Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

International Nuclear Information System (INIS)

Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Hae Yong; Chang, Won Pyo; Seok, Su Dong; Lee, Yong Bum

2010-02-01

A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

International Nuclear Information System (INIS)

Fernandez, A.; McGinley, J.; Somers, J.

2008-01-01

The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

Energy Technology Data Exchange (ETDEWEB)

Fernandez, A.; McGinley, J.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements P.O.Box 2340, Karlsruhe, D-76125 (Germany)

2008-07-01

The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors

International Nuclear Information System (INIS)

Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

2008-01-01

Many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important criterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals

A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.

Energy Technology Data Exchange (ETDEWEB)

Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

2008-04-23

many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

Generation-IV nuclear reactors, SFR concept

International Nuclear Information System (INIS)

Dufour, P.

2010-01-01

In this presentation author deals with development of sodium-cooled fast reactors and lead-cooled fast reactors. He concluded that: - SFR is a proved concept that has never achieved industrial deployment; - The GEN IV objectives need to reconsider the design of both the core and the reactor design : innovations are being analysed; Future design will benefit from considerable feedback of design, licensing, construction and operation of PX, SPX, etc.

Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

Science.gov (United States)

Murty, K. L.; Charit, I.

2008-12-01

Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

Fuel research for subcritical and critical GEN-IV systems cooled by heavy liquid metal

International Nuclear Information System (INIS)

Sobolev, V.; Verwerft, M.

2009-01-01

The participation of the Belgian Nuclear Research Centre SCK-CEN in the worldwide GEN-IV research can be considered as an opportunity. Today's GEN-IV research at SCK-CEN is mainly driven by the interests of the project MYRRHA (Multipurpose hYbrid Research Reactor for High-tech Applications). The main goal of this project is to build at SCK-CEN in Mol a new generation fast spectrum, subcritical, research and materials testing reactor MYRRHA driven by a high-energy proton accelerator. This GEN-IV MTR is cooled by heavy liquid metal (Pb-Bi) and will be used for the ADS concept demonstration, testing and qualification of new fuels, transmutation targets and innovative materials. On the European scale, MYRRHA is integrated in the Euratom FP6 Integrated Project (IP) EUROTRANS (EUROpean research programme for TRANSmutation of high level nuclear waste in an accelerator driven system), as the small-scale experimental machine for transmutation demonstration called XT-ADS. Last but not least, this experimental facility will also demonstrate the technological feasibility of the LFR (Lead-cooled Fast Reactor) GEN-IV concept; in EU the LFR design studies are performed in the framework of the Euratom FP6 ELSY (European Lead-cooled SYstem) project, where SCK-CEN is a partner. Among the research needed to ensure a safe and reliable operation of the MYRRHA/XT ADS reactor, the development and qualification of fuel and cladding materials have been recognized as one of the main key issues to be addressed

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

Analysis of a Spanish energy scenario with Generation IV nuclear reactors

International Nuclear Information System (INIS)

Ochoa, Raquel; Jimenez, Gonzalo; Perez-Martin, Sara

2013-01-01

Highlights: • Spanish energy scenario for the hypothetical deployment of Gen-IV SFR reactors. • Availability of national resources is assessed, considering SFR’s breeding. • An assessment of the impact of transmuting MA on the final repository. • SERPENT code with own pre- and post-processing tools were employed. • The employed SFR core design is based on the specifications of the CP-ESFR. - Abstract: The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed

Safeguards Licensing Aspects of a Future Gen IV Test Facility - a Case Study

International Nuclear Information System (INIS)

Lindell, M. Aberg; Grape, S.; Hakansson, A.; Svaerd, S. Jacobsson

2010-01-01

The scope of this study covers safeguards licensing aspects of a possible future Gen IV demonstration facility. As a basis for the investigation, the facility was assumed to be located in Sweden, comprising a lead-cooled fast reactor and a reprocessing plant with fuel fabrication. The aim has been to identify safeguards requirements that may be set by the IAEA and the Swedish Radiation Safety Authority, and also to suggest how the safeguards system could be implemented in practice. The changed usage and handling of nuclear fuel, as compared to that of today, has been examined in order to determine how today's safeguards measures can be modified and extended to meet the needs of the demonstration facility. This work is part of GENIUS, the Swedish Gen IV research and development programme, which emphasizes lead-cooled fast reactors. (author)

Gen IV Materials Handbook Implementation Plan

International Nuclear Information System (INIS)

Rittenhouse, P.; Ren, W.

2005-01-01

A Gen IV Materials Handbook is being developed to provide an authoritative single source of highly qualified structural materials information and materials properties data for use in design and analyses of all Generation IV Reactor Systems. The Handbook will be responsive to the needs expressed by all of the principal government, national laboratory, and private company stakeholders of Gen IV Reactor Systems. The Gen IV Materials Handbook Implementation Plan provided here addresses the purpose, rationale, attributes, and benefits of the Handbook and will detail its content, format, quality assurance, applicability, and access. Structural materials, both metallic and ceramic, for all Gen IV reactor types currently supported by the Department of Energy (DOE) will be included in the Gen IV Materials Handbook. However, initial emphasis will be on materials for the Very High Temperature Reactor (VHTR). Descriptive information (e.g., chemical composition and applicable technical specifications and codes) will be provided for each material along with an extensive presentation of mechanical and physical property data including consideration of temperature, irradiation, environment, etc. effects on properties. Access to the Gen IV Materials Handbook will be internet-based with appropriate levels of control. Information and data in the Handbook will be configured to allow search by material classes, specific materials, specific information or property class, specific property, data parameters, and individual data points identified with materials parameters, test conditions, and data source. Details on all of these as well as proposed applicability and consideration of data quality classes are provided in the Implementation Plan. Website development for the Handbook is divided into six phases including (1) detailed product analysis and specification, (2) simulation and design, (3) implementation and testing, (4) product release, (5) project/product evaluation, and (6) product

Uranium enrichment reduction in the Prototype Gen-IV sodium-cooled fast reactor (PGSFR) with PBO reflector

Energy Technology Data Exchange (ETDEWEB)

Hartanto, Donny; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

2016-04-15

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations

International Nuclear Information System (INIS)

Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Bruna, Giovanni; Hache, Georges; Repussard, Jacques

2012-01-01

The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

Thermal stability study for candidate stainless steels of GEN IV reactors

International Nuclear Information System (INIS)

Simeg Veternikova, J.; Degmova, J.; Pekarcikova, M.; Simko, F.; Petriska, M.; Skarba, M.; Mikula, P.; Pupala, M.

2016-01-01

Highlights: • Thermal resistance of advanced stainless steels were observed at 1000 °C. • GEN IV candidate steels were confronted to classic AISI steels. • ODS AISI 316 has weaker thermal resistance than classic AISI steel. • Ferritic ODS steels and NF 709 has better thermal resistance than AISI steels. - Abstract: Candidate stainless steels for GEN IV reactors were investigated in term of thermal and corrosion stability at high temperatures. New austenitic steel (NF 709), austenitic ODS steel (ODS 316) and two ferritic ODS steels (MA 956 and MA 957) were exposed to around 1000 °C in inert argon atmosphere at pressure of ∼8 MPa. The steels were further studied in a light of vacancy defects presence by positron annihilation spectroscopy and their thermal resistance was confronted to classic AISI steels. The thermal strain supported a creation of oxide layers observed by scanning electron microscopy (SEM).

Thermal stability study for candidate stainless steels of GEN IV reactors

Energy Technology Data Exchange (ETDEWEB)

Simeg Veternikova, J., E-mail: jana.veternikova@stuba.sk [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Degmova, J. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Pekarcikova, M. [Institute of Materials Science, Faculty of Materials Science and Technology, Slovak University of Technology, Paulinska 16, 917 24 Trnava (Slovakia); Simko, F. [Department of Molten Salts, Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dubravska cesta 9, 845 36 Bratislava (Slovakia); Petriska, M. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Skarba, M. [Slovak University of Technology, Vazovova 5, 812 43 Bratislava (Slovakia); Mikula, P. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Pupala, M. [Department of Molten Salts, Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dubravska cesta 9, 845 36 Bratislava (Slovakia)

2016-11-30

Highlights: • Thermal resistance of advanced stainless steels were observed at 1000 °C. • GEN IV candidate steels were confronted to classic AISI steels. • ODS AISI 316 has weaker thermal resistance than classic AISI steel. • Ferritic ODS steels and NF 709 has better thermal resistance than AISI steels. - Abstract: Candidate stainless steels for GEN IV reactors were investigated in term of thermal and corrosion stability at high temperatures. New austenitic steel (NF 709), austenitic ODS steel (ODS 316) and two ferritic ODS steels (MA 956 and MA 957) were exposed to around 1000 °C in inert argon atmosphere at pressure of ∼8 MPa. The steels were further studied in a light of vacancy defects presence by positron annihilation spectroscopy and their thermal resistance was confronted to classic AISI steels. The thermal strain supported a creation of oxide layers observed by scanning electron microscopy (SEM).

European project SARGEN IV: safety approach and assessment of GEN IV reactors

International Nuclear Information System (INIS)

Ammirabile, L.

2013-01-01

• SARGEN I V has elaborated a proposal for the harmonization of safety assessment practices for GEN IV NPP. • An overall reinforcement of DiD is expected for GEN I V NPP, including improved independence between all levels of DiD. • An inherent approach should reinforce the fulfillment of fundamental safety functions e.g. the consequences for some situations should be reduced and the grace periods should be extended. For the same reason, the use of passive systems can be envisaged. • The need of complementary and integrated deterministic and probabilistic approaches is reiterated. • Methodologies: Some of them are not yet applied. • Assessment of hazards would be a challenging aspect of next generation of NPP safety assessment and should be improved, which is confirmed by the first insights of Fukushima Daiichi TEPCO reactors accidents. • Provisions to cope with extreme events notably to improve the grace period before cliff-edge effects and thus allowing back-up measures to be implemented have to be defined and should be considered as hardened equipments

The safety R and D for GEN-IV reactors in the European nuclear energy technology platform strategic research agenda

International Nuclear Information System (INIS)

Bruna, G.

2009-01-01

In the fall 2007 EC launched the Sustainable Nuclear Energy Technology Platform (SNE-TP). The SNE-TP governing board set-up three working groups (WG): 1) Strategic Research Agenda (SRA) WG, in charge of drafting road-maps to support research, development and demonstration for current and future NPPs; 2) Deployment Strategy (DS) WG, in charge of defining the research road-map implementation and 3) Education, Training and Knowledge management (ETKM) WG, which was aimed at issuing proposal to reinforce European education and attract young in the nuclear field. The SRA WG was mandated to prepare the SRA vision document based on the preliminary road-map sketched in the document published by the Commission earlier in 2007. The SRA WG was originally organized in 5 sub-groups covering specific topics (1) GEN II and III, III+, including Advanced LWR, 2) Advanced Fuel Cycle for waste minimization and resource optimization; 3) GEN IV Fast Systems (SFR, LFR, GFR, ADS); 4) GEN IV (V) HTR and non-electricity-production applications; 5) New Nuclear Large Research Infrastructures) and 5 other sub-groups dealing with more generic cross-cutting research activities applicable to many specific topics, namely: 1) Structural material research; 2) modeling, simulation and methods, including physical data and tools and means for qualification and validation; 3) Reactor Safety, including severe accidents and human factor; 4) Advanced Driver and Minor Actinide Fuels: science and properties; 5) Pre-normative Research, Codes and Standards.The present paper is mainly aimed at summarizing the content of the SRA Safety sub-chapter focusing on GEN-IV aspects

Fast reactor development and worldwide cooperation in Generation-IV International Forum

International Nuclear Information System (INIS)

Sagayama, Yutaka

2013-01-01

Objectives of Gen-IV systems development: Goals: Four challenging technology goals have been defined to be applied to innovative nuclear reactor concepts in the 21st century: 1) Safety and Reliability (safe and reliable operation, no offsite emergency response); 2) Sustainability (effective fuel utilization, minimization of nuclear waste); 3) Proliferation Resistance & Physical Protection (to assure unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism); 4) Economic Competitiveness (life-cycle cost advantage over other energy resources). Phase: Each Generation-IV reactor system is one of three stages. 1) Viability Phase; 2) Performance Phase; 3) Demonstration Phase. Target: Commercial Deployment is expected around 2030s or beyond

Policy-induced market introduction of Generation IV reactor systems

International Nuclear Information System (INIS)

Heek, Aliki Irina van; Roelofs, Ferry

2011-01-01

Almost 10 years ago the U.S. Department of Energy (DOE) started the Generation IV Initiative (GenIV) with 9 other national governments with a positive ground attitude towards nuclear energy. Some of these Generation IV systems, like the fast reactors, are nearing the demonstration stage. The question on how their market introduction will be implemented becomes increasingly urgent. One main topic for future reactor technologies is the treatment of radioactive waste products. Technological solutions to this issue are being developed. One possible process is the transformation of long-living radioactive nuclides into short living ones; a process known as transmutation, which can be done in a nuclear reactor only. Various Generation IV reactor concepts are suitable for this process, and of these systems most experience has been gained with the sodium-cooled fast reactor (SFR). However, both these first generation SFR plants and their Generation IV successors are designed as electricity generating plants, and therefore supposed to be commercially viable in the electricity markets. Various studies indicate that the generation costs of a combined LWR-(S)FR nuclear generating park (LWR: light water reactor) will be higher than that of an LWR-only park. To investigate the effects of the deployment of the different reactors and fuel cycles on the waste produced, resources used and costs incurred as a function of time, a dynamic fuel cycle assessment is performed. This study will focus on the waste impact of the introduction of a fraction of fast reactors in the European nuclear reactor park with a cost increase as described in the previous paragraph. The nuclear fuel cycle scenario code DANESS is used for this, as well as the nuclear park model of the EU-27 used for the previous study. (orig.)

Fuel, structural material and coolant for an advanced fast micro-reactor

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

2011-01-01

The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

Design and Selection of Innovative Primary Circulation Pumps for GEN-IV Lead Fast Reactors

Directory of Open Access Journals (Sweden)

Walter Borreani

2017-12-01

Full Text Available Although Lead-cooled Fast Reactor (LFR is not a new concept, it continues to be an example of innovation in the nuclear field. Recently, there has been strong interest in liquid lead (Pb or liquid lead–bismuth eutectic (LBE both critical and subcritical systems in a relevant number of Countries, including studies performed in the frame of GENERATION-IV initiative. In this paper, the theoretical and computational findings for three different designs of Primary Circulation Pump (PCP evolving liquid lead (namely the jet pump, the Archimedean pump and the blade pump are presented with reference to the ALFRED (Advanced Lead Fast Reactor European Demonstrator design. The pumps are first analyzed from the theoretical point of view and then modeled with a 3D CFD code. Required design performance of the pumps are approximatively around an effective head of 2 bar with a mass flow rate of 5000 kg/s. Taking into account the geometrical constraints of the reactor and the fluid dynamics characteristics of the molten lead, the maximum design velocity for molten lead fluid flow of 2 m/s may be exceeded giving rise to unacceptable erosion phenomena of the blade or rotating component of the primary pumping system. For this reason a deep investigation of non-conventional axial pumps has been performed. The results presented shows that the design of the jet pump looks like beyond the current technological feasibility while, once the mechanical challenges of the Archimedean (screw pump and the fluid-dynamic issues of the blade pump will be addressed, both could represent viable solutions as PCP for ALFRED. Particularly, the blade pump shows the best performance in terms of pressure head generated in normal operation conditions as well as pressure drop in locked rotor conditions. Further optimizations (mainly for what the geometrical configuration is concerned are still necessary.

