor (MPFR) as a hydrogen production system

    International Nuclear Information System (INIS)

    Endo, H.; Ninokata, H.; Netchaev, A.; Sawada, T.

    2001-01-01

    Nuclear energy is the only effective alternative energy source to fossil fuels in the next century. Therefore future nuclear power plants should satisfy the following three requirements: i) multiple energy conversion capability with high temperature not only for electricity generation but also for hydrogen production, ii) extended siting capability so as to eliminate on-site refuelling, and iii) passive safety features. An aim of this paper is to describe the basic concept of the multi-purpose liquid metallic fuelled fast reactor system (MPFR). The MPFR introduces the U-Pu-X (X: Mn, Fe, Co) liquid metallic alloy with Ta and Ta/TaC structural materials, and satisfies all of the conditions listed above based on the following characteristics of the liquid metallic fuel: high temperature operation between 650 deg C (sodium-cooled system) and 1 200 deg C (lead-cooled system), a core lifetime of 15-30 years without radiation damage of fuel materials, and enhanced passive safety by the thermal expansion of liquid fuel and the avoidance of re-criticality due to local core fuel dispersion at fuel failure events. (authors)

  15. Status of fast reactor development in India. April 1998 - March 1999

    International Nuclear Information System (INIS)

    Lee, S.M.

    1999-01-01

    Electricity growth rate in India in 1998-99 improved compared to the previous year and the installed electric capacity reached 93.25 GWe. The thermal nuclear power plants performed very well with average capacity factor of over 72%. The Kalpakkam Reprocessing Plant was commissioned. FBTR was operated at various power levels and a peak fuel burn-up of 49000 MWd/t achieved. Test irradiation of Zr-Nb was undertaken in FBTR for the PHWR programme. Refurbishing of the plant included new state of the art neutronic channels. Detailed design of PFBR was continued. The review of the chapters of the PSAR by an IGCAR Internal Safety Committee and by the AERB PFBR-Project Design Safety Committee was continued. Work on Environmental Impact Assessment Report, for obtaining clearance from the concerned environmental authorities, for the project has been started Technology development for PFBR included core subassemblies, main vessel, inner vessel, IHX, steam generator, roof slab, drive mechanism, control plug etc. Indigenous manufacture of raw materials has been also taken up. R and D in reactor physics, shielding, engineering development, instrumentation, thermal hydraulics, structural mechanics, metallurgy, non-destructive examination, chemistry, reprocessing and safety was continued. These include cover gas heat and mass transfer, SA hydraulic tests, thermal striping studies, fuel development for PFBR, corrosion and material property studies on steels, PIE of FBTR fuel and developments for the pilot plant for fast reactor fuel reprocessing. (author)

  16. Status of fast reactor development in India. April 1998 - March 1999

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S M [Safety Research, Health Physics, Information Services, Instrumentation and Electronics Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)

    1999-07-01

    Electricity growth rate in India in 1998-99 improved compared to the previous year and the installed electric capacity reached 93.25 GWe. The thermal nuclear power plants performed very well with average capacity factor of over 72%. The Kalpakkam Reprocessing Plant was commissioned. FBTR was operated at various power levels and a peak fuel burn-up of 49000 MWd/t achieved. Test irradiation of Zr-Nb was undertaken in FBTR for the PHWR programme. Refurbishing of the plant included new state of the art neutronic channels. Detailed design of PFBR was continued. The review of the chapters of the PSAR by an IGCAR Internal Safety Committee and by the AERB PFBR-Project Design Safety Committee was continued. Work on Environmental Impact Assessment Report, for obtaining clearance from the concerned environmental authorities, for the project has been started Technology development for PFBR included core subassemblies, main vessel, inner vessel, IHX, steam generator, roof slab, drive mechanism, control plug etc. Indigenous manufacture of raw materials has been also taken up. R and D in reactor physics, shielding, engineering development, instrumentation, thermal hydraulics, structural mechanics, metallurgy, non-destructive examination, chemistry, reprocessing and safety was continued. These include cover gas heat and mass transfer, SA hydraulic tests, thermal striping studies, fuel development for PFBR, corrosion and material property studies on steels, PIE of FBTR fuel and developments for the pilot plant for fast reactor fuel reprocessing. (author)