A Qualitative Assessment of Diversion Scenarios for a GEN IV Example Sodium Fast Reactor Using the GEN IV PR and PP Methodology

Energy Technology Data Exchange (ETDEWEB)

Zentner, M.D.; Coles, G.A. [PNNL, P.O. Box 999, 902 Battelle Boulvard, Richland, WA 99336 (United States); Therios, I.U. [Argonne National Lab. - ANL (United States)

2009-06-15

An experts working group was created in 2002 by The Generation IV International Forum for the purpose of developing an internationally accepted methodology for assessing the proliferation resistance of a nuclear energy system (NES) and its individual elements. A two year case study was performed by the working group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information to designers at various levels of details, including pre-conceptual design stage. The study analyzes the response of the ESFR entire nuclear energy system to different proliferation and theft strategies. The challenges considered comprise concealed diversion, concealed misuse and abrogation strategies. This paper describes the work done in performing a qualitative assessment of potential concealed diversion scenarios from the ESFR, and includes an evaluation of the potential effect of changes in the conversion ratio on diversion strategies. (authors)

Trends and Developments for Fast Neutron Reactors and Related Fuel Cycles

International Nuclear Information System (INIS)

Carré, Frank

2013-01-01

• FR13 – A unique and dedicated framework to share updates on national programs of Fast Reactor developments, projects of new builds and plans for the future: - Near term projects of sodium and lead-alloy Fast Reactors; - Gen-IV visions of sodium-cooled and alternative types of Fast Neutron Reactors (GFR, LFR…). • FR13 – A special emphasis put on Fast Reactor Safety, Sustainability of nuclear fuel cycle and Young Generation perspective. • FR13 – A catalyst for further collaborations and alliances: - To share visions of goals and advisable options for future Fast Reactors and Nuclear Fuel Cycle; - To share cost of R&D and large demonstrations (safety, security, recycling); - To progress towards harmonized international standards; - To integrate national projects into a consistent international roadmap

Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

Energy Technology Data Exchange (ETDEWEB)

Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Halsey, William [Lawrence Livermore National Laboratory (LLNL); Hayner, George [Idaho National Laboratory (INL); Katoh, Yutai [ORNL; Klett, James William [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Stoller, Roger E [ORNL; Wilson, Dane F [ORNL

2005-12-01

The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor - SCWR and the Very High Temperature Reactor - VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor - GFR, the Lead-cooled Fast Reactor - LFR, and the Sodium-cooled Fast Reactor - SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides, and may provide an alternative to accelerator-driven systems. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development.

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 ...

Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations. Published on September 24, 2012

International Nuclear Information System (INIS)

Couturier, Jean; Bruna, Giovanni; Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Hache, Georges

2012-01-01

The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

Gas Cooled Fast Reactor Research and Development in the European Union

Directory of Open Access Journals (Sweden)

Richard Stainsby

2009-01-01

Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

2008-08-01

Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

The status of proliferation resistance evaluation methodology development in GEN IV international forum

International Nuclear Information System (INIS)

Inoue, Naoko; Kawakubo, Yoko; Seya, Michio; Suzuki, Mitsutoshi; Kuno, Yusuke; Senzaki, Masao

2010-01-01

The Generation IV Nuclear Energy Systems International Forum (GIF) Proliferation Resistance and Physical Protection Working Group (PR and PP WG) was established in December 2002 in order to develop the PR and PP evaluation methodology for GEN IV nuclear energy systems. The methodology has been studied and established by international consensus. The PR and PP WG activities include development of the measures and metrics; establishment of the framework of PR and PP evaluation, the demonstration study using Example Sodium Fast Reactor (ESFR), which included the development of three evaluation approaches; the Case Study using ESFR and four kinds of threat scenarios; the joint study with GIF System Steering Committees (SSCs) of the six reactor design concepts; and the harmonization study with the IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). This paper reviews the status of GIF PR and PP studies and identifies the challenges and directions for applying the methodology to evaluate future nuclear energy systems in Japan. (author)

Comparative analysis of power conversion cycles optimized for fast reactors of generation IV

International Nuclear Information System (INIS)

Perez Pichel, G. D.

2011-01-01

For the study, which is presented here, has been chosen as the specific parameters of each reactor, which are today the three largest projects within generation IV technology development: ESFR for the reactor's sodium, LEADER for the lead reactor's and finally, GoFastR in the case of reactor gas-cooled.

Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-98

Energy Technology Data Exchange (ETDEWEB)

Jammes, C.; Filliatre, P.; De Izarra, G. [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 Saint-Paul-lez-Durance, (France); Elter, Zs.; Pazsit, I. [Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Goteborg, (Sweden); Verma, V.; Hellesen, C.; Jacobsson, S. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala, (Sweden); Hamrita, H.; Bakkali, M. [CEA, DRT, LIST, Sensors and Electronic Architecture Laboratory, Saclay, F-91191 Gif Sur Yvette, (France); Chapoutier, N.; Scholer, A-C.; Verrier, D. [AREVA NP, 10 rue Juliette Recamier F-69456 Lyon, (France); Cantonnet, B.; Nappe, J-C. [PHONIS France S.A.S, Nuclear Instrumentation, Avenue Roger Roncier, B.P. 520, F-19106 Brive Cedex, (France); Molinie, P.; Dessante, P.; Hanna, R.; Kirkpatrick, M.; Odic, E. [Supelec, Department of Power and Energy System, F-91192 Gif Sur Yvette, (France); Jadot, F. [CEA, DEN, DER, ASTRID Project Group, Cadarache, F-13108 Saint-Paul-lez-Durance, (France)

2015-07-01

The neutron flux monitoring system of the French GEN-IV sodium-cooled fast reactor will rely on high temperature fission chambers installed in the reactor vessel and capable of operating over a wide-range neutron flux. The definition of such a system is presented and the technological solutions are justified with the use of simulation and experimental results. (authors)

On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

Directory of Open Access Journals (Sweden)

Jong-Bum Kim

2016-10-01

Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

International Nuclear Information System (INIS)

Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook

2016-01-01

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results

A Qualitative Assessment of Diversion Scenarios for an Example Sodium Fast Reactor Using the GEN IV PR and PP Methodology

International Nuclear Information System (INIS)

Zentner, Michael D.; Coles, Garill A.; Therios, Ike

2012-01-01

FAST REACTORS;NUCLEAR ENERGY;NUCLEAR MATERIALS MANAGEMENT;PROLIFERATION;SAFEGUARDS;THEFT; A working group was created in 2002 by the Generation IV International Forum (GIF) for the purpose of developing an internationally accepted methodology for assessing the Proliferation Resistance of a nuclear energy system (NES) and its individual elements. A two year case study is being performed by the experts group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information at various levels of details to NES designers, safeguard administrators and decision makers. The study analyzes the response of the complete ESFR nuclear energy system to different proliferation and theft strategies. The challenges considered include concealed diversion, concealed misuse and 'break out' strategies. This paper describes the work done in performing a qualitative assessment of concealed diversion scenarios from the ESFR.

Generation IV nuclear reactors: Current status and future prospects

International Nuclear Information System (INIS)

Locatelli, Giorgio; Mancini, Mauro; Todeschini, Nicola

2013-01-01

Generation IV nuclear power plants (GEN IV NPPs) are supposed to become, in many countries, an important source of base load power in the middle–long term (2030–2050). Nowadays there are many designs of these NPPs but for political, strategic and economic reasons only few of them will be deployed. International literature proposes many papers and reports dealing with GEN IV NPPs, but there is an evident difference in the types and structures of the information and a general unbiased overview is missing. This paper fills the gap, presenting the state-of-the-art for GEN IV NPPs technologies (VHTR, SFR, SCWR, GFR, LFR and MSR) providing a comprehensive literature review of the different designs, discussing the major R and D challenges and comparing them with other advanced technologies available for the middle- and long-term energy market. The result of this research shows that the possible applications for GEN IV technologies are wider than current NPPs. The economics of some GEN IV NPPs is similar to actual NPPs but the “carbon cost” for fossil-fired power plants would increase the relative valuation. However, GEN IV NPPs still require substantial R and D effort, preventing short-term commercial adoption. - Highlights: • Generation IV reactors are the middle–long term technology for nuclear energy. • This paper provides an overview and a taxonomy for the designs under consideration. • R and D efforts are in the material, heat exchangers, power conversion unit and fuel. • The life cycle costs are competitive with other innovative technologies. • The hydrogen economy will foster the development of Generation IV reactors

JRC-IE's research of safety of Gen IV systems

International Nuclear Information System (INIS)

Tsige-Tamirat, H.; Ranguelova, V.; Feutterer, M.; Ammirabile, L.; Carlsson, J.; D'Agata, E.; Laurie, M.; Magallon, D.

2010-01-01

The Institute for Energy (IE), one of the seven scientific Institutes of the Joint Research Centre (JRC) of the European Commission, has the mission to provide scientific and technical support for the conception, development, implementation and monitoring of community policies related to energy. To accomplish its mission, IE performs research in the areas of renewable energies, safety and sustainability of nuclear energy for current and future reactor systems, energy technic/economic assessment, and security of energy supply. The Generation IV International Forum (GIF) is a cooperative international endeavour organized to carry out R and D needed to establish the feasibility and performance capabilities of the next generation nuclear energy systems and support the progress towards their realization. The EU, represented by EURATOM and with the JRC as implementing agent, is working together with other GIF partners to perform pre-competitive R and D on key technologies to be implemented in future nuclear systems. IE is engaged in experimental research, simulation and modeling, scientific, feasibility and engineering studies on innovative nuclear reactor systems needed to support the EURATOM contribution to GEN IV initiative, in particular in assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions and knowledge management and preservation. IE's research activities on Generation IV reactor systems are focused on the assessment of the potential of such systems to meet long term EU energy needs with respect to economical advantages, enhanced safety, sustainability, and proliferation resistance. IE participates in international collaborations and has bilateral research cooperation both with European and non-European partners. This paper gives an overview of IE's current research activities on the Gen IV reactor systems related to safety. (authors)

Minor actinides transmutation potential: state of art for GEN IV sodium cooled fast reactors

International Nuclear Information System (INIS)

Buiron, Laurent

2015-01-01

In the frame of the R and D program relative to the 1991 French act on nuclear waste management, fast neutron systems have shown relevant characteristics that meet both requirements on sustainable resources management and waste minimization. They also offer flexibility by mean of burner or breeder configurations allowing mastering plutonium inventory without significant impact on core safety. From the technological point of view, sodium cooled fast reactor are considered in order to achieve mean term industrial deployment. The present document summaries the main results of R and D program on minor actinides transmutation in sodium fast reactor since 2006 following recommendation of the first part of the 1991 French act. Both homogeneous and heterogeneous management achievable performances are presented for 'evolutionary' SFR V2B core as well as low void worth CFV core for industrial scale configurations (1500 MWe). Minor actinides transmutation could be demonstrated in the ASTRID reactor with the following configurations: - a 2%vol Americium content for the homogeneous mode, - a 10%vol Americium content for the heterogeneous mode, without any substantial modification of the main core safety parameters and only limited impacts on the associated fuel cycle (manufacturing issues are not considered here). In order to achieve such goal, a wide range of experimental irradiations driven by transmutation scenarios have to be performed for both homogeneous and heterogeneous minor actinides management. (author) [fr

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

ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

International Nuclear Information System (INIS)

Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

2010-01-01

The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

The generation IV nuclear reactor systems - Energy of future

International Nuclear Information System (INIS)

Ohai, Dumitru; Jianu, Adrian

2006-01-01

Ten nations joined within the Generation IV International Forum (GIF), agreeing on a framework for international cooperation in research. Their goal is to develop future-generation nuclear energy systems that can be licensed, constructed, and operated in an economically competitive way while addressing the issues of safety, proliferation, and other public perception concerns. The objective is for the Gen IV systems to be available for deployment by 2030. Using more than 100 nuclear experts from its 10 member nations, the GIF has developed a Gen IV Technology Roadmap to guide the research and development of the world's most advanced, efficient and safe nuclear power systems. The Gen IV Technology Roadmap calls for extensive research and development of six different potential future reactor systems. These include water-cooled, gas-cooled, liquid metal-cooled and nonclassical systems. One or more of these reactor systems will provide the best combination of safety, reliability, efficiency and proliferation resistance at a competitive cost. The main goals for the Gen IV Nuclear Energy Systems are: - Provide sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel use for worldwide energy production; - Minimize and manage their nuclear waste and noticeably reduce the long-term stewardship burden in the future, improving the protection of public health and the environment; - Increase the assurance that these reactors are very unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased protection against acts of terrorism; - Have a clear life-cycle cost advantage over other energy sources; - Have a level of financial risk comparable to other energy projects; - Excel in safety and reliability; - Have a low likelihood and degree of reactor core damage. (authors)

Report and analysis on 'PR and PP evaluation. Example sodium fast reactor full system case study'

International Nuclear Information System (INIS)

Sagara, Hiroshi; Inoue, Naoko; Kawakubo, Yoko; Watahiki, Masaru

2011-01-01

The Generation IV (GEN IV) Nuclear Energy Systems International Forum (GIF) Proliferation Resistance and Physical Protection Working Group (PRPP WG) was established in December 2002 in order to develop the PR and valuation methodology for GEN IV nuclear energy systems. In the final report of 'PR and PP Evaluation: Example Sodium Fast Reactor (ESFR) Full System Case Study,' issued in October 2009, the demonstration study of PR and PP evaluation with the qualitative approach are summarized using ESFR with four scenario threats. The present paper reviews and analyzes some results of the ESFR case study, and identifies the challenges and direction for the PR and PP evaluation methodology with quantitative approach. (author)

Comparative analysis of power conversion cycles optimized for fast reactors of generation IV; Analisis comparativo de ciclos de conversion de potencia optimizados para reactores rapidos de generacion IV

Energy Technology Data Exchange (ETDEWEB)

Perez Pichel, G. D.

2011-07-01

For the study, which is presented here, has been chosen as the specific parameters of each reactor, which are today the three largest projects within generation IV technology development: ESFR for the reactor's sodium, LEADER for the lead reactor's and finally, GoFastR in the case of reactor gas-cooled.

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

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2005-09-15

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

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

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2005-09-01

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

Nordic Nuclear Materials Forum for Generation IV Reactors

Energy Technology Data Exchange (ETDEWEB)

Anghel, C. (Studsvik Nuclear AB, Nykoeping (Sweden)); Penttilae, S. (Technical Research Centre of Finland, VTT (Finland))

2010-03-15

A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

Nordic Nuclear Materials Forum for Generation IV Reactors

International Nuclear Information System (INIS)

Anghel, C.; Penttilae, S.

2010-03-01

A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

Sandia Pulse Reactor-IV Project

International Nuclear Information System (INIS)

Reuscher, J.A.

1983-01-01

Sandia National Laboratories has developed, designed and operated fast burst reactors for over 20 years. These reactors have been used for a variety of radiation effects programs. During this period, programs have required larger irradiation volumes primarily to expose complex electronic systems to postulated threat environments. As experiment volumes increased, a new reactor was built so that these components could be tested. The Sandia Pulse Reactor-IV is a logical evolution of the two decades of fast burst reactor development at Sandia

GIF (Gen-IV International Forum) Symposium 2009. Proceedings

International Nuclear Information System (INIS)

2009-01-01

The objective of this symposium is to give a well documented state of the art of the initiative and to report and discuss the most significant technical progress and evolution in the different areas during these last ten years. Another significant objective is to provide a forum for an open and hopefully lively discussion of the perspectives, priorities and challenges for the next few years, accounting for a rapidly evolving environment. The symposium has been organized into three sessions that have dealt with the following issues: -) Generation IV International Forum (GIF): 10 years of achievements and the path forward, -) Methodology Overviews and Focus on Applications, -) Very High Temperature Reactor (VHTR), -) Gas-cooled Fast Reactor (GFR), -) Super-Critical Water-cooled Reactor (SCWR), -) Lead-cooled Fast Reactor (LFR), -) Molten Salt Reactor (MSR), -) Sodium-cooled Fast Reactor (SFR), -) International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) and its potential synergy with GIF, and -) GIF priority objectives for the next 5 years

Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

Energy Technology Data Exchange (ETDEWEB)

Behafarid, F.; Shaver, D. R. [Rensselaer Polytechnic Inst., Troy, NY (United States); Bolotnov, I. A. [North Carolina State Univ., Raleigh, NC (United States); Jansen, K. E. [Univ. of Colorado, Boulder, CO (United States); Antal, S. P.; Podowski, M. Z. [Rensselaer Polytechnic Inst., Troy, NY (United States)

2012-07-01

The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

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.

Discussion on safety analysis approach for sodium fast reactors

International Nuclear Information System (INIS)

Hong, Soon Joon; Choo, Yeon Joon; Suh, Nam Duk; Shin, Ahn Dong; Bae, Moo Hoon

2012-01-01

Utilization of nuclear energy is increasingly necessary not only because of the increasing energy consumption but also because of the controls on greenhouse emissions against global warming. To keep step with such demands, advanced reactors are now world widely under development with the aims of highly economical advances, and enhanced safety. Recently, further elaborating is encouraged on the research and development program for Generation IV (GEN IV) reactors, and in collaboration with other interested countries through the Generation IV International Forum (GIF). Sodium cooled Fast Reactor (SFR) is a strong contender amongst the GEN IV reactor concepts. Korea also takes part in that program and plans to construct demonstration reactor of SFR. SFR is under the development for a candidate of small modular reactors, for example, PRISM (Power Reactor Innovative Small Module). Understanding of safety analysis approach has also advanced by the demand of increasing comprehensive safety requirement. Reviewing the past development of the licensing and safety basis in the advanced reactors, such approaches seemed primarily not so satisfactory because the reference framework of licensing and safety analysis approach in the advanced reactors was always the one in water reactors. And, the framework is very plant specific one and thereby the advanced reactors and their frameworks don't look like a well assorted couple. Recently as a result of considerable advances in probabilistic safety assessment (PSA), risk informed approaches are increasingly applied together with some of the deterministic approaches like as the ones in water reactors. Technology neutral framework (TNF) can be said to be the utmost works of such risk informed approaches, even though an intensive assessment of the applicability has not been sufficiently accomplished. This study discusses the viable safety analysis approaches for the urgent application to the construction of pool type SFR. As discussed in

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

Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium

Directory of Open Access Journals (Sweden)

Baschwitz Anne

2016-01-01

Full Text Available International organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR. For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU, though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment

Uranium Enrichment Reduction in the Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR with PBO Reflector

Directory of Open Access Journals (Sweden)

Chihyung Kim

2016-04-01

Full Text Available The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

Nordic forum for generation IV reactors, status and activities in 2012

International Nuclear Information System (INIS)

Van Nieuwenhove, R.; Lauritzen, B.; Nonboel, E.

2012-12-01

The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risoe, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many aspects in Gen-IV reactor research and gave a good overview of the activities within this field at the various institutes and universities. The present report contains book of abstracts. The individual Power Point presentations are indexed in INIS and may be found at http://nordic-gen4.org/seminars/nordic-gen4-riso-2012-2/ (LN)

Nordic forum for generation IV reactors, status and activities in 2012

Energy Technology Data Exchange (ETDEWEB)

Van Nieuwenhove, R. [Institutt for Energiteknikk, OECD Halden Reactor Project, Kjeller (Norway); Lauritzen, B.; Nonboel, E. [Technical Univ. of Denmark. DTU Nutech, Roskilde (Denmark)

2012-12-15

The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risoe, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many aspects in Gen-IV reactor research and gave a good overview of the activities within this field at the various institutes and universities. The present report contains book of abstracts. The individual Power Point presentations are indexed in INIS and may be found at http://nordic-gen4.org/seminars/nordic-gen4-riso-2012-2/ (LN)

A Qualitative Assessment Of Diversion Scenarios For A Example Sodium Fast Reactor Using The Gen IV PR And PP Methodology

International Nuclear Information System (INIS)

Zentner, Michael D.

2008-01-01

A working group was created in 2002 by the Generation IV International Forum (GIF) for the purpose of developing an internationally accepted methodology for assessing the Proliferation Resistance of a nuclear energy system (NES) and its individual elements. A two year case study is being performed by the experts group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information at various levels of details to NES designers, safeguard administrators and decision makers. The study analyzes the response of the complete ESFR nuclear energy system to different proliferation and theft strategies. The challenges considered include concealed diversion, concealed misuse and 'break out' strategies. This paper describes the work done in performing a qualitative assessment of concealed diversion scenarios from the ESFR.

Current status of NPP generation IV

International Nuclear Information System (INIS)

Yohanes Dwi Anggoro; Dharu Dewi; Nurlaila; Arief Tris Yuliyanto

2013-01-01

Today development of nuclear technology has reached the stage of research and development of Generation IV nuclear power plants (advanced reactor systems) which is an innovative development from the previous generation of nuclear power plants. There are six types of power generation IV reactors, namely: Very High Temperature Reactor (VHTR), Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), and Super Critical Water-cooled Reactor (SCWR). The purpose of this study is to know the development of Generation IV nuclear power plants that have been done by the thirteen countries that are members of the Gen IV International Forum (GIF). The method used is review study and refers to various studies related to the current status of research and development of generation IV nuclear power. The result of this study showed that the systems and technology on Generation IV nuclear power plants offer significant advances in sustainability, safety and reliability, economics, and proliferation resistance and physical protection. In addition, based on the research and development experience is estimated that: SFR can be used optimally in 2015, VHTR in 2020, while NPP types GFR, LFR, MSR, and SCWR in 2025. Utilization of NPP generation IV said to be optimal if fulfill the goal of NPP generation IV, such as: capable to generate energy sustainability and promote long-term availability of nuclear fuel, minimize nuclear waste and reduce the long term stewardship burden, has an advantage in the field of safety and reliability compared to the previous generation of NPP and VHTR technology have a good prospects in Indonesia. (author)

A Global Assessment of Fast Reactors in the Future

International Nuclear Information System (INIS)

Devezeaux de Lavergne, J-G.; Mathonnière, G.

2013-01-01

Conclusions: • Fast reactors are the only way to fully achieve nuclear sustainability. • The SFR market cannot exist if a recycling market is not already present. • SFR has many other advantages that clearly outwheight the disadvantages (this trend is increasing). • Large data uncertainties (on uranium resources, world nuclear fleet deployment) return the little precise period at which economic competitiveness will be reached. Anyway, it is most likely to occur sometime in the second half of the century. • However, the market will start earlier, as it is splitted in two phases: before and after the economic competitiveness (this event is in fact country-dependant): – In the first phase 0-2 reactors will be built every year; – In the second phase up to 10-15 reactors will be built every year. • It is rather probable that there will be no more than two or three different Gen IV technologies in the world, because of the market size

Building competencies for New Gen IV Reactors

International Nuclear Information System (INIS)

Pavel, G.L.; Ghitescu, P.

2015-01-01

The Advanced Lead Fast Reactor European Demonstrator - ALFRED is designed and sustained by several European countries. It is a 300 MWt (125 MWe) reactor, intended to be built in Romania, near the Pitesti site. Pure lead is used as primary coolant and it is foreseen to have a 40% thermal efficiency. Secondary cycle contains superheated water steam at around 450 Celsius degrees. Through ARCADIA cooperation, 26 partners from all over Europe joined their forces to provide the necessary research support for ALFRED. In Romania, several entities are providing nuclear courses but only the University Politechnica of Bucharest is offering a complete training program for nuclear industry but targeted courses for LFR technology need to be developed and implemented. Issues like physics of breeding, coolant analysis and behavior, targeted computer codes, core design and dynamics, safety still needs to be tackled

Conceptual core designs for a 1200 MWe sodium cooled fast reactor

International Nuclear Information System (INIS)

Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

2008-01-01

The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

R and D Trends For The Future Sodium Fast Reactors In France

International Nuclear Information System (INIS)

Dufour, Ph.; Anzieu, P.; Lecarpentier, D.; Serpantie, JP.

2006-01-01

The sodium fast reactors are the natural Generation IV candidate, thanks to their strong potential for incineration and/or breeding that allow drastic fissile materials economy and fission waste products recycling or transmutation. The question is now to make evolve the existing or past projects of reactors to systems fully compatible with Generation IV objectives, in particular with regard to the economy, durability and safety. This work must be achieved in an international frame which requires a sharing of the objectives and will allow, in the long term, the sharing of the activities. However, in order to ensure the overall coherence of the various development programs defined within the Gen-IV framework, it is necessary to define a new SFR development plan based on the experience gained in France (Phenix, Superphenix) and Europe, in the EFR project. The commonly agreed SFR system issues to be improved or further investigated are its capital cost, safety issues (sodium risks, core criticality accidents), and in-service inspection and maintenance technology. (authors)

Gen IV International Forum - GIF, 2010 Annual Report

International Nuclear Information System (INIS)

Anon.

2011-01-01

The Generation IV International Forum (GIF), created in 2000 to foster international collaboration at a detailed level of actual R and D, is a cooperative international endeavor, organized to develop the research necessary to test the feasibility and performance capabilities of fourth generation nuclear systems, with the goal of making such systems deployable in large numbers around 2030. Since its beginning, GIF members stated the following goals for the fourth generation of nuclear power plants when compared to previous generations: a) improve sustainability (including effective fuel utilization and minimization of waste); b) improve economics (competitiveness with respect to other energy sources); c) improve safety and reliability (e.g. no need for offsite emergency response); and d) improve proliferation resistance and physical protection. After an in-depth analysis of the different available concepts, whatever their level of development, the Forum selected six concepts as the most promising, and decided to focus R and D on these systems: - the very-high-temperature reactor (VHTR); - the sodium-cooled fast reactor (SFR); - the supercritical-water-cooled reactor (SCWR); - the gas-cooled fast reactor (GFR); - the lead-cooled fast reactor (LFR); - the molten salt reactor (MSR). Active members of the GIF are Canada, Euratom, France, Japan, People's Republic of China, Republic of Korea, Republic of South Africa, Russian Federation, Switzerland and the United States. Altogether, they represent around 90% of the world installed nuclear capacity for producing electricity, and all key technology holders. The forum is led by the policy group, where all members are represented, and currently chaired by Japan since December 2009, assisted by vice-chairs from France and United States. The year 2010 has seen some important achievements and decisions regarding these six systems. For example, two sodium-cooled fast reactors (re)started this year: Monju in Japan restarted after

Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors (Workshop Report)

Energy Technology Data Exchange (ETDEWEB)

Stoller, RE

2004-07-15

The ''Workshop on Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors'' was convened to determine the degree to which an increased effort in modeling and simulation could help bridge the gap between the data that is needed to support the implementation of these advanced nuclear technologies and the data that can be obtained in available experimental facilities. The need to develop materials capable of performing in the severe operating environments expected in fusion and fission (Generation IV) reactors represents a significant challenge in materials science. There is a range of potential Gen-IV fission reactor design concepts and each concept has its own unique demands. Improved economic performance is a major goal of the Gen-IV designs. As a result, most designs call for significantly higher operating temperatures than the current generation of LWRs to obtain higher thermal efficiency. In many cases, the desired operating temperatures rule out the use of the structural alloys employed today. The very high operating temperature (up to 1000 C) associated with the NGNP is a prime example of an attractive new system that will require the development of new structural materials. Fusion power plants represent an even greater challenge to structural materials development and application. The operating temperatures, neutron exposure levels and thermo-mechanical stresses are comparable to or greater than those for proposed Gen-IV fission reactors. In addition, the transmutation products created in the structural materials by the high energy neutrons produced in the DT plasma can profoundly influence the microstructural evolution and mechanical behavior of these materials. Although the workshop addressed issues relevant to both Gen-IV and fusion reactor materials, much of the discussion focused on fusion; the same focus is reflected in this report. Most of the physical models and computational methods

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

Directory of Open Access Journals (Sweden)

Chan Bock Lee

2016-10-01

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

Thermodynamic Data to Model the Interaction Between Coolant and Fuel in Gen IV Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Dinsdale, Alan; Gisby, John; Davies, Hugh; Konings, Rudy; Benes, Ondrej

2013-06-01

Understanding the behaviour of nuclear fuels in various environments is vital to the design and safe operation of nuclear reactors. While this is true if the reactor is operating within its design specification, it is even more so if accidents occur and the fuel is exposed to unexpected temperatures, pressures or chemical environments. It is clearly hazardous and costly to explore all such scenarios experimentally and therefore it is necessary to undertake modelling where possible using well-grounded theoretical approaches. This paper will show examples of where calculations of chemical and phase equilibria have been applied successfully to the long term storage of nuclear waste, phase formation during core meltdown and prediction of fission product release into the atmosphere. It will also highlight the development of thermodynamic data carried out during the European Metrology Research Project Metrofission required to model the potential interaction between the coolant, nuclear fuel, containment materials and atmosphere of a sodium cooled fast reactor. (authors)

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

Energy Technology Data Exchange (ETDEWEB)

Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

2017-02-15

Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

Environmental sensitivity studies for Gen-IV roadmap DUPIC scenario

International Nuclear Information System (INIS)

Jeong, Chang Joon

2004-03-01

The environmental effect of the DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel cycle, which is considered as one of the partial recycle scenario in Gen-IV roadmap, has been analyzed by using the dynamic analysis method. Through the parametric calculations for the DUPIC fuel cycle deployment time and the fraction of the DUPIC reactors, the environmental effects of the fuel cycle for important parameters such as the amount of spent fuel and the combined amounts of plutonium and minor actinides were estimated and compared to those of the once-through LWR fuel cycle. The results of the sensitivity calculations showed that an early deployment of the DUPIC fuel cycle with a high DUPIC reactor fraction can reduce the accumulation of spent fuel by up to 40%. More important is the associated reduction in the combined amount of plutonium and minor actinides, which may reduce the key repository parameter (long term decay heat). Therefore it is expected that favorable environmental effects will be the outcome of the implementation of the DUPIC fuel cycle

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

International Nuclear Information System (INIS)

Yun, Sunghwan; Kim, Sang Ji

2015-01-01

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

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

Development and application of an advanced fuel model for the safety analysis of the generation IV gas-cooled fast reactor

International Nuclear Information System (INIS)

Petkevich, P.