  17. European Fast Reactor IWGFR/FRCC-report. A review of the collaborative programme on the European Fast Reactor (EFR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    The design work for the 1500 MWe European Fast Reactor EFR was started in 1988. Two years during phase 1 were devoted to the concept design; the subsequent concept validation phase 2 will last until March 1993. In autumn 1991 the 'concept design '91, CD91, was put forward; its major design features and the R and D support are described briefly together with the organisational structures. The European Fast Reactor Utilities Group 'EFRUG' presently comprises EdF (France), ENEL (Italy), Nuclear Electric (UK) and Bayernwerk, PreulsenElektra and RWE (Germany). For design and construction of EFR the group 'EFR Associates' is responsible, combining the companies Siemens (formerly Interatom, Germany), NNC Ltd. (UK) and Framatome/Div. Novatome (France). The necessary R and D support is given by CEA (France), UKAEA (UK) and KfK/Siemens (Germany). The R and D work is executed in the various national research centres ranging from Dounreay via Bensberg and Karlsruhe to Cadarache. The design work is done at Bensberg, Lyon and Risley. The present programme of design work extends to early 1993 and is aimed at producing a detailed consistent design for the nuclear part of the plant and a non site specific safety report. By that date the basic feasibility of the main design features will have been underwritten by the joint R and D programme and there will be an informal assessment of the general licensibility of the concept by the Ad Hoc Safety Club. In follow-up the utilities will then be in the position to decide whether to proceed with the next steps. The key issues of this phase will include the specification of the plant, the siting, the detailed engineering, licensing with possibly a public enquiry and the question of ownership and financing. In the international arena the collaboration with USSR is proceeding well on the basis of an USSR-Europe Agreement from January 8, 1991; it foresees review and specialists meetings in the field of fast breeder research. On the occasion of

  18. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  19. Prediction of degradation and fracture of structural materials

    International Nuclear Information System (INIS)

    Tomkins, B.

    1992-01-01

    Prediction of materials performance in an engineering integrity context requires the underpinning of predictive modelling tuned by inputs from design, fabrication, operating experience, and laboratory testing. In this regard, in addition to fracture resistance four important areas of time dependent degradation are considered - mechanical, environmental, irradiation and thermal. The status of prediction of materials performance is discussed in relation to a number of important components such as LWR reactor pressure vessels and steam generators, and Fast Reactor high temperature structures. In each case the role of materials modelling is examined and the balance of factors which contribute to the overall prediction of component integrity/reliability noted. Structural integrity arguments must follow a clear strategy if the required level of confidence is to be established. Various strategies and their evolution are discussed. (author)

  20. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Juutilainen, Pauli

    2008-01-01

    Motivated by the development of fast reactors the behaviour of the Japanese experimental fast reactor Joyo is simulated with two Monte Carlo codes: Monte Carlo NParticle (MCNP) and Probabilistic Scattering Game (PSG). The simulations are based on the benchmark study 'Japan's Experimental Fast Reactor Joyo MKI core: Sodium-Cooled Uranium-Plutonium Mixed Oxide Fueled Fast Core Surrounded by UO 2 Blanket'. The study is focused on the criticality of the reactor, control rod worth, sodium void reactivity and isothermal temperature coefficient of the reactor. These features are calculated by applying both homogeneous and heterogeneous reactor core models that are built according to the benchmark instructions. The results of the two models obtained by the two codes are compared with each other and especially with the experimental results presented in the benchmark. (author)

  2. A review of the Italian fast reactor programme

    International Nuclear Information System (INIS)

    Pierantoni, F.; Tavoni, R.

    1984-01-01

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

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

    International Nuclear Information System (INIS)

    Ding Xiaoming

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Vrban, B.

    2015-01-01

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

  5. Sodium fires at fast reactors: RF status report

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  6. A review of fast reactor progress in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Tomabechi, K [Power Reactor and Nuclear Fuel Development Corporation, Tokyo (Japan)

    1978-07-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  8. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  9. Upgrading program of the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Yoshida, A.; Yogo, S.

    2001-01-01

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

  10. 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)

  11. Status of national programmes on fast reactors in Korea

    International Nuclear Information System (INIS)

    Kim, Y.I.; Hahn, D.

    2002-01-01

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

  12. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  13. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  14. On the extension of the analytic nodal diffusion solver ANDES to sodium fast reactors

    International Nuclear Information System (INIS)

    Ochoa, R.; Herrero, J.J.; Garcia-Herranz, N.