2008-09-01

Until about the year 2030, current-day nuclear power plants (NPPs) will be replaced by so-called Gen-III or Gen-III+ units, which are mainly based on light water reactor technology. The principal new features are increased safety and improved economical effectiveness. However, these systems use the same fuel forms and are based on the same fuel cycle. Beyond 2030, the interest is likely to shift towards fourth generation NPPs, which offer the possibility of complete fuel cycle closure. Generation-IV reactor concepts include both thermal and fast systems, and involve a wide range of fuel forms and compositions. The present research has been focused on the development of a thermo-mechanical model for the innovative fuel design of the Generation-IV Gas-cooled Fast Reactor (GFR). The principal distinctive feature of the fuel is that the fuel pellets are arranged within plates which enclose an inner honeycomb structure. Apart from the geometry, the usage of new materials is foreseen. Thus, the fuel pellets are of mixed uranium-plutonium carbide, and the cladding is bulk or fiber-reinforced SiC. The setting up of an appropriate materials database was thus the very first task which had to be carried out in the current work. The main purpose of the currently developed model is to provide reliable data, in the context of transient analysis, for the calculation of the principal neutronic feedbacks in the GFR core, viz. the fuel temperature for the Doppler effect and the fuel plate deformation for the axial core expansion effect. None of the available fuel modeling codes is suitable for a realistic simulation of the GFR fuel, as the inner honeycomb structure cannot be explicitly taken into account. The development work has been carried out largely in the context of PSI’s generic code system for fast reactor safety analysis, FAST. Thereby, it has mainly involved extension of the thermo-mechanical code FRED, developed originally for the modeling of traditional rodded fuel

Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

NARCIS (Netherlands)

van Rooijen, W.F.G.

2006-01-01

This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor

International Nuclear Information System (INIS)

Stauff, N.

2011-01-01

Compared with earlier plant designs (Phenix, Super-Phenix, EFR), Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel

Thermal analysis of supercritical CO2 power cycles: Assessment of their suitability to the forthcoming sodium fast reactors

International Nuclear Information System (INIS)

Pérez-Pichel, G.D.; Linares, J.I.; Herranz, L.E.; Moratilla, B.Y.

2012-01-01

Highlights: ► This paper investigates the potential use of S-CO 2 cycles in SFRs. ► A wide range of configurations have been explored. ► It is feasible to reach a thermal efficiency as high as 43.5%. ► A sensitivity analysis together with an exergy study have been done. ► Potential use in SFRs of recompression S-CO 2 cycles for their balance of plant. - Abstract: Sodium fast reactors (SFRs) potential to meet Gen. IV requirements is broadly acknowledged worldwide. The scientific and technological experience accumulated by operating test reactors and, even, by running commercial reactors, makes them be considered as the closest Gen. IV option in the near future. In the past their balance of plant has been always based on Rankine cycles. This paper investigates the potential use of supercritical recompression CO 2 cycles (S-CO 2 ) in SFRs on the basis of the working parameters foreseen within the European Sodium Fast Reactor (ESFR) project. A wide range of configurations have been explored, from the simplest one to combined cycles (with organic Rankine cycles, ORC), and a comparison has been set in terms of thermal efficiency. As a result, it has been found out that the most basic configuration could reach a thermal efficiency as high as 43.31%, which is comparable to that obtained through super-critical Rankine cycles proposed elsewhere. A sensitivity analysis together with an exergy study of this configuration, pointed the pre-cooler and IHX Na–CO 2 as key components in the cycle performance. These results highlight a main conclusion: the potential use in SFRs of recompression S-CO 2 cycles for their balance of plant, whenever a sound and extensive database is built-up on S-CO 2 turbo-machinery and IHX performance.

Generation IV reactors: international projects

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

2003-01-01

Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

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

Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

International Nuclear Information System (INIS)

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

2008-01-01

This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities

Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

Energy Technology Data Exchange (ETDEWEB)

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

2008-01-15

This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

Directory of Open Access Journals (Sweden)

Pengcheng Zhao

2016-01-01

Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

International Nuclear Information System (INIS)

Van Nieuwenhove, R.

2012-01-01

A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

Energy Technology Data Exchange (ETDEWEB)

Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (Norway))

2012-01-15

A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

Nuclear data uncertainty analysis for the generation IV gas-cooled fast reactor

International Nuclear Information System (INIS)

Pelloni, S.; Mikityuk, K.

2012-01-01

For the European 2400 MW Gas-cooled Fast Reactor (GoFastR), this paper summarizes a priori uncertainties, i.e. without any integral experiment assessment, of the main neutronic parameters which were obtained on the basis of the deterministic code system ERANOS (Edition 2.2-N). JEFF-3.1 cross-sections were used in conjunction with the newest ENDF/B-VII.0 based covariance library (COMMARA-2.0) resulting from a recent cooperation of the Brookhaven and Los Alamos National Laboratories within the Advanced Fuel Cycle Initiative. The basis for the analysis is the original GoFastR concept with carbide fuel pins and silicon-carbide ceramic cladding, which was developed and proposed in the first quarter of 2009 by the 'French alternative energies and Atomic Energy Commission', CEA. The main conclusions from the current study are that nuclear data uncertainties of neutronic parameters may still be too large for this Generation IV reactor, especially concerning the multiplication factor, despite the fact that the new covariance library is quite complete; These uncertainties, in relative terms, do not show the a priori expected increase with bum-up as a result of the minor actinide and fission product build-up. Indeed, they are found almost independent of the fuel depletion, since the uncertainty associated with 238 U inelastic scattering results largely dominating. This finding clearly supports the activities of Subgroup 33 of the Working Party on International Nuclear Data Evaluation Cooperation (WPEC), i.e. Methods and issues for the combined use of integral experiments and covariance data, attempting to reduce the present unbiased uncertainties on nuclear data through adjustments based on available experimental data. (authors)

The European Lead Fast Reactor Strategy and the Roadmap for the Demonstrator ALFRED

International Nuclear Information System (INIS)

Alemberti, A.; De Bruyn, D.; Grasso, G.; Mansani, L.; Mattioli, D.; Roelofs, F.

2013-01-01

Expected impacts: → To ensure that nuclear energy remains a long-term contributor to a low carbon economy it is necessary to increase its sustainability through demonstrating the technical, industrial and economic viability of Gen IV fast nuclear reactors; → With the construction and operation of MYRRHA and ALFRED, Europe will be in an excellent position to secure the development of a safe, sustainable and competitive fast nuclear technology; → ALFRED Demonstrator Roadmap will: • play a key role by involving European industry and maintaining and developing European leadership in nuclear technologies worldwide; • allow to investigate and address the main technological issues that can be implemented in the LFR prototype (2035); • make possible commercial deployment, by the European industry, of these technologies by 2050 and beyond; • contribute significantly to the development of a sustainable and secure energy supply for Europe from the second half of this century onwards

Improvement of Steam Generator Reliability for GEN-IV SFR

Energy Technology Data Exchange (ETDEWEB)

Kim, Seong O; Kim Se Yun; Kim, Seok Hoon; Eoh, Jae Hyuk; Lee, Hyeong Yeon; Choi, Byung Seon

2005-11-15

The R and D items performed in this study were selected from the R and D task of ' Reliability improvement of Steam Generator' of GEN-IV SFR Component Design and BOP. Since this project deals with one of the most important issues for a GEN-IV SFR system, it needs to enhance the domestic technical backgrounds associated with the corresponding R and D items even for a very short period by 2005. This study provides the R and D results for i) Development of assessment methodology for dissimilar metal weld and ii) Development of multi-dimensional simulation methodology for a SWR event in a SFR steam generator.

Improvement of Steam Generator Reliability for GEN-IV SFR

International Nuclear Information System (INIS)

Kim, Seong O; Kim Se Yun; Kim, Seok Hoon; Eoh, Jae Hyuk; Lee, Hyeong Yeon; Choi, Byung Seon

2005-11-01

The R and D items performed in this study were selected from the R and D task of ' Reliability improvement of Steam Generator' of GEN-IV SFR Component Design and BOP. Since this project deals with one of the most important issues for a GEN-IV SFR system, it needs to enhance the domestic technical backgrounds associated with the corresponding R and D items even for a very short period by 2005. This study provides the R and D results for i) Development of assessment methodology for dissimilar metal weld and ii) Development of multi-dimensional simulation methodology for a SWR event in a SFR steam generator

Fast wave current drive in reactor scale tokamaks

International Nuclear Information System (INIS)

Moreau, D.

1992-01-01

The IAEA Technical Committee Meeting on Fast Wave Current Drive in Reactor Scale Tokamaks, hosted by the Commissariat a l'Energie Atomique (CEA), Departement de Recherches sur la Fusion Controlee (Centres d'Etudes de Cadarache, under the Euratom-CEA Association for fusion) aimed at discussing the physics and the efficiency of non-inductive current drive by fast waves. Relevance to reactor size tokamaks and comparison between theory and experiment were emphasized. The following topics are described in the summary report: (i) theory and modelling of radiofrequency current drive (theory, full wave modelling, ray tracing and Fokker-Planck calculations, helicity injection and ponderomotive effects, and alternative radio-frequency current drive effects), (ii) present experiments, (iii) reactor applications (reactor scenarios including fast wave current drive; and fast wave current drive antennas); (iv) discussion and summary. 32 refs

Developing new nuclear curricula for GEN IV needs

International Nuclear Information System (INIS)

Ghitescu, P.; Pavel, G.L.

2014-01-01

States who wish to start and develop a nuclear program must take into consideration a strong proven strategy for developing a sustainable program. A complete nuclear research program must include: a good national strategy and support on the topic; strong research laboratories supported by good personnel; education component to provide sustainable and qualified workforce; national/international interest from stakeholders and governments and a well informed society. New demonstrators are foreseen for the next period to be built in Europe and skilled supporting personnel is strongly needed. Current situation in nuclear higher education with perspective will be analysed. EURATOM strongly supports development of multidisciplinary co-operational projects in order to built such novel initiatives. An example of such program supported by European Commission, ARCADIA, will be given. The project is based on the cooperation of a large number of participants all over Europe and the main purpose is to develop a road-map for Gen IV reactor. (authors)

JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors. (3) Progress of component design

International Nuclear Information System (INIS)

Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi; Eto, Masao; Miyagawa, Takayuki

2015-01-01

In the frame work of generation IV international forum (GIF), safety design criteria (SDC) and safety design guideline (SDG) for the generation IV sodium-cooled fast reactors have been developing in the circumstance of worldwide deployment of SFRs. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC in the feasibility study of SDG for Sodium-cooled Fast Reactor (SFR). In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX (pump/IHX) was modified for the primary heat exchanger (PHX), which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator (SG), protective wall tube type design is under investigation as an option with less R and D risks. (author)

JSFR design progress related to development of safety design criteria for Generation IV sodium-cooled fast reactors. (1) Overview

International Nuclear Information System (INIS)

Kamide, Hideki; Ando, Masato; Ito, Takaya

2015-01-01

JAEA, JAPC and MFBR have been conducting design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. As the result of the design study and R and D activity related the innovative technologies incorporated in the design in the Fast Reactor Cycle Technology Development (FaCT) project up to 2010, basic design concept of JSFR was established and its development process to the commercialization including construction and operation of a demonstration version of JSFR was outlined. JSFR is a looptype next generation sodium-cooled fast reactor (SFR), which is aiming at achieving development targets of Generation IV reactors concerning sustainability, safety and reliability, economics and proliferation resistance and physical protection by introducing the innovative technologies such as shortened high-chromium steel piping. The output power is assumed for the design study as 1,500 MWe for the commercial version and 750 MWe for the demonstration version. In FaCT phase I up to 2010, in order to evaluate feasibility to achieve the development targets, the design study has been conducted on the main components and systems. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lessons learned from the TEPCO's Fukushima Dai-ichi nuclear power plants accident, in the frame work of Generation IV International Forum (GIF), the design study is focusing on the design measures against severe external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated for the safety issues on SDC and SDG of a SFR. (author)

Overview of materials R and D for fusion and Gen-4

Energy Technology Data Exchange (ETDEWEB)

Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan); Tavassoli, F.; Carre, F.; Billot, P. [CEA Saclay, 91 - Gif sur Yvette (France); Zinide, S. [Oak Ridge National Laboratory, Materials Science and Technology Div., AK TN (United States)

2007-07-01

Full text of publication follows: In view of the growing need for energy, the risk of exhaustion of fossil fuel and the problem of global warming, the nuclear energy is receiving added attention as a realistic and viable advanced solution. International collaborations on Generation IV (Gen-IV) fission reactors and on ITER and DEMO fusion reactors are developing. This is particularly the case in the sector of materials, where they hold the key to success of these systems. The international community has recognized and planned its materials R and D work for Fusion and Gen-IV reactors with the following considerations: 1- The time allotted to materials R and D is short and may not allow development of totally new materials. 2- Activities required, to cover existing materials variations and service conditions necessary for reactor design, are very time consuming. 3- The work to be done must build upon the existing knowledge of materials and avoid duplications. Although ITER for fusion and Generation four International Forum (GIF) for Gen-IV are important international collaborative programs, they are insufficient to meet all the national energy policies of the participating countries. This paper provides an overview of the materials R and D carried out for fusion and Gen-IV reactors at international and national levels. Materials programs discussed include both cross-cutting and reactor specific actions, where major tasks can be defined as: + Cross-cutting materials tasks: - materials for high temperature service; - materials with neutron damage tolerance; - materials behavior analysis and modeling; - high temperature design methodology. + Reactor specific materials tasks: - very high temperature alloys; - carbon, high temperature ceramics and their composites; - materials compatibilities. Starting with a brief introduction of materials R and D strategies, ITER and Broader Approach (BA), overall activities for fusion and GIF for Gen-IV will be reviewed. Domestic

Status of the design and safety project for the sodium-cooled fast reactor as a generation IV nuclear energy system

International Nuclear Information System (INIS)

Niwa, Hajime; Fiorini, Gian-Luigi; Sim, Yoon-Sub; Lennox, Tom; Cahalan, James E.

2005-01-01

The Design and Safety Project Management Board (DSPMB) was established under the Sodium Cooled Fast Reactor (SFR) System Steering Committee (SSC) in the Generation IV international Forum. The DSPMB will promote collaborative R and D activities on reactor core design, and safety assessment for candidate systems, and also integrate these results together with those from other PMBs such as advanced fuel and component to a whole fast reactor system in order to develop high performance systems that will satisfy the goals of Generation IV nuclear energy systems. The DSPMB has formulated the present R and D schedules for this purpose. Two SFR concepts were proposed: a loop-type system with primarily a MOX fuel core and a pool-type system with a metal fuel core. Study of innovative systems and their evaluation will also be included. The safety project will cover both the safety assessment of the design and the preparation of the methods/tools to be used for the assessment. After a rather short viability phase, the project will move to the performance phase for development of performance data and design optimization of conceptual designs. This paper describes the schedules, work packages and tasks for the collaborative studies of the member countries. (author)

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

High temperature fast reactor for hydrogen production in Brazil

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F.

2008-01-01

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

A Review of PSA Technology Applications according to the Development of Sodium-cooled Fast Reactors in the World

International Nuclear Information System (INIS)

Kim, Tae Woon; Lee, Yong Bum; Jung, Hae Yong; Kim, Sang Ji; Hahn, Do Hee; Yang, Joon Eon

2008-12-01

The international nuclear societies request to perform Probabilistic Safety Assessment (PSA) according to the development of Gen IV Sodium-cooled Fast Reactors (SFR). One of the major tasks of the PSA is to identify various sequences of events which could lead to the release of radioactivity. However, due to the limited operating and SFR PSA experiences, it will be difficult to derive and to quantify core damage frequency for SFR under development in Korea, so called KALIMER. Hence, in this report, the foreign PSA results, such as USA and Japan, are analyzed based on the obtained documents. Finally an approach on how to perform PSA for KALIMER is suggested

Conceptual Design on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

International Nuclear Information System (INIS)

Lee, Jae Han; Koo, Gyeong Hoi

2013-01-01

This paper describes the key concept of the drive mechanism, and suggests a required motor power and reducer gears to meet the functional design requirements, and a seismic response analysis of CRDM housing is performed to check its structural integrity. An AC servo motor is selected as a CRA driving power because it uses permanent magnets and is brushless type while DC motor needs a brush and a coil rotates. The control shim motor size is constrained by a housing diameter of 250mm. The driving system has several design requirements. To calculate the motor power, the drive shaft torque is needed. One part of the drive shaft has a lead screw, driving by a ball-nut. The ball screw driver torque (Tr) is calculated by some equations as follow; A servo motor with a nominal power of 100W, a nominal torque of 0.32 N-m (max. 0.48N-m) is selected considering a safety margin. Its diameter is about 50mm. The fast drive-in motor needs a strong power to insert enforcedly the stuck CRA into core within a required time. The motor sizes are calculated by the same procedure. The diameters are in the range of 80mm to 110mm by the insertion time (10 ∼ 24 seconds). The prototype Gen-IV SFR (sodium-cooled Fast Reactor) is of 150MWe capacity. The reactor has six primary control rod assemblies(CRAs). The primary control rod is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. The control rod drive mechanism (CRDM) consists of the drive motor assembly, the driveline, and its housing. The driveline consists of three concentric members of a drive shaft, a tension tube, and a position indicator rod, and it connects the drive motor assembly to the CRA. Main issue is that these many driving parts shall be enclosed within a limited housing diameter because the available pitch of CRDMs is limited by 300mm

Conceptual Design on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Han; Koo, Gyeong Hoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

This paper describes the key concept of the drive mechanism, and suggests a required motor power and reducer gears to meet the functional design requirements, and a seismic response analysis of CRDM housing is performed to check its structural integrity. An AC servo motor is selected as a CRA driving power because it uses permanent magnets and is brushless type while DC motor needs a brush and a coil rotates. The control shim motor size is constrained by a housing diameter of 250mm. The driving system has several design requirements. To calculate the motor power, the drive shaft torque is needed. One part of the drive shaft has a lead screw, driving by a ball-nut. The ball screw driver torque (Tr) is calculated by some equations as follow; A servo motor with a nominal power of 100W, a nominal torque of 0.32 N-m (max. 0.48N-m) is selected considering a safety margin. Its diameter is about 50mm. The fast drive-in motor needs a strong power to insert enforcedly the stuck CRA into core within a required time. The motor sizes are calculated by the same procedure. The diameters are in the range of 80mm to 110mm by the insertion time (10 ∼ 24 seconds). The prototype Gen-IV SFR (sodium-cooled Fast Reactor) is of 150MWe capacity. The reactor has six primary control rod assemblies(CRAs). The primary control rod is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. The control rod drive mechanism (CRDM) consists of the drive motor assembly, the driveline, and its housing. The driveline consists of three concentric members of a drive shaft, a tension tube, and a position indicator rod, and it connects the drive motor assembly to the CRA. Main issue is that these many driving parts shall be enclosed within a limited housing diameter because the available pitch of CRDMs is limited by 300mm.

Thermal analysis of supercritical CO{sub 2} power cycles: Assessment of their suitability to the forthcoming sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Perez-Pichel, G.D., E-mail: gdp@icai.es [Rafael Marino Chair on New Energy Technologies, Comillas Pontifical University, Madrid (Spain); Linares, J.I. [Rafael Marino Chair on New Energy Technologies, Comillas Pontifical University, Madrid (Spain); Herranz, L.E.; Moratilla, B.Y. [Unit of Nuclear Safety Research, CIEMAT, Madrid (Spain)

2012-09-15

Highlights: Black-Right-Pointing-Pointer This paper investigates the potential use of S-CO{sub 2} cycles in SFRs. Black-Right-Pointing-Pointer A wide range of configurations have been explored. Black-Right-Pointing-Pointer It is feasible to reach a thermal efficiency as high as 43.5%. Black-Right-Pointing-Pointer A sensitivity analysis together with an exergy study have been done. Black-Right-Pointing-Pointer Potential use in SFRs of recompression S-CO{sub 2} cycles for their balance of plant. - Abstract: Sodium fast reactors (SFRs) potential to meet Gen. IV requirements is broadly acknowledged worldwide. The scientific and technological experience accumulated by operating test reactors and, even, by running commercial reactors, makes them be considered as the closest Gen. IV option in the near future. In the past their balance of plant has been always based on Rankine cycles. This paper investigates the potential use of supercritical recompression CO{sub 2} cycles (S-CO{sub 2}) in SFRs on the basis of the working parameters foreseen within the European Sodium Fast Reactor (ESFR) project. A wide range of configurations have been explored, from the simplest one to combined cycles (with organic Rankine cycles, ORC), and a comparison has been set in terms of thermal efficiency. As a result, it has been found out that the most basic configuration could reach a thermal efficiency as high as 43.31%, which is comparable to that obtained through super-critical Rankine cycles proposed elsewhere. A sensitivity analysis together with an exergy study of this configuration, pointed the pre-cooler and IHX{sub Na-CO{sub 2}} as key components in the cycle performance. These results highlight a main conclusion: the potential use in SFRs of recompression S-CO{sub 2} cycles for their balance of plant, whenever a sound and extensive database is built-up on S-CO{sub 2} turbo-machinery and IHX performance.

Analysis and evaluation of the Dual Fluid Reactor concept

Energy Technology Data Exchange (ETDEWEB)

Wang, Xiang

2017-06-27

The Dual Fluid Reactor is a molten salt fast reactor developed by IFK in Berlin based on the Gen-IV Molten-Salt Reactor concept and the Liquid-Metal Cooled Reactor. The design aims to combine these two concepts to improve these two concepts. The Dissertation focuses on the concept and performs diverse calculations and estimations on the subjects of neutron physics, depletion and thermal-hydraulic behaviors to validate the new features of the concept. Based on the results it is concluded that this concept is feasible to its desired purpose and with great potential.

The ENEN-III project: Technical Training on the Concepts and Design of GEN IV nuclear reactors

International Nuclear Information System (INIS)

Berkvens, T.; Renault, C.; Alonso, M.; Salomaa, R.; Schönfelder, C.

2013-01-01

Some conclusions: • Not enough training courses to cover the LO’s: – Especially GEN IV; – Many introductory courses, little specific courses; – Reach out to other partners for more courses. • Skills and Attitudes: – Much more difficult to train/measure; – To be treated in a separate project. • Use of Learning Outcomes must be promoted; • Involvement of human resources necessary for the successful implementation of the schemes: – End of project workshop

MYRRHA – A multi-purpose fast spectrum research reactor

International Nuclear Information System (INIS)

Aït Abderrahim, Hamid; Baeten, Peter; De Bruyn, Didier; Fernandez, Rafael

2012-01-01

Highlights: ► Historical evolution of the MYRRHA project. ► Detail design of the MYRRHA Accelerator Driven System. ► Irradiation performance simulation of the MYRRHA ADS. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental Accelerator-Driven System (ADS) currently under development at SCK⋅CEN and will replace the Material Testing Reactor (MTR) BR2. The MYRRHA facility is currently being developed with the aid of the FP7-project “Central Design Team” and will be as a flexible irradiation facility, able to work in both subcritical and critical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV systems, material developments for fusion reactors, radioisotope production for medical and industrial applications, and Si-doping. MYRRHA will also demonstrate the full concept of Accelerator Driven Systems by coupling the requisite three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow for the study of efficient transmutation of high-level nuclear waste. Since MYRRHA is based on the heavy liquid metal technology, Lead–Bismuth Eutectic, it will be able to significantly contribute to the development of Lead Fast Reactor (LFR) technology. Further, in critical mode, MYRRHA will play the role of European Technology Pilot Plant in the path forward for LFR. In this paper we present the historical perspectives, international and high profile membership within the consortium of the MYRRHA project and the rationale for the design choices are presented. Also, the latest configuration of the reactor system is described together with the different irradiation capabilities. More specifically, the possibilities and performances for fuel irradiations are presented in detail.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Choi, Jin Ha; Kim, Myung Hyun

2016-01-01

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

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)

Transmutation of Thermocouples in Thermal and Fast Nuclear Reactors

International Nuclear Information System (INIS)

Scervini, M.; Rae, C.; Lindley, B.

2013-06-01

Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. Their role is fundamental for the control of current nuclear reactors and for the development of the nuclear technology needed for the implementation of GEN IV nuclear reactors. When used for in-core measurements thermocouples are strongly affected not only by high temperatures, but also by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition in the thermoelements and, as a consequence, a time dependent drift in the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. In this work, undertaken as part of the European project METROFISSION, the change in composition occurring in irradiated thermocouples has been calculated using the software ORIGEN 2.2. Several thermocouples have been considered, including Nickel based thermocouples (type K and type N), Tungsten based thermocouples (W-5%Re vs W-26%Re and W- 3%Re vs W-25%Re), Platinum based thermocouples (type S and Platinum vs Palladium) and Molybdenum vs Niobium thermocouples. The transmutation induced by both thermal flux and fast flux has been calculated. Thermocouples undergo more pronounced transmutation in thermal fluxes rather than in fast fluxes, as the neutron cross section of an element is higher for thermal energies. Nickel based thermocouples have a minimal change in composition, while Platinum based and Tungsten based thermocouples experience a very significant transmutation. The use of coatings deposited on the sheath of a thermocouple has been considered as a mean to reduce the neutron flux the thermoelements inside the thermocouple sheath

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

International Nuclear Information System (INIS)

Kim, Sang Ji; Kim, Yong Hee; Hong, Ser Gi

2003-06-01

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

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

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.)

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.

Fast breeder reactors

International Nuclear Information System (INIS)

Heinzel, V.

1975-01-01

The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

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

ASTRID, Generation IV advanced sodium technological reactor for industrial demonstration

International Nuclear Information System (INIS)

Gauche, F.

2013-01-01

ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) is an integrated technology demonstrator designed to demonstrate the operability of the innovative choices enabling fast neutron reactor technology to meet the Generation IV criteria. ASTRID is a sodium-cooled fast reactor with an electricity generating power of 600 MWe. In order to meet the generation IV goals, ASTRID will incorporate the following decisive innovations: -) an improved core with a very low, even negative void coefficient; -) the possible installation of additional safety devices in the core. For example, passive anti-reactivity insertion devices are explored; -) more core instrumentation; -) an energy conversion system with modular steam generators, to limit the effects of a possible sodium-water reaction, or sodium-nitrogen exchangers; -) considerable thermal inertia combined with natural convection to deal with decay heat; -)elimination of major sodium fires by bunkerization and/or inert atmosphere in the premises; -) to take into account off-site hazards (earthquake, airplane crash,...) right from the design stage; -) a complete rethink of the reactor architecture in order to limit the risk of proliferation. ASTRID will also include systems for reducing the length of refueling outages and increasing the burn-up and the duration of the cycle. In-service inspection, maintenance and repair are also taken into account right from the start of the project. The ASTRID prototype should be operational by about 2023. (A.C.)

Fast Spectrum Reactors

CERN Document Server

Todd, Donald; Tsvetkov, Pavel

2012-01-01

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

Drop performance test of conceptually designed control rod assembly for prototype generation IV sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Young Kyu; Lee, Jae Han; Kim, Hoe Woong; KIm, Sung Kyun; Kim, Jong Bum [Sodium-cooled Fast Reactor NSSS Design Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-06-15

The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

Study on high temperature design methodology of heat-resistant materials for GEN-IV systems

Energy Technology Data Exchange (ETDEWEB)

Ryu, Woo Seog; Kim, D. W.; Kim, S. H.; Kim, W. G.; Kim, J. H.; Park, D. G.; Yoon, J. H.; Lee, H. Y.; Hing, J. H

2005-08-15

Analysis of the existing high temperature design and assessment codes such as US(ASME-NH,Draft Code Case for Alloy 617), France(RCC-MR), UK(R5), Japan(BDS/DDS/FDS) for Gen IV reactor structure has been carried out. In addition the scope and fields for research and development is needed in the future have been defined. For assessing the high temperature creep cracks, time dependent fracture mechanics (TDFM) parameters of the C and Ct were analyzed. The creep propagation data were obtained from the creep crack growth tests for type 316LN stainless steels, and creep crack growth testing machine for Gen-IV system up to 950 .deg. C was set up. Damage mechanism and causes for creep-fatigue were investigated. The difference between prediction creep-fatigue life and experimental life were investigated. Material properties for analysis creep-fatigue damage were recommended. The assessment procedure (Draft) on creep-fatigue crack initiation has been developed based on the technical appendix A16 of French RCC-MR code. Ultrasonic wave signal against creep ruptured specimens of type 316LN stainless steel was obtained. It was identified that creep damage can be evaluated by ultrasonic method. The NDT techniques evaluated include Barkhausen noise, magnetic hysteresis parameters, positron annihilation, X-ray diffraction and small angle neutron scattering. Experimental procedure and evaluation method of material integrity were developed through the fracture toughness test of Cr-Mo steel.

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.)

Summary of Structural Concept Development and High Temperature Structural Integrity Evaluation Technology for a Gen-IV SFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Han; Joo, Young Sang; Lee, Hyeong Yeon (and others)

2008-04-15

The economic improvement is a hot issue as one of Gen IV nuclear plant goals. It requires many researches and development works to meet the goal by securing the same level of plant safety. One of the key research items is the increase of the plant capacity with the minimum number of components and loops. Through the successful conceptual design experience for the KALIMER-600, the structural design study for a 1200MWe large capacity of sodium-cooled fast reactor has been performed to achieve the above plant size effects. The component number and reactor structural sizing were determined based on the core and fluid system design information. Several researches were performed to reduce the construction cost of NSSS in structural point of view, for example, a simplified component arrangement, concept proposals of integrated components, a high temperature LBB application technology, and an innovative in-service inspection (ISI) tool, and a computer program development of the ASME-NH design procedure of the class 1 structure and component under high temperature over 500 .deg. C. The IHTS piping arrangement was also proposed to minimize the length through the properly locating the SG and pump by 126m. Further studies of these concepts are required to confirm on the fabricability and the structural integrity for the operating and design loads. The proposed concepts will be optimized to a unified conceptual design through several trade-off studies.

The fast breeder reactor

International Nuclear Information System (INIS)

Collier, J.

1990-01-01

The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

Energy Technology Data Exchange (ETDEWEB)

Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

Development of multi-group xs libraries for the gfr 2400 reactor

International Nuclear Information System (INIS)

Cerba, Š.; Vrban, B.; Lüley, J.; Necas, V.

2016-01-01

GFR 2400 is considered as a conceptual design of the large scale GEN IV Gas-Cooled Fast Reactor. In general, the GEN IV technologies are seen as reliable but also very challenging reactor concepts. Since GFR 2400 lacks any experimental data, the questions on its safety are even more complex and the assessment of its performance could be made only based on computational experience. The paper deals with the development process of multi-group XS libraries based on a hybrid deterministic-Stochastic methodology, using the NJOY99, TRANSX, DIF3D, PARTISN and MCNP5 codes. A new optimized 25 group SBJ E 71 2 5G cross section library was developed based on ENDF/B-VII.1 evaluated data, ZZ-KAFAX-E70 background cross sections and GFR 2400 neutron spectrum. The created library was validated through integral experiments evaluated on the HEX-Z deterministic models in DIF3D. The results were also compared with MCNP5 calculations. (authors)

Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

International Nuclear Information System (INIS)

Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco

2016-01-01

Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

2016-08-15

Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

Study of nitrogen injection to enhance forced convection for gas fast reactors

International Nuclear Information System (INIS)

Tauveron, N.; Dor, I.; Bentivoglio, F.