    2011-01-01

    Within the framework of the Collaborative Project for a European Sodium Fast Reactor, the reactor physics group at UPM is working on the extension of its in-house multi-scale advanced deterministic code COBAYA3 to Sodium Fast Reactors (SFR). COBAYA3 is a 3D multigroup neutron kinetics diffusion code that can be used either as a pin-by-pin code or as a stand-alone nodal code by using the analytic nodal diffusion solver ANDES. It is coupled with thermal-hydraulics codes such as COBRA-TF and FLICA, allowing transient analysis of LWR at both fine-mesh and coarse-mesh scales. In order to enable also 3D pin-by-pin and nodal coupled NK-TH simulations of SFR, different developments are in progress. This paper presents the first steps towards the application of COBAYA3 to this type of reactors. ANDES solver, already extended to triangular-Z geometry, has been applied to fast reactor steady-state calculations. The required cross section libraries were generated with ERANOS code for several configurations. Here some of the limitations encountered when attempting to apply the Analytical Coarse Mesh Finite Difference (ACMFD) method - implemented inside ANDES - to fast reactor calculations are discussed and the sensitivity of the method to the energy-group structure is studied. In order to reinforce some of the conclusions obtained two calculations are presented. The first one involves a 3D mini-core model in 33 groups, where the ANDES solver presents several issues. And secondly, a benchmark from the NEA for a small 3D FBR in hexagonal-Z geometry in 4 energy groups is used to verify the good convergence of the code in a few-energy-group structure. (author)

  15. Time dependent design curves for a high nitrogen grade of 316LN stainless steel for fast reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Ganesh Kumar, J.; Ganesan, V.; Laha, K.; Mathew, M.D., E-mail: mathew@igcar.gov.in

    2013-12-15

    Highlights: • 316LN SS is an important high temperature structural material for sodium cooled fast reactors. • Creep strength of 316LN SS has been increased substantially by increasing the nitrogen content. • Creep design curves based on RCC-MR code procedures have been generated for this new material. • 100,000 h allowable stress at 600 °C increased by more than 40% as a result of doubling the nitrogen content in the steel. - Abstract: Type 316L(N) stainless steel (SS) containing 0.06–0.08 wt.% nitrogen is the major material for reactor assembly components of sodium cooled fast reactors (SFRs). With a view to increase the design life of SFRs to 60 years from the current life of 40 years, studies are being carried out to improve the high temperature creep and low cycle fatigue properties of 316LN SS by increasing the nitrogen content above 0.08 wt.%. In this investigation, the creep properties of a high nitrogen grade of 316LN SS containing 0.14 wt.% nitrogen have been studied. Creep tests were carried out at 550 °C, 600 °C and 650 °C at various stress levels in the range of 140–350 MPa. Creep strength was found to be significantly improved by doubling the nitrogen content in this steel. The maximum rupture life in these tests was 33,000 h. The creep data has been analyzed according to RCC-MR nuclear code procedures in order to generate the creep design curves for the high nitrogen grade of 316LN SS. Allowable stress for 100,000 h at 600 °C increased by more than 38% as a result of doubling the nitrogen content in the steel.

  16. Current status and perspective of advanced loop type fast reactor in fast reactor cycle technology development project

    International Nuclear Information System (INIS)

    Niwa, Hajime; Aoto, Kazumi; Morishita, Masaki

    2007-01-01

    After selecting the combination of the sodium-cooled fast reactor (SFR) with oxide fuel, the advanced aqueous reprocessing and the simplified pelletizing fuel fabrication as the most promising concept of FR cycle system, 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was finalized in 2006. Instead, a new project, Fast Reactor Cycle Technology Development Project (FaCT Project) was launched in Japan focusing on development of the selected concepts. This paper describes the current status and perspective of the advanced loop type SFR system in the FaCT Project, especially on the design requirements, current design as well as the related innovative technologies together with the development road-map. Some considerations on advantages of the advanced loop type design are also described. (authors)

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

    International Nuclear Information System (INIS)

    1998-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-01

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

  19. NEA Activities in Preserving, Evaluating and Applying Data from Fast Reactor Experiments

    International Nuclear Information System (INIS)

    Gulliford, N.T.; Cornet, S.M.; Hill, I.; Yamaji, A.

    2015-01-01

    also reported, along with an overview of some other NEA nuclear science activities related to fast reactors in the fuel cycle, materials and nuclear data areas. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    2004-08-01

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

  2. European commission - 7th framework programme. The collaborative project on European sodium fast reactor (CP ESFR)

    International Nuclear Information System (INIS)

    Fiorini, G.L.