2011-01-01

Highlights: → The present study concerns the use of blowers in case of nitrogen injection. It is a well-known fact that heavier gases (than helium) enhance natural circulation. The use of such heavier gases (nitrogen is considered here) can also enhance forced convection. → A specific work on the impact of the use of alternative gas on blower behaviour is presented. → These developments are used in a simplified system analysis and in a complete transient behaviour analysis in depressurised situations computed with the CATHARE2 code. - Abstract: In the frame of the international forum GenIV, the gas fast reactor is considered as a promising concept, combining the benefits of fast spectrum and high temperature, using helium as coolant. In the current preliminary viability GFR studies safety system relies on blowers in case of depressurised conditions. The present study concerns the use of blowers in case of nitrogen injection. It is a well-known fact that heavier gases (than helium) enhance natural circulation. The use of such gases (nitrogen is considered) can also enhance forced convection. A specific work on the impact of the use of alternative gas on blower behaviour is presented. Transient behaviours in depressurised situations are computed with the CATHARE2 code and analyzed.

The fast breeder reactor

International Nuclear Information System (INIS)

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

1990-01-01

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

Fast breeder reactors

International Nuclear Information System (INIS)

Waltar, A.E.; Reynolds, A.B.

1981-01-01

This book describes the major design features of fast breeder reactors and the methods used for their design and analysis. The foremost objective of this book is to fulfill the need for a textbook on Fast Breeder Reactor (FBR) technology at the graduate level or the advanced undergraduate level. It is assumed that the reader has an introductory understanding of reactor theory, heat transfer, and fluid mechanics. The book is expected to be used most widely for a one-semester general course on fast breeder reactors, with the extent of material covered to vary according to the interest of the instructor. The book could also be used effectively for a two-quarter or a two-semester course. In addition, the book could serve as a text for a course on fast reactor safety since many topics other than those appearing in the safety chapters relate to FBR safety. Methodology in fast reactor design and analysis, together with physical descriptions of systems, is emphasized in this text more than numerical results. Analytical and design results continue to change with the ongoing evolution of FBR design whereas many design methods have remained fundamentally unchanged for a considerable time

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.)

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.)

{sup 20}F power measurement for generation IV sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Coulon, R.; Normand, S.; Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.M.; Kondrasovs, V.; Frelin-Labalme, A.M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); BAN, G. [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Dautremer, T.; Montagu, T.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Portier, J.L. [Centrale PHENIX, Centre de Marcoule, Groupe Essais Statistiques, F-30207 Bagnols sur Ceze (France); Jousset, P. [CEA, LIST, Laboratoire Capteurs Diamant, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille, France.F-84500 Bollene (France)

2010-07-01

The Phenix nuclear power plant has been a French Sodium Fast Reactor (SFR) prototype producing electrical power between 1973 and 2010. The power was monitored using ex-core neutron measurements. This kind of measurement instantly estimates the power but needs to be often calibrated with the heat balance thermodynamic measurement. Large safety and security margins have then been set not to derive above the nominal operating point. It is important for future SFR to reduce this margin and working closer to the nominal operating point. This work deals with the use of delayed gamma to measure the power. The main activation product contained in the primary sodium coolant is the {sup 24}Na which is not convenient for neutron flux measurement due to its long decay period. The experimental study done at the Phenix reactor shows that the use of {sup 20}F as power tagging agent gives a fast and accurate power measurement closed to the thermal balance measurement thanks to its high energy photon emission (1.634 MeV) and its short decay period (11 s). (authors)

Gen IV Materials Handbook Functionalities and Operation

International Nuclear Information System (INIS)

Ren, Weiju

2009-01-01

This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.

Gen IV Materials Handbook Functionalities and Operation

Energy Technology Data Exchange (ETDEWEB)

Ren, Weiju [ORNL

2009-12-01

This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.

DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

Energy Technology Data Exchange (ETDEWEB)

William E. Kastenberg; Edward Blandford; Lance Kim

2009-03-31

This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public.

DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

International Nuclear Information System (INIS)

Kastenberg, William E.; Blandford, Edward; Kim, Lance

2009-01-01

This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public

Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

Science.gov (United States)

Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

2017-01-01

Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

Fast breeder reactor research

International Nuclear Information System (INIS)

1975-01-01

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

New Materials for NGNP/Gen IV

International Nuclear Information System (INIS)

Swindeman, Robert W.; Marriott, Douglas L.

2009-01-01

The bounding conditions were briefly summarized for the Next Generation Nuclear Plant (NGNP) that is the leading candidate in the Department of Energy Generation IV reactor program. Metallic materials essential to the successful development and proof of concept for the NGNP were identified. The literature bearing on the materials technology for high-temperature gas-cooled reactors was reviewed with emphasis on the needs identified for the NGNP. Several materials were identified for a more thorough study of their databases and behavioral features relative to the requirements ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NH.

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.

Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

International Nuclear Information System (INIS)

Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Dařílek, Petr; Zajac, Radoslav; Nečas, Vladimír; Haščik, Ján

2014-01-01

Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice

Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

International Nuclear Information System (INIS)

Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E.; Lovera, P.; Fleche, J. L.; Lacroix, M.; Carra, O.; Dechelette, F.; Prele, G.; Rodriguez, G.

2012-01-01

Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO 2 interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

Energy Technology Data Exchange (ETDEWEB)

Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Lovera, P.; Fleche, J. L. [CEA, DEN, DPC Saclay, F-91191 Gif-sur-Yvette (France); Lacroix, M. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Carra, O. [AREVA / NP, 10 Rue Juliette Recamier, 69003 Lyon (France); Dechelette, F. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Prele, G. [EDF/SEPTEN, 12-14 avenue Dutrievoz, 69628 Villeurbane Cedex (France); Rodriguez, G. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France)

2012-07-01

Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO{sub 2} interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

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.)

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)

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

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.

Fast nuclear reactors. Associated international projects. State of the art and assessment of the concepts

International Nuclear Information System (INIS)

Azpitarte, O.; Ramilo, L.

2013-01-01

The recognition of the strategic importance of nuclear energy as a source of sustainable energy may be perceived in the continuous development, in many countries, of the technology of fast nuclear reactors with an associated closed fuel cycle, assuming that these Generation IV innovative systems will be required in the future. These reactors fulfill international requirements for safety and reliability, economic competitiveness, sustainability and proliferation resistance. They have the potential of using more efficiently the natural resources of Uranium and of reducing the volume and radiotoxicity of the nuclear waste by partitioning and transmutation of Minor Actinides. The national and international programs being carried out today are concentrated in the following concepts: Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), Gas Fast Reactor (GFR), Super Critical Water Reactor (SCWR) and Molten Salt Reactor (MSR). This article presents a short review of the technology of the mentioned concepts and details the current state of the main national and international related projects. (author)

Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

International Nuclear Information System (INIS)

Timothy Shaw; Anthony Baratta; Vaughn Whisker

2005-01-01

Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning

Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

Energy Technology Data Exchange (ETDEWEB)

Timothy Shaw; Anthony Baratta; Vaughn Whisker

2005-02-28

Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning.

Preliminary conceptual design of the secondary sodium circuit-eliminated JSFR (Japan Sodium Fast Reactor) adopting a supercritical CO2 turbine system (2). Turbine system and plant size

International Nuclear Information System (INIS)

Kisohara, Naoyuki; Sakamoto, Yoshihiko; Kotake, Shoji

2014-09-01

Research and development of the supercritical CO 2 (S-CO 2 ) cycle turbine system is underway in various countries for further improvement of the safety and economy of sodium-cooled fast reactors. The Component Design and Balance-Of-Plant (CD and BOP) of the Generation IV International Nuclear Forum (Gen-IV) has addressed this study, and their analytical and experimental results have been discussed between the relevant countries. JAEA, who is a member of the CD and BOP, has performed a design study of an S-CO 2 gas turbine system applied to the Japan Sodium-cooled Fast Reactor (JSFR). In this study, the S-CO 2 cycle turbine system was directly connected to the primary sodium system of the JSFR to eliminate the secondary sodium circuit, aiming for further economical improvement. This is because there is no risk of sodium-water reaction in the S-CO 2 cycle turbine system of SFRs. This report describes the system configuration, heat/mass balance, and main components of the S-CO 2 turbine system, based on the JSFR specifications. The layout of components and piping in the reactor and turbine buildings were examined and the dimensions of the buildings were estimated. The study has revealed that the reactor and turbine buildings could be reduced by 7% and 40%, respectively, in comparison with those in the existing JSFR design with the secondary sodium circuit employing the steam turbine. The cycle thermal was also calculated as 41.9-42.3%, which is nearly the same as that of the JSFR with the water/steam system. (author)

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.)

Preliminary Comparative Evaluation Study on Reference Design of GEN-IV SFR

Energy Technology Data Exchange (ETDEWEB)

Sim, Yoon Sub; Kim, Yeong Il; Hong, Ser Gi (and others)

2005-11-15

A fast reactor has a good transmutation capability and it enables breeding of fuel and use of a closed fuel cycle. By these characteristics of a fast reactor, the limited uranium resources of the world can be much more effectively utilized and the nuclear wastes of a high level of radioactivity and toxicity from the current nuclear power reactors of LWRs and HWRs can be drastically reduced in its volume and the management of the wastes can be easily treated. Also electricity can be generated more effectively since a fast reactor has the feature of high operation temperature. These features of a fast reactor makes it inevitable on a long term basis to construct fast reactors in Korea. The domestic fast reactor technology level, however, is at the level of coming out of a beginning stage and needs utilization of international expertise. Recently an international cooperation program called GIF has been formulated and our KALIMER was selected as one of the two reference designs for the international joint R and D works with JSFR of Japan. In the current frame of the GIF program, the two selected reference designs are supposed to be evaluated against each other in future and one design is to be finally selected. To make the international cooperation program directed more useful to our fast reactor technology development, it is required to strengthen the competitiveness of KALIMER so that it can be selected. To meet the necessity, a study was made in this research for pre-evaluation of the GIF reference designs and setting up plans for development of designs and technology that will enhance the competitiveness of KALIMER.

Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

Energy Technology Data Exchange (ETDEWEB)

Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

2010-09-28

The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

OpenAIRE

Verma, Vasudha

2017-01-01

Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

International Nuclear Information System (INIS)

Atkinson, C.A.; Collins, P.J.

1987-01-01

Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

Fast Breeder Reactor studies

International Nuclear Information System (INIS)

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

1980-07-01

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

Fast Breeder Reactor studies

Energy Technology Data Exchange (ETDEWEB)

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

1980-07-01

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

Optimal reactor strategy for commercializing fast breeder reactors

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

Fast reactor database. 2006 update

International Nuclear Information System (INIS)

2006-12-01

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

R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

2005-01-01

The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

Fast breeder reactors an engineering introduction

CERN Document Server

Judd, A M

1981-01-01

Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

Potential application of Rankine and He-Brayton cycles to sodium fast reactors

International Nuclear Information System (INIS)

Perez-Pichel, G.D.; Linares, J.I.; Herranz, L.E.; Moratilla, B.Y.

2011-01-01

Highlights: → This paper has been focused on thermal efficiency of several Rankine and Brayton cycles for SFR. → A sub-critical Rankine configuration could reach a thermal efficiency higher than 43%. → It could be increased to almost 45% using super-critical configurations. → Brayton cycles thermal performance can be enhanced by adding a super-critical organic fluid Rankine cycle. → The moderate coolant temperature at the reactor makes Brayton configurations have poorer. - Abstract: Traditionally all the demos and/or prototypes of the sodium fast reactor (SFR) technology with power output, have used a steam sub-critical Rankine cycle. Sustainability requirement of Gen. IV reactors recommends exploring alternate power cycle configurations capable of reaching high thermal efficiency. By adopting the anticipated working parameters of next SFRs, this paper investigates the potential of some Rankine and He-Brayton layouts to reach thermal efficiencies as high as feasible, so that they could become alternates for SFR reactor balance of plant. The assessment has encompassed from sub-critical to super-critical Rankine cycles and combined cycles based on He-Brayton gas cycles of different complexity coupled to Organic Rankine Cycles. The sub-critical Rankine configuration reached at thermal efficiency higher than 43%, which has been shown to be a superior performance than any of the He-Brayton configurations analyzed. By adopting a super-critical Rankine arrangement, thermal efficiency would increase less than 1.5%. In short, according to the present study a sub-critical layout seems to be the most promising configuration for all those upcoming prototypes to be operated in the short term (10-15 years). The potential of super-critical CO 2 -Brayton cycles should be explored for future SFRs to be deployed in a longer run.

Generation IV reactors and the ASTRID prototype: lessons from the Fukushima accident

International Nuclear Information System (INIS)

Gauche, F.

2012-01-01

In France, the ASTRID prototype is an industrial demonstrator of a sodium-cooled fast neutron reactor (SFR), fulfilling the criteria for Generation IV reactors. ASTRID will meet safety requirements as stringent as for third generation reactors, and it takes into account lessons from the Fukushima accident. The objectives are to reinforce the robustness of the safety demonstration for all safety functions. ASTRID will feature an innovative core with a negative sodium void coefficient, it will take advantage of the large thermal inertia of SFR for decay heat removal, and will provide for a design either eliminating the sodium-water reaction, or guaranteeing no consequences for safety in case such reaction would take place. (author)

SSTAR: The US lead-cooled fast reactor (LFR)

International Nuclear Information System (INIS)

Smith, Craig F.; Halsey, William G.; Brown, Neil W.; Sienicki, James J.; Moisseytsev, Anton; Wade, David C.

2008-01-01

It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV lead-cooled fast reactor system

Development of Core Heat Removal Objective Provision Trees for Sodium-Cooled Fast Reactor Defense-in-Depth Evaluation

Energy Technology Data Exchange (ETDEWEB)

Yang, Huichang; Kang, Bongsuk; Lee, Youngho [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

Based on the definition of Defense-in-Depth levels and safety functions for KALIMER sodium-cooled fast reactor, suggested in the reference and, OPTs for level 1, 2, and 3 defense-in-depth and core heat removal safety function, were developed and suggested in this paper. The purpose of this OPT is first to assure the defensein-depth design during the licensing of Sodium-Cooled Fast Reactors (SFR), but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of PGSFR (Prototype Gen-IV SFR) is not mature yet, the OPT is developed for KALIMER design. Developed OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defensein-depth evaluation frame for the regulatory reviews for the licensing process. At this moment, the identified provisions have both aspects as requirements and design features already adopted in KALIMER design. In the next stage of this study, derived provisions to be adopted will be compared with the actual design features and findings can be suggested as recommendations for the safety improvement.

Neutronic/Thermal-hydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

International Nuclear Information System (INIS)

Ragusa, Jean; Siegel, Andrew; Ruggieri, Jean-Michel

2010-01-01

The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

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.)

Transient Analysis Needs for Generation IV Reactor Concepts

International Nuclear Information System (INIS)

Siefken, L.J.; Harvego, E.A.; Coryell, E.W.; Davis, C.B.

2002-01-01

The importance of nuclear energy as a vital and strategic resource in the U. S. and world's energy supply mix has led to an initiative, termed Generation IV by the U.S. Department of Energy (DOE), to develop and demonstrate new and improved reactor technologies. These new Generation IV reactor concepts are expected to be substantially improved over the current generation of reactors with respect to economics, safety, proliferation resistance and waste characteristics. Although a number of light water reactor concepts have been proposed as Generation IV candidates, the majority of proposed designs have fundamentally different characteristics than the current generation of commercial LWRs operating in the U.S. and other countries. This paper presents the results of a review of these new reactor technologies and defines the transient analyses required to support the evaluation and future development of the Generation IV concepts. The ultimate objective of this work is to identify and develop new capabilities needed by INEEL to support DOE's Generation IV initiative. In particular, the focus of this study is on needed extensions or enhancements to SCDAP/RELAP5/3D code. This code and the RELAP5-3D code from which it evolved are the primary analysis tools used by the INEEL and others for the analysis of design-basis and beyond-design-basis accidents in current generation light water reactors. (authors)

4. generation sodium-cooled fast reactors. The ASTRID technological demonstrator

International Nuclear Information System (INIS)

2012-12-01

The sodium-cooled fast reactor (SFR) concept is one of the four fast neutron concepts selected by the Generation IV International Forum (GIF). SFRs have favourable technical characteristics and they are the sole type of reactor for which significant industrial experience feedback is available. After a discussion of the past experience gained on fast breeder reactors in the world (benefits, difficulties and problematics), the authors discuss the main improvement domains and the associated R and D advances (reactor safety, prevention and mitigation of severe accidents, the sodium-water risk, detection of sodium leaks, increased availability, instrumentation and inspection, control and repairability, assembly handling and washing). Then, they describe the technical requirements and safety objectives of the ASTRID experimental project, notably with its reactivity management, cooling management, and radiological containment management functions. They describe and discuss requirements to be met and choices made for Astrid, and the design options for its various components (core and fuels, nuclear heater, energy conversion system, fuel assembly handling, instrumentation and in-service inspection, control and command). They present the installations which are associated with the ASTRID cycle, evoke the development and use of simulations and codes, describe the industrial organization and the international collaboration about the ASTRID project, present the planning and cost definition

Condensation and homogenization of cross sections for the deterministic transport codes with Monte Carlo method: Application to the GEN IV fast neutron reactors

International Nuclear Information System (INIS)

Cai, Li

2014-01-01

In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core

Fast reactor shield sensitivity studies for steel--sodium--iron systems

International Nuclear Information System (INIS)

Oblow, E.M.; Weisbin, C.R.

1977-01-01

A study was made of the adequacy of the current ENDF/B-IV sodium and iron neutron cross section data files for fast reactor shield design work. Experimental data from 21 fast reactor shield configurations containing large thicknesses of steel, sodium, and iron were analyzed with discrete ordinates calculations and sensitivity methods to assess the data files. This study represents the largest full-scale sensitivity analysis of benchmark quality experimental data to date. Included in the sensitivity studies were the results of the new cross section adjustment algorithms added to the FORSS code system. Conclusions were drawn about the need for more accurate data for sodium and iron elastic and discrete inelastic cross sections above 1 MeV and the values of the total cross section in the vicinity of important minima

Status of the French R/D program on the severe accident issue to develop Gen IV SFRs - 15373

International Nuclear Information System (INIS)

Serre, F.; Bertrand, F.; Journeau, C.; Suteau, C.; Verwaede, D.; Schmitt, D.; Farges, B.

2015-01-01

The ASTRID reactor (Advanced Sodium Technological Reactor for Industrial Demonstration) is a technological demonstrator designed by the CEA with its industrial partners, with very high levels of requirements. In the ASTRID project, the safety objectives are to prevent core melting, in particular by the development of an innovative core with complementary safety prevention devices, and to enhance the reactor resistance to severe accident by design. To mitigate the consequences of hypothetical core melting situations, specific dispositions or mitigation devices will be added to the core and to the reactor. It is also required to provide a robust safety demonstration (with high level of confidence). Therefore a new approach for severe accident issue has been defined: to the well-known 'lines of defense' method, a 'lines of mitigation' method is added. To meet these ASTRID, or future SFR, requirements, a large R/D program was launched in the Severe Accident domain, with a large number of partners. This paper will present the status of the CEA R/D related to the SFR Severe Accident issue, the collaboration framework (with industrial partners and R/D foreign organizations), and the future R/D plans to support the ASTRID project and possible developments for future Gen IV commercial SFR. (authors)

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)

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

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

International Nuclear Information System (INIS)

Friess, Friederike Renate

2017-01-01

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

Energy Technology Data Exchange (ETDEWEB)

Friess, Friederike Renate

2017-07-12

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Fast Reactor Research in Europe: The Way Towards Sustainability (Summary)

International Nuclear Information System (INIS)

Schenkel, R.

2012-01-01

Full text: The European Union (EU) has taken the lead in responding to climate change, announcing far-reaching initiatives ranging from promoting energy efficient light bulbs and cars to new building codes, carbon trading schemes, development of low carbon technologies and greater competition in energy markets. Nuclear energy remains central to the energy debate in Europe. One third of EU electricity is produced via nuclear fission and eight new reactors are under construction. Traditionally non-nuclear countries are manifesting an interest in building nuclear power plants while the clock is ticking down on Belgium, Germany and the United Kingdom's decision to renew or close existing nuclear infrastructures. Sustainability in nuclear energy production is ensured in the medium term as a result of the large and diverse uranium resources available in politically stable countries around the world. The quantities available with high probability ensure more than one hundred years of nuclear energy production. This extrapolation depends, however, on the forecast for future nuclear energy production. The use of fast neutron breeder reactors would lead to a much more efficient utilization of the uranium, extending the sustainable energy production to several thousands of years. The presentation will outline the fast reactors of the new generation currently being developed within the Generation IV initiative. Broad conclusions of the presentation are that: - There is a growing nuclear renaissance in Europe for good reason; - Nuclear energy is a green and sustainable option for Europe and indeed the world's energy needs; - Nuclear energy is a competitive energy that makes economic sense; - Nuclear fission reactors have a safety and environmental track record that is second to none, yet public misperceptions persist and must be tackled; - Waste management solutions exist while new developments hold great promise; - The evolution and promise of nuclear technologies must also be

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)

The fast breeder reactor

International Nuclear Information System (INIS)

Patterson, W.

1990-01-01

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

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.)

Key Factors for the Linkage Strategy between R and D and Commercialization for Gen-ΙV

International Nuclear Information System (INIS)

Lee, Kyoungmi; Hong, Jung Suk

2013-01-01

The Fukushima nuclear disaster has leaded to enhance the safety and the cost-effectiveness of technology for the future so that advanced countries such as United Sates and France have concerned about a next generation nuclear power plant, Gen-IV(Generation-IV Reactor). Considering various characteristics of nuclear R and D, it is necessary to have more elaborated strategies for the effective development of the next generation of nuclear technology. In this study, we suggest 5 key factors for the successful commercialization of Gen-IV by analyzing the distinct characteristics of nuclear R and D with Gen-IV and CSF(Critical Success Factor)s of several cases in these field and conducting the FGI(Focus Group Interview). Considering these results, we could find and suggest some important points for further strategy for Gen-IV. That is, following five key factors for the linkage improvement between R and D and commercialization of Gen-IV should be considered: the participation of nuclear power plant operators from the beginning, the establishment of consistent and comprehensive plan/roadmap/detailed strategy, the technology development based on global energy issues and international cooperation, the stable and clear funding plans for long-term projects, the cooperation of relative ministries. Gen-IV system is getting a positive response in that it accompanies long-term R and D plans in Korea. We think that the standard of Gen-IV would lead the next generation of nuclear industry if the proper strategy for the cooperation between the private sector and the regulation from the beginning. Moreover, we expect that this study will facilitate its development process from R and D to commercialization

Key Factors for the Linkage Strategy between R and D and Commercialization for Gen-ΙV

Energy Technology Data Exchange (ETDEWEB)

Lee, Kyoungmi; Hong, Jung Suk [Korean Institute of S and T Evaluation and Planning, Seoul (Korea, Republic of)

2013-05-15

The Fukushima nuclear disaster has leaded to enhance the safety and the cost-effectiveness of technology for the future so that advanced countries such as United Sates and France have concerned about a next generation nuclear power plant, Gen-IV(Generation-IV Reactor). Considering various characteristics of nuclear R and D, it is necessary to have more elaborated strategies for the effective development of the next generation of nuclear technology. In this study, we suggest 5 key factors for the successful commercialization of Gen-IV by analyzing the distinct characteristics of nuclear R and D with Gen-IV and CSF(Critical Success Factor)s of several cases in these field and conducting the FGI(Focus Group Interview). Considering these results, we could find and suggest some important points for further strategy for Gen-IV. That is, following five key factors for the linkage improvement between R and D and commercialization of Gen-IV should be considered: the participation of nuclear power plant operators from the beginning, the establishment of consistent and comprehensive plan/roadmap/detailed strategy, the technology development based on global energy issues and international cooperation, the stable and clear funding plans for long-term projects, the cooperation of relative ministries. Gen-IV system is getting a positive response in that it accompanies long-term R and D plans in Korea. We think that the standard of Gen-IV would lead the next generation of nuclear industry if the proper strategy for the cooperation between the private sector and the regulation from the beginning. Moreover, we expect that this study will facilitate its development process from R and D to commercialization.

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

Contribution of the IV generation fast reactors to the sustainable development

International Nuclear Information System (INIS)

Mendoza G, G.; Klapp E, J.L.

2007-01-01

During the XXI century all the energy forms are necessary for the sustainable development. A balanced energy politics has to use a mixture of energy sources that completes the objective of responding to the increase in the demand and that it uses non emitting gases sources of greenhouse effect like the nuclear one. It is evident the great existent difficulty to turn the objectives of emissions for the coming years without having the nuclear energy. Later on, the process continued outlining serious commitments among the development necessity, the improvement of the level of life and the competitiveness, and the execution from the established environmental requirements to world level. It is very foregone that the energy nuclear become the best energy source to improve the environmental conditions and that new initiatives are determined in those that this energy will have an important paper. The solution is to build a nuclear central of advanced design, using technologies that its help to brake the diffusion of the nuclear weapons. The nucleo electric energy at great scale should be developed on the base of designs of reactors and innovative processes of fuel that can lend technological support to the not nuclear proliferation regime, and that at the same time they contribute to satisfy the electricity demand in the world. In a scenario of increase of energy demand, mainly in the development countries, and of growing interest in the pollutants reduction originated by the use of fossil fuels, the nuclear reactors of IV Generation arise as proposal and challenge. Meanwhile the search of new technologies and innovations become imperative, translating an enormous evolution, not only in the conceptual projects, as well as in the fuel cycle so that, in a scenario of open economy, turn its more competitive. Inside the reactors of fourth generation, the quick reactors are configured as those that more assist to such demands and they will be, without a doubt, the reactors in

Effects of Ta addition on the Microstructural and Mechanical Properties of 9Cr-0.5Mo-2W F/M Steels for a SFR Fuel Cladding

International Nuclear Information System (INIS)

Baek, Jong-Hyuk; Han, Chang-Hee; Kim, Tae-Kyu; Kim, Sung-Ho; Lee, Chan-Bock

2007-01-01

Today twenty fission reactors provide about 40% of the domestic electricity supply. The world-wide distribution of some nuclear reactors will be aging and will need replacement and enhancement to both keep pace with and to take up a large share of the growing world-wide electricity demand. A new generation (Gen IV) of nuclear plant concepts has become the focus of international advanced reactor activity. Gen IV nuclear systems embodies greater improvements and innovative advances in technology over earlier ones. The Gen IV systems are to have a considerable increase in safety and be economically competitive when compared with the existed commercial reactors. In particular, the systems should produce a significantly reduced volume of nuclear wastes. From this point of view, sodium-cooled Fast Reactor (SFR) is strongly considered as a future nuclear energy system in Korea

MYRRHA a multi-purpose hybrid research reactor for high-tech applications

International Nuclear Information System (INIS)

Abderrahim, H. A.; Baeten, P.

2012-01-01

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental accelerator driven system (ADS) in development at SCK-CEN. MYRRHA is able to work both in subcritical (ADS) as in critical mode. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for generation IV (GEN IV) systems, material developments for fusion reactors, radioisotope production and industrial applications, such as Si-doping. MYRRHA will also demonstrate the ADS full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. MYRRHA is based on the heavy liquid metal technology and so it will contribute to the development of lead fast reactor (LFR) technology and in critical mode, MYRRHA will play the role of European technology pilot plant in the roadmap for LFR. In this paper the historical evolution of MYRRHA and the rationale behind the design choices is presented and the latest configuration of the reactor core and primary system is described. (authors)

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.

Physics of Fast and Intermediate Reactors. V. I. Proceedings of the Seminar on the Physics of Fast and Intermediate Reactors. V. I

Energy Technology Data Exchange (ETDEWEB)

NONE

1962-03-15

It is generally agreed that the ultimate economic advantage of power produced by nuclear fission over that produced by conventional sources depends on the ability of a certain type of reactor to breed precious nuclear fuel out of the plentiful but not readily fissionable isotope of uranium. This fact is mainly responsible for the importance attached to the development of fast power reactors, but many other interesting properties of unmoderated or weakly moderated reactor systems have also been brought to light by reactor physicists. In August 1961 the Agency organized in Vienna a Seminar on the Physics of Past and Intermediate Reactors, at which all the topics relating to this important branch, of reactor science were discussed. The main feature of this meeting was extensive discussion of the 66 written contributions, which set the stage for a wide exchange of experience and ideas throughout 13 half-day sessions. The Seminar was attended by 132 scientists from 22 Member States and two international organizations. It is hoped that these Proceedings of the Seminar, which include both the papers presented and a record of the discussions, will be useful as a reference work both to research workers in the field and to newcomers to it for many years to come. The Agency's thanks are due to all the participating scientists for their written or oral contributions and especially to those among them who, as session chairmen, led the discussions and contributed greatly to the success of the meeting. During the Seminar, sixty-five papers were orally presented, and seven more were accepted for publication in the Proceedings. In order that these Proceedings might be in the hands of their users at an early date, the method of presentation of the papers and of the extensive session discussions had to be somewhat different from the one usually followed. The complete record of the sessions will be found at the end of Volume III. The order in which the papers are presented here is not

Improvement of computer programs 'BAMBOO' and 'ASFRE-IV' for coupling analysis of deformation and thermal-hydraulics in a high burn-up fuel subassembly of fast reactor

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ohshima, Hiroyuki; Imai, Yasutomo

2003-04-01

A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBOO. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly. (author)

Argentinean activities related to Fast Reactors

International Nuclear Information System (INIS)

Azpitarte, Osvaldo

2012-01-01

CNEA objectives in the area of Generation IV nuclear reactors: Implement a programme for the monitoring of the global progress of new technologies for Generation IV nuclear reactors and their fuel cycles, in order to generate and assess associated lines of R&D. – Perform studies and evaluations for defining the Generation IV line or lines on which CNEA would be interested; – Promote the participation on specific international projects; – Implementation of experimental facilities

Safeguards challenges of Fast Breeder Reactor

International Nuclear Information System (INIS)

Ko, H. S.

2010-01-01

Although the safeguards system of Sodium Fast Reactor (SFR) seems similar to that of Light Water Reactor (LWR), it was raised safeguards challenges of SFR that resulted from the visual opacity of liquid sodium, chemical reactivity of sodium and other characteristics of fast reactor. As it is the basic concept stage of the safeguards of SFR in Korea, this study tried to analyze the latest similar study of safeguards issues of the Fast Breeder Reactor (FBR) at Joyo and Monju in Japan. For this reason, this study is to introduce some potential safeguards challenges of Fast Breeder Reactor. With this analysis, future study could be to address the safeguards challenges of SFR in Korea

Simulation tools and new developments of the molten salt fast reactor

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Doligez, X.; Heuer, D.; Allibert, M.; Ghetta, V.