    2009-01-01

    The paper summarizes the key characteristics of the four years large Collaborative Project on European Sodium Fast Reactor (CP ESFR - 2009-2012); the CP ESFR follows the 6th FP project named 'Roadmap for a European Innovative SOdium cooled FAst Reactor - EISOFAR' further identifying, organizing and implementing a significant part of the needed R and D effort. The CP ESFR merges the contribution of 25 european partners; it will be realized under the aegis of the 7th FP under the Area - Advanced Nuclear Systems with a refund from the European Commission of 5.8 M euro (11.55 M euro total budget). It will be a key component of the European Sustainable Nuclear Energy Technology Platform (SNE TP) and its Strategic Research Agenda (SRA). The inputs for the project are the key research goals for fourth generation of European sodium cooled fast reactors which can be summarized as follow: an improved safety with in particular the achievement of a robust architecture vis a vis of abnormal situations and the robustness of the safety demonstrations; the guarantee of a financial risk comparable to that of the other means of energy production; a flexible and robust management of the nuclear materials and especially the waste reduction through the Minor Actinides burning. (author)

  3. The collaborative project on European sodium fast reactor (CP ESFR project)

    International Nuclear Information System (INIS)

    Fiorini, Gian-Luigi

    2010-01-01

    The paper summarizes the key characteristics of the four years large Collaborative Project on European Sodium Fast Reactor (CP ESFR - 2009-2012); the CP ESFR follows the 6th FP project named 'Roadmap for a European Innovative SOdium cooled FAst Reactor - EISOFAR' further identifying, organizing and implementing a significant part of the needed R and D effort. The paper also gives insights concerning the so called 'working horses' cores and systems which are provided by CEA and AREVA and that will be used as a basis to test the performances and assess the pertinence of innovative solutions. The CP ESFR merges the contribution of 25 European partners (EU + CH); it will be performed under the aegis of the 7th Euratom FP under the Area - Advanced Nuclear Systems with a refund from the European Commission. It will be a key component of the European Sustainable Nuclear Energy Technology Platform (SNE TP) and its Strategic Research Agenda (SRA). The inputs for the project are the key research goals for fourth generation of European sodium cooled fast reactors which can be summarized as follows: an improved safety with in particular the achievement of a robust architecture vis-a-vis of abnormal situations and the robustness of the safety demonstrations; the guarantee of a financial risk similar to that of the other means of energy production; a flexible and robust management of nuclear materials and especially waste reduction through Minor Actinides burning

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

    International Nuclear Information System (INIS)

    Beddingfield, David H.; Lafleur, Adrienne M.

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Orlov, V.V.

    2001-01-01

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

  6. Study on an innovative fast reactor utilizing hydride neutron absorber - Final report of phase I study

    International Nuclear Information System (INIS)

    Konashi, K.; Iwasaki, T.; Itoh, K.; Hirai, M.; Sato, J.; Kurosaki, K.; Suzuki, A.; Matsumura, Y.; Abe, S.

    2010-01-01

    These days, the demand to use nuclear resources efficiently is growing for long-term energy supply and also for solving the green house problem. It is indispensable to develop technologies to reduce environmental load with the nuclear energy supply for sustainable development of human beings. In this regard, the development of the fast breeder reactor (FBR) is preferable to utilize nuclear resources effectively and also to burn minor actinides which possess very long toxicity for more than thousands years if they are not extinguished. As one of the FBR developing works in Japan this phase I study started in 2006 to introduce hafnium (Hf) hydride and Gadolinium-Zirconium (Gd-Zr) hydride as new control materials in FBR. By adopting them, the FBR core control technology is improved by two ways. One is extension of control rod life time by using long life Hf hydride which leads to reduce the fabrication and disposal cost and the other is reduction of the excess reactivity by adopting Gd-Zr hydride which leads to reduce the number of control rods and simplifies the core upper structure. This three year study was successfully completed and the following results were obtained. The core design was performed to examine the applicability of the Hf hydride absorber to Japanese Sodium Fast Reactor (JSFR) and it is clarified that the control rod life time can be prolonged to 6 years by adopting Hf hydride and the excess reactivity of the beginning of the core cycle can be reduced to half and the number of the control rods is also reduced to half by using the Gd-Zr hydride burnable poison. The safety analyses also certified that the core safety can be maintained with the same reliability of JSFR Hf hydride and Gd-Zr hydride pellets were fabricated in good manner and their basic features for design use were measured by using the latest devices such as SEM-EDX. In order to reduce the hydrogen transfer through the stainless steel cladding a new technique which shares calorizing

  7. Benchmark on traveling wave fast reactor with negative reactivity feedback obtained with MCNPX code

    International Nuclear Information System (INIS)

    Gann, V.V.; Gann, A.V.

    2012-01-01

    This paper presents results of computer simulations of traveling wave fast reactor with negative reactivity feedback. The results were obtained using MCNPX code combined with CINDER90 subroutine for depletion calculations. We considered 1-D model of TWR containing 4 m long core made of mixture of 66 at. % 238 U and 34 at. % 10 B. Ignitor made of 235 U was located in the center of the core. Boron was included as imitator of structural in-core materials and coolant. Negative reactivity feedback was adjusted to reactor power of 500 MW. In this case two burning waves originated from the igniter and travel to the ends of the core during the following 40 years; coefficient of utilization of 238 U reached 80 %. Distribution of specific power in traveling wave, isotope concentration of fission products and actinides, neutron flux, fast neutron spectrum, specific activity were calculated. Data of the computer simulation is in qualitative agreement with theoretical results obtained in slow burning wave approximation

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

    International Nuclear Information System (INIS)

    Poplavski, V.M.; Ashurko, Y.M.; Zverev, K.V.; Sarayev, O.M.; Oshkanov, N.N.; Korol'kov, A.S.

    1999-01-01

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

  9. 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.

  10. Development of a three dimension multi-physics code for molten salt fast reactor

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2014-01-01

    Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-07-01

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

  13. Technology development of fast reactor fuel reprocessing technology in India

    International Nuclear Information System (INIS)

    Natarajan, R.; Raj, Baldev

    2009-01-01

    India is committed to the large scale induction of fast breeder reactors beginning with the construction of 500 MWe Prototype Fast Breeder Reactor, PFBR. Closed fuel cycle is a prerequisite for the success of the fast reactors to reduce the external dependence of the fuel. In the Indian context, spent fuel reprocessing, with as low as possible out of pile fissile inventory, is another important requirement for increasing the share in power generation through nuclear route as early as possible. The development of this complex technology is being carried out in four phases, the first phase being the developmental phase, in which major R and D issues are addressed, while the second phase is the design, construction and operation of a pilot plant, called CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell. The third phase is the construction and operation of Demonstration of Fast Reactor Fuel Reprocessing Plant (DFRP) which will provide experience in fast reactor fuel reprocessing with high availability factors and plant throughput. The design, construction and operation of the commercial plant (FRP) for reprocessing of PFBR fuel is the fourth phase, which will provide the requisite confidence for the large scale induction of fast reactors

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

    International Nuclear Information System (INIS)

    Hammer, P.

    1980-09-01

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

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

    International Nuclear Information System (INIS)

    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 ampersand D

  16. Linear and nonlinear stability analysis, associated to experimental fast reactors

    International Nuclear Information System (INIS)

    Amorim, E.S. do; Moura Neto, C. de; Rosa, M.A.P.

    1980-07-01

    Phenomena associated to the physics of fast neutrons were analysed by linear and nonlinear Kinetics with arbitrary feedback. The theoretical foundations of linear kinetics and transfer functions aiming at the analysis of fast reactors stability, are established. These stability conditions were analitically proposed and investigated by digital and analogic programs. (E.G.) [pt

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

    International Nuclear Information System (INIS)

    Pierantoni, F.

    1978-01-01

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

  18. Integral Fast Reactor Program. Annual progress report, FY 1992

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-06-01

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

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

    International Nuclear Information System (INIS)

    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 ampersand D

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

    International Nuclear Information System (INIS)

    Donaldson, D.M.; Bray, J.A.; Newson, I.H.

    1978-01-01

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

  1. International standardization of safety requirements for fast reactors

    International Nuclear Information System (INIS)

    2011-06-01

    Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Donaldson, D M [FRDD, UKAEA, Risley (United Kingdom); Bray, J A; Newson, I H [UKAEA, Dounreay Nuclear Power Establishment, Thurso (United Kingdom)

    1978-08-01

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

  3. Dose rate analyses for fast reactor fuel manufacturers

    International Nuclear Information System (INIS)

    Smith, R.C.; Strode, J.N.; Brackenbush, L.W.; Faust, L.G.

    1976-01-01

    An early appraisal of the radiation exposure situation in the fabrication of plutonium enriched mixed-oxide fuels for fast reactors is presented. Radiation data are presented on fuel process operations measured under actual operating conditions using plutonium containing up to 19 wt percent 240 Pu

  4. Fuel Behavior Modeling Issues Associated with Future Fast Reactor Systems

    International Nuclear Information System (INIS)

    Yacout, A.M.; Hofman, G.L.; Lambert, J.D.B.; Kim, Y.S.

    2007-01-01

    Major issues of concern related to advanced fast reactor fuel behavior are discussed here with focus on phenomena that are encountered during irradiation of metallic fuel elements. Identification of those issues is part of an advanced fuel simulation effort that aims at improving fuel design and reducing reliance on conventional approach of design by experiment which is both time and resource consuming. (authors)

  5. 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.

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

    International Nuclear Information System (INIS)

    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

  7. Review of fast reactor activities in India (1982-83)

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1983-01-01

    A review of fast reactor activities in India in 1982-1983 is given. One stage of construction of Fast Breeder Test Reactor (FBTR) is briefly described. The emphasis is on design studies for the 500 MWe Prototype Fast Breeder Reactor (PFBR). The main features of this design are introduced

  8. U.S. activities related to fast reactors and ADS

    International Nuclear Information System (INIS)

    Finck, Phillip J.

    2001-01-01

    , lower cost, and improved safety and proliferation resistance; Innovative nuclear plant design, manufacturing, construction, operation, maintenance and decommissioning technologies; Advanced nuclear fuels; and Relevant areas of fundamental science. In 2000, the US Fast Reactor and ADS programs include the EBR-II Electrometallurgical Treatment program, FFTF activities in view of a decision whether to restart or deactivate the facility, several NERI awards, and the Advanced Accelerators Applications program. A 10 year R and D plan concentrated on defining the key technologies to be used for transmutation of nuclear waste (plutonium, minor actinides and long lived fission products). This activity includes fuels development and performance testing, development and testing of dedicated separation technologies, design of transmutation systems, and materials and physics research. It is expected that after 10 years, suitable technologies will have been demonstrated for practical waste transmutation demonstration. The Accelerator Driven Test Facility (ADTF) will be built over 10 years. It will consist of a large proton linear accelerator (600MeV, 13mA), coupled to two testing stations; the Target and Multiplier Test Station will consist of a large (8MW) spallation target surrounded by test loops for materials and small amounts of fuels; the Sub Critical Multiplier (SCM) will consist of a 100MW subcritical fast reactor driven by a 4MW spallation target and will be used to demonstrate the safe and efficient operations of accelerator driven systems, and will serve to irradiate experimental fuels. The ADTF will serve as the principal test station for a Proof of Performance series of tests, to demonstrate the safety and operations of Accelerator Driven Systems, and to demonstrate efficient transmutation and recycling of minor actinides and long lived fission products

  9. Safety considerations in the fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Baker, A.R.; Burton, W.R.; Taylor, H.A.

    1977-01-01

    The fuel cycle safety problems for fast reactors, as compared with thermal reactors, are enhanced by the higher fissile content and heat rating of the fuel. Additionally recycling leads to the build up of substantial isotopes which contribute to the alpha and neutron hazards. The plutonium arisings in a nuclear power reactor programme extending into the next century are discussed. A requirement is to be able to return the product plutonium to a reactor about 9 months after the end of irradiation and it is anticipated that progress will be made slowly towards this fuel cycle, having regard to the necessity for maintaining safe and reliable operations. Consideration of the steps in the fuel cycle has indicated that it will be best to store the irradiated fuel on the reactor sites while I131 decays and decay heat falls before transporting and a suitable transport flask is being developed. Reprocessing development work is aimed at the key area of fuel breakdown, the inter-relation of the fuel characteristics on the dissolution of the plutonium and a solvent extract cycle leading to a product suitable for a co-located fabrication plant. Because of the high activity of recycled fuel it is considered that fabrication must move to a fully remote operation as is already the case for reprocessing, and a gel precipitation process producing a vibro compacted fuel is under development for this purpose. The waste streams from the processing plants must be minimised, processed for recovery of plutonium where applicable and then conditioned so that the final products released from the processing cycle are acceptable for ultimate disposal. The safety aspects reviewed cover protection of operators, containment of radioactive materials, criticality and regulation of discharges to the environment

  10. Investigation of small and modular-sized fast reactor

    International Nuclear Information System (INIS)

    Kubota, Kenichi; Kawasaki, Nobuchika; Umetsu, Yoichiro; Akatsu, Minoru; Kasai, Shigeo; Konomura, Mamoru; Ichimiya, Masakazu

    2000-06-01

    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)

  11. Modelling of an ULOF transient in a sodium fast reactor

    International Nuclear Information System (INIS)

    Droin, Jean-Baptiste

    2016-01-01

    particular attention is paid to stabilize boiling occurrence which leads to minimize severe accident consequences.The phenomena occurring during the various ULOF transients are modelled in accordance to the level of details required to catch all the possible bifurcations of the transient. The tool coupled different (2D, 1D and 0D) models of thermics, thermo-hydraulics, core degradation (material melting and motions) and neutronics. The assumptions associated to these models are highlighted, discussed and validated. The physical tool capability of simulating the various realistic ULOF transients (without boiling, with stabilized boiling or flow excursion after boiling) is demonstrated by comparisons to experimental results (GR19, SCARABEE experiments) and to mechanistic simulations (CATHARE2 and SIMMER III).Parametric studies are then carried out on two variables: the fuel burn-up and the model of neutronic feedbacks. They underline the important influence of these parameters on the transient and the final core state. Finally, a preliminary sensitivity analysis (2000 simulations) is performed on 26 uncertain parameters (linked to initial core configuration, accident features, model uncertainties and radial nodalization). The variability of the final core state is underlined and quantified; only around 25% of cases lead to core degradation. The main influent parameters on transient phenomena are also identified, enabling to prioritize core design and safety studies. In the future, this tool will be used for safety-informed design and stability analyses of fast reactor systems, allowing to emphasize the main dominant phenomena and trends of significance for safety assessment. (author) [fr

  12. Solution treatment of fast reactor claddings

    International Nuclear Information System (INIS)

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro

    1974-01-01

    The fuel cladding tubes for Joyo (experimental FBR) are required to be a material corresponding to AISI Type 316 and cold-rolled after solution treatment. It is necessary to have no precipitation of carbide and to make the grain size smaller than ASTM No.6. It is very difficult to obtain the fine grains without the precipitation, however. In this connection, the behavior of carbide solution at high temperature and the annealing behavior of the material cold-worked and solution-treated were studied. The following matters are described: the solid solubility line of AISI Type 316; the behavior of carbide at solution treatment temperature; and the annealing behavior of the cold-worked material. (Mori, K.)

  13. Recent development of radioanalytical method at IBR-2 pulsed fast reactor of the JINR

    International Nuclear Information System (INIS)

    Nazarov, V.M.; Pavlov, S.S.; Herrera, E.

    1991-01-01

    The experience of the use of radioanalytical methods, including NAA at IBR-2 pilsed fast reactor of the JINR, is discussed. Physical and technical parameters of the experimental installation designed for NAA and radiography are given. The detailed examples of the application of resonance neutrons to the control of the environment in the geology of oil, in multi-element analysis of food products and superpure materials as well as in nuclear physics are reviewed. The works on the application of the neutron isotopes sources for express determination of nitrogen content in original and synthetic materials are introduced. 7 refs.; 8 figs.; 3 tabs

  14. Measurement of the physics properties of gas-cooled fast reactors in the zero energy reactor PROTEUS and analysis of the results

    International Nuclear Information System (INIS)

    Richmond, R.

    1982-12-01

    The main aim of the fast reactor physics measurements carried out in the zero energy reactor PROTEUS was to check the performance of data sets and calculation methods used in the design of fast breeder reactors. This allowed the accuracy of the power reactor calculations to be determined and enabled an assessment to be made of whether this accuracy would be sufficient to allow the design, construction and licensing of the GCFR power reactor. In order to carry out the physics measurements an existing zero energy reactor was converted to a form in which a central fast reactor lattice was surrounded by thermal zones to drive the reactor critical. One of the most important measuring techniques used to check the performance of data sets and calculation methods was the determination of reaction rate ratios and, by using an appropriate range of nuclides, it was possible to obtain a detailed picture covering 70% of reactions taking place in the central part of the fast reactor zone and with an accuracy of +-1.5% in a typical ratio. A further technique used during the work on GCFR-PROTEUS was the measurement of neutron spectrum which was carried out in a wide range of environments and, in the later stages of the work, covered the energy range from 9 keV to 2.3 MeV. These measurements, in particular, indicated significant errors in the FGL4 scattering cross-sections. A third technique, which was developed to a high degree of accuracy, was the measurement of reactivity worths. This was used in measurements of the worths of small samples and also in the application of the null reactivity technique to determine k-infinity and hence the absorption cross-sections of reactor structural materials. (Auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Balz, W [Commission of the European Communities, Brussels (Belgium)

    1981-05-01

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

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

    International Nuclear Information System (INIS)

    Balz, W.

    1981-01-01

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

  17. Uranium-plutonium fuel for fast reactors

    International Nuclear Information System (INIS)

    Antipov, S.A.; Astafiev, V.A.; Clouchenkov, A.E.; Gustchin, K.I.; Menshikova, T.S.

    1996-01-01

    Technology was established for fabrication of MOX fuel pellets from co-precipitated and mechanically blended mixed oxides. Both processes ensure the homogeneous structure of pellets readily dissolvable in nitric acid upon reprocessing. In order to increase the plutonium charge in a reactor-burner a process was tested for producing MOX fuel with higher content of plutonium and an inert diluent. It was shown that it is feasible to produce fuel having homogeneous structure and the content of plutonium up to 45% mass

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

    International Nuclear Information System (INIS)

    Wu Xingman

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  20. Fast Reactor Knowledge Preservation Efforts. An Overview

    International Nuclear Information System (INIS)

    Grandy, Christopher

    2013-01-01

    • ARC-AFR Program is involved in a number of knowledge preservation activities; • Recovery of Information from EBR-II, FFTF, and TREAT is very important; • Recovery of Information from Office of Scientific and Technical Information (OSTI) and conversion to electronic format; • Organizing some data into electronic databases – EBR-II Plant Testing Data, FFTF Plant Testing Data, TREAT Test Data, SFR Fuels and Materials Irradiation Data, etc.; • Information is being used to support existing U.S. SFR programs along with international programs such as the IAEA CRP EBR-II Safety Benchmark

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

    International Nuclear Information System (INIS)

    1998-01-01

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

  2. Application of artificial intelligence to the detection and diagnosis of core faults in fast reactors

    International Nuclear Information System (INIS)

    Le Guillou, G.; Berlin, C.; Parcy, J.P.

    1983-01-01

    An automatic diagnostic system designed specially for fuel assembly accidents in fast reactors of the French type is described. It has an ''expert system'' structure and, at the present stage of the project, is operating on a laboratory scale in an experimental unit (Solar 16/65). The software aspects of the development of such systems (structure and formulation of rules, language, etc.) are presented in detail. Various simulation tests are outlined. ''Real-time'' application of the system will begin in 1983 on the Phenix reactor. (author)

  3. 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.

  4. Effect of neutron anisotropic scattering in fast reactor analysis

    International Nuclear Information System (INIS)

    Chiba, Gou

    2004-01-01

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

  5. Development of fast reactor metal fuels containing minor actinides

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  7. Trial visualization of fast reactor design knowledge

    International Nuclear Information System (INIS)

    Yoshikawa, Shinji; Minami, Masaki; Takahashi, Tadao

    2011-01-01

    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)

  8. Materials with structural hierarchy

    Science.gov (United States)

    Lakes, Roderic

    1993-01-01

    The role of structural hierarchy in determining bulk material properties is examined. Dense hierarchical materials are discussed, including composites and polycrystals, polymers, and biological materials. Hierarchical cellular materials are considered, including cellular solids and the prediction of strength and stiffness in hierarchical cellular materials.

  9. Design characteristics of zero power fast reactor Lasta

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  10. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

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

    2010-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bramman, J I [International Relations Department, United Kingdom Atomic Energy Authority, Risley, Warrington (United Kingdom)

    1992-07-01

    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.

  12. Software for selection of optimal layouts of fast reactors

    International Nuclear Information System (INIS)

    Geraskin, N.I.; Kuz'min, A.M.; Morin, D.V.

    1983-01-01

    A complex program for the calculation and optimization of a two-dimensional cylindrical fast reactor consisting of two axial layers and having up to 10 zones of different compositions in each layer is described. Search for optimal parameters is performed by the successive linearization method based on the small perturbation theory and linear programming. The complex program is written for the BESM-6 computer in the FORTRAN language

  13. Specialists' meeting on fast reactor cover gas purification

    International Nuclear Information System (INIS)

    1987-01-01

    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

  14. Specialists' meeting on fast reactor cover gas purification

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1987-07-01

    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. Teaching Sodium Fast Reactors in CEA-INSTN

    International Nuclear Information System (INIS)

    Dufour, Ph.; Latgé, C.; Gicquel, L.

    2013-01-01

    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

  16. Analyses of hypothetical FCI's in a fast reactor

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Martin, F.J.; Niccoli, L.G.

    1981-01-01

    Parametric analyses using the SIMMER code were performed to evaluate the potential for a severe recriticality from a pressure-driven recompaction caused by an energetic FCI during the transition phase of a hypothetical accident in a fast reactor. For realistic and reasonable estimates for the assumed accident conditions, a severe recriticality was not predicted. The conditions under which a severe recriticality would be obtained or averted were identified. 10 figures, 2 tables

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

    Energy Technology Data Exchange (ETDEWEB)

    De Luca, B [CNEN-RIT/MAT - Laboratorio Sviluppo Processi - C.S.N. Cassacia, Rome (Italy); Labanti, V [CNEN-DRV, Bologna (Italy); Mennucci, M [NIRA, Genoa (Italy)

    1978-08-01

    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.

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

    International Nuclear Information System (INIS)

    Waintraub, M.

    1986-01-01

    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)

  19. Actinide recycle potential in the integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. In the IFR pyroprocessing, minor actinides accompany plutonium product stream, and therefore, actinide recycle occurs naturally. The fast neutron spectrum of the IFR makes it an ideal actinide burner, as well. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and potential implications on long-term waste management

  20. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

    2013-01-01

    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