2010-01-01

Starting from the Molten Salt Breeder Reactor project of Oak-Ridge, we have performed parametric studies in terms of safety coefficients, reprocessing requirements and breeding capabilities. In the frame of this major re-evaluation of the molten salt reactor (MSR), we have developed a new concept called Molten Salt Fast Reactor or MSFR, based on the Thorium fuel cycle and a fast neutron spectrum. This concept has been selected for further studies by the MSR steering committee of the Generation IV International Forum in 2009. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for materials evolution which resolves the Bateman's equations giving the population of each nucleus inside each part of the reactor at each moment. Because of MSR's fundamental characteristics compared to classical solid-fuelled reactors, the classical Bateman equations have to be modified by adding two terms representing the reprocessing capacities and the fertile or fissile alimentation. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system (reactor and reprocessing unit) and radioprotection issues. (authors)

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

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)

Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-08-15

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

The Atalante facility at CEA/Marcoule: towards Gen IV systems fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Bordier, Gilles; Warin, Dominique; Masson, Michel [CEA/Marcoule Direction, BP 17171 - 30207 - Bagnols-sur-Ceze Cedex (France)

2008-07-01

The Atalante facility is a complete set of 18 hot labs and 9 shielded cells devoted to the research and development on fuel cycle. The activities correspond to 4 major sectors of nuclear research: -) to support the operation of actual reprocessing plants with the aim of adapting the head of the process to the increase of the spend fuel burn-up and to different types of new burnt fuels to be reprocessed (including MOX, USi or UMo fuels); -) to develop the COEX{sup TM} process that jointly manages uranium and plutonium from the dissolution of spent fuel to the production of UPuO{sub 2} powder and the fabrication of MOX fuel pellets; -) to prepare the recycling of minor actinides (MA) by partitioning or by grouped actinide extraction, and by MA bearing fuel fabrication; -) to study the long term behavior of high level waste conditioning matrices and especially self irradiation and leaching of vitrified waste. The first hot lab of Atalante was operated in 1992, the process shielded cell (CBP) in 2003 and the last LN1 lab in 2005, while at the same time a large scale demonstration test on the DIAMEX-SANEX MA partitioning process was performed. Now some new challenges involve further necessary evolutions of the facility. Some are related to safety assessment and operating flexibility; the major evolutions will come from new scientific goals and research programs. Furthermore, minor actinides materials irradiation tests in fast reactors will be prepared in the framework of a large international cooperation (GACID program) and need the production of significant amounts of MA bearing mixed U-Pu oxide compounds in new hot labs and shielded cells equipment. The major new research tools are presented and we highlight how Atalante is a unique facility which brings a real opportunity to reinforce the European and international scientific cooperation in order to prepare the next Gen IV fuel cycle. (authors)

The Atalante facility at CEA/Marcoule: towards Gen IV systems fuel cycle

International Nuclear Information System (INIS)

Bordier, Gilles; Warin, Dominique; Masson, Michel

2008-01-01

The Atalante facility is a complete set of 18 hot labs and 9 shielded cells devoted to the research and development on fuel cycle. The activities correspond to 4 major sectors of nuclear research: -) to support the operation of actual reprocessing plants with the aim of adapting the head of the process to the increase of the spend fuel burn-up and to different types of new burnt fuels to be reprocessed (including MOX, USi or UMo fuels); -) to develop the COEX TM process that jointly manages uranium and plutonium from the dissolution of spent fuel to the production of UPuO 2 powder and the fabrication of MOX fuel pellets; -) to prepare the recycling of minor actinides (MA) by partitioning or by grouped actinide extraction, and by MA bearing fuel fabrication; -) to study the long term behavior of high level waste conditioning matrices and especially self irradiation and leaching of vitrified waste. The first hot lab of Atalante was operated in 1992, the process shielded cell (CBP) in 2003 and the last LN1 lab in 2005, while at the same time a large scale demonstration test on the DIAMEX-SANEX MA partitioning process was performed. Now some new challenges involve further necessary evolutions of the facility. Some are related to safety assessment and operating flexibility; the major evolutions will come from new scientific goals and research programs. Furthermore, minor actinides materials irradiation tests in fast reactors will be prepared in the framework of a large international cooperation (GACID program) and need the production of significant amounts of MA bearing mixed U-Pu oxide compounds in new hot labs and shielded cells equipment. The major new research tools are presented and we highlight how Atalante is a unique facility which brings a real opportunity to reinforce the European and international scientific cooperation in order to prepare the next Gen IV fuel cycle. (authors)

A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

International Nuclear Information System (INIS)

Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

2010-01-01

Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks. (authors)

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

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

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.

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)

Status of National Programmes on Fast Breeder Reactors. International Working Group on Fast Reactors Twenty-First Annual Meeting, Seattle, USA, 9-12 May 1988

International Nuclear Information System (INIS)

1988-11-01

The following papers on the status of national programmes on fast breeder reactors are presented in this report: Fast breeder reactor development in France during 1987; Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands; A review of the Indian fast reactor programme; A review of the Italian fast reactor programme; A review of the fast reactor programme in Japan; Status of fast reactor activities in the USSR; A review of the United Kingdom fast reactor programme; Status of liquid metal reactor development in the United States of America; Review of activities of the Commission of European Communities relating to fast reactors in 1987; European co-operation in the field of fast reactor research and development — 1987 progress report; A review of fast reactor activities in Switzerland

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.)

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.)

Fast breeder reactors--lecture 4

International Nuclear Information System (INIS)

Marshall, W.; Davies, L.M.

1986-01-01

This paper discusses the economics of fast breeder reactors. An algebraic background is presented which represents the various views expressed by different nations regarding the cost of fast breeder reactors and their associated fuel cycle services, the timescale by which they might be available, and the simultaneous variations in the price of uranium. Actual presentations made by individual countries in recent discussions serve to verify the general nature of this present discussion. It is assumed that if nuclear power is to make a long term contribution to the needs of the world, the introduction of fast breeder reactors is both essential and necessary

Best-practices guidelines for L2PSA development and applications. Volume 2 - Best practices for the Gen II PWR, Gen II BWR L2PSAs. Extension to Gen III reactors

International Nuclear Information System (INIS)

Raimond, E.; Durin, T.; Rahni, N.; Meignen, R.; Cranga, M.; Pichereau, F.; Bentaib, A.; Guigueno, Y.; Loeffler, H.; Mildenberger, O.; Lajtha, G.; Santamaria, C.S.; Dienstbier, J.; Rydl, A.; Holmberg, J.E.; Lindholm, I.; Maennistoe, I.; Pauli, E.M.; Dirksen, G.; Grindon, L.; Peers, K.; Hulqvist, G.; Parozzi, F.; Polidoro, F.; Cazzoli, E.; Vitazkova, J.; Burgazzi, L.; Oury, L.; Ngatchou, C.; Siltanen, S.; Niemela, I.; Routamo, T.; Helstroem, P.; Bassi, C.; Brinkman, H.; Seidel, A.; Schubert, B.; Wohlstein, R.; Guentay, S.; Vincon, L.

2010-01-01

The objective of this coordinated action was to develop best practice guidelines for the performance of Level 2 PSA methodologies with a view of harmonisation at EU level and to allow meaningful and practical uncertainty evaluations in a Level 2 PSA. Specific relationships with community in charge of nuclear reactor safety (utilities, safety authorities, vendors, and research or services companies) have been established in order to define the current needs in terms of guidelines for level 2 PSA development and applications. An international workshop was organised in Hamburg, with the support of VATTENFALL, in November 2008. The level 2 PSA experts from the ASAMPSA2 project partners have proposed some guidelines for the development and application of L2PSA based on their experience and on information available from international cooperation (EC Severe Accident network of Excellence - SARNET, IAEA standards, OECD-NEA publications and workshop) or open literature. The number of technical issues addressed in the guideline is very large and all are not covered with the same relevancy in the first version of the guideline. This version is submitted for external review in November 2010 by severe accident experts and PSA, especially, from SARNET and OECD-NEA members. The feedback of the external review will be dis cussed during an international open works hop planned in March 2011 and all outcomes will be taken into consideration in the final version of this guideline (June 2011). The guideline includes 3 volumes: - Volume 1 - General considerations on L2PSA. - Volume 2 - Technical recommendations for Gen II and III reactors. - Volume 3 - Specific considerations for future reactor (Gen IV). The recommendations formulated in the guideline should not be considered as 'mandatory' but should help the L2PSA developers to achieve high quality studies with limited time and resources. It may also help the L2PSA reviewers by positioning one specific study in comparison with some

Astrid (fast breeder nuclear reactor)

International Nuclear Information System (INIS)

2014-01-01

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

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

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

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Primary system thermal hydraulics of future Indian fast reactors

Energy Technology Data Exchange (ETDEWEB)

Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

2015-12-01

Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

Development of Basic Key Technologies for Gen IV SFR Safety Evaluation

International Nuclear Information System (INIS)

Jeong, Hae Yong; Kwon, Young Min; Kim, Tae Woon; Park, Soo Yong; Suk, Soo Dong; Lee, Kwi Lim; Lee, Yong Bum; Chang, Won Pyo; Ha, Kwi Seok; Hahn, Sang Hoon

2010-07-01

Safety issues and design requirements on control rod worth were identified through the evaluation of safety design characteristics and the preliminary safety evaluation. This results will be taken into account for the conceptual design studies of the demonstration reactor in the next stage. The Level-1 Pasa has been performed and a quantitative Cdf value was produced for the selected design from the several candidates. The inherent safety characteristics of the selected design were evaluated through the DBE and ATWS analyses. A surrogate material for Tru has been selected which is applicable to the study of liquidus/solidus temperature test for the metallic fuel containing Tru. A methodology for the regression analysis with surrogate material has been developed and valuable data on metal fuel liquidus/solidus temperature have been measured. A simple mechanistic model describing a bending of subassemblies has been formulated based on the foreign test data and existing models. Its applicability has been evaluated for the Phenix design. New criteria of the core damage for the SFR PSA were identified. The list of initiating events, system response event tree, and core response event tree, which constitute a PSA methodology for an SFR, have been introduced. By developing the SFR PIRT, phenomenological model features, which have to be satisfied in a safety code, were defined and the PIRT results were applied to the design of the PDRC test facility. Bases for a safety evaluation methodology for the SFR DBEs have been also prepared. A draft version of the topical report on the code for local fault analysis has been completed. Since 2007, the MARS-LMR code has been developed and assessments for model validation with the test data from EBR-II and Phenix reactor have been continued. The code has been applied to the evaluation of passive safety of a conceptual design of Gen IV SFR

The fast breeder reactor

International Nuclear Information System (INIS)

Keck, O.

1984-01-01

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

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

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

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)

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)

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.)

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.

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.

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.

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.

Experimental facilities for Generation IV reactors research

International Nuclear Information System (INIS)

Krecanova, E.; Di Gabriele, F.; Berka, J.; Zychova, M.; Macak, J.; Vojacek, A.

2013-06-01

Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

Safeguards Considerations for the Design of a Future Fast Neutron Sodium Cooled Reactor

International Nuclear Information System (INIS)

Cazalet, J.; Raymond, P.; Masson, M.; Saturnin, A.

2015-01-01

Incorporating safeguards at an early stage of a reactor design is a way to increase the effectiveness and efficiency of safeguards measures minimizing the possibilities of misuse of the plant or nuclear material diversion. It also reduces the impact on the construction and operation cost. At the preliminary phase, the design will integrate: confinement, containment, surveillance features and non-destructive assay equipment. Taking into account these requirements will help the operator in the approval of the plant at the design phase by national and international authorities in charge of Nuclear Material accounting and safeguards. A large amount of work has been made by the GEN IV International Forum to assess the proliferation resistance of nuclear systems. The IAEA has developed guidelines on ''Safeguards by design'' describing reference requirements for future nuclear facilities. Based on these studies, this communication details implementation of safeguards in the design of a sodium cooled fast neutron reactor (SFR) currently studied in France. Specificities are the use of MOX fuel with high concentration of plutonium and the potential capacity of breeding. A great attention should be paid to avoid diversion of nuclear material contained in fresh or irradiated fuel. Scenarios of reactor misuse are analyzed. The identification of diversion pathways and requirements for nuclear material accountancy, leads to an approach of safeguards, specific to SFR: Material Balance Areas (MBA) and some key measurement points (KMP) are characterized. Specific instrumentation assay helping in the identification and/or characterization of fuel elements and the inventory of nuclear material is described. As concerns the fuel cycle, the safeguards of the reprocessing unit will be progressively increased through the development of materials monitoring and the implementation of these measures at strategic locations of buildings, thus providing real-time information

The irradiation test program for transmutation in the French Phenix fast reactor

International Nuclear Information System (INIS)

Guidez, J.; Chaucheprat, P.; Fontaine, B.; Brunon, E.

2004-01-01

Put on commercial operation in July 1974, the French fast reactor Phenix reached a 100 000 hours operation time in september 2003. When the French law relative to long lived radioactive waste management was promulgated on December 1991, priority was given to Phenix to be run as a research reactor and to carry on a wide irradiation program dedicated to study transmutation of minor actinides and long-lived fission products. After a major renovation program required to extend the reactor lifetime, Phenix power buildup took place in 2003. Experimental irradiations have been loaded in the core, involving components for heterogeneous and homogeneous transmutation modes, americium targets, technetium 99 metal pins and isolated isotopes for integral cross-sections measurements. Associated post- irradiated examination programs are already underway or planned. With new experiments to be loaded in the core in 2006 the Phenix reactor remains to be a powerful tool providing an important experimental data on fast reactors and on transmutation of minor actinides and long-lived fission products, as well as it will contribute to gain further experience in the framework of the GENERATION IV International Forum. (authors)

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

Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

Directory of Open Access Journals (Sweden)

Sungjoo Lee

2016-09-01

Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

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

International Nuclear Information System (INIS)

2012-12-01

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

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

Fast mixed spectrum reactor concept

International Nuclear Information System (INIS)

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

1979-04-01

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

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

Development of an advanced code system for fast-reactor transient analysis

International Nuclear Information System (INIS)

Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

2005-01-01

FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)

Development of alternate extractant systems for fast reactor fuel cycle

International Nuclear Information System (INIS)

Vasudeva Rao, P.R.; Suresh, A.; Venkatesan, K.A.; Srinivasan, T.G.; Raj, Baldev

2007-01-01

Due to the limitations of TBP in processing of high burn-up, Pu-rich fast reactor fuels, there is a need to develop alternate extractants for fast reactor fuel processing. In this context, our Centre has been examining the suitability of alternate tri-alkyl phosphates. Third phase formation in the extraction of Th(IV) by TBP, tri-n-amyl phosphate (TAP) and tri-2-methyl-butyl phosphate (T2MBP) from nitric acid media has been investigated under various conditions to derive conclusions on their application for extraction of Pu at macro levels. The chemical and radiolytic degradation of tri-n-amyl-phosphate (TAP) diluted in normal paraffin hydrocarbon (NPH) in the presence of nitric acid has been investigated by the measurement of plutonium retention in organic phase. The potential application of room temperature ionic liquids (RTILs) for reprocessing of spent nuclear fuel has been explored. Extraction of uranium (VI) and palladium (II) from nitric acid medium by commercially available RTIL and tri-n-butyl phosphate solution in RTIL have been studied and the feasibility of electrodeposition of uranium as uranium oxide (UO 2 ) and palladium (II) as metallic palladium from the loaded organic phase have been demonstrated. This paper describes results of the above studies and discusses the suitability of the systems for fast reactor fuel reprocessing. (authors)

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

The web-enabled database of JRC-EC, a useful tool for managing European Gen IV materials data

International Nuclear Information System (INIS)

Over, H.H.; Dietz, W.

2008-01-01

Materials and document databases are important tools to conserve knowledge and experimental materials data of European R and D projects. A web-enabled application guarantees a fast access to these data. In combination with analysis tools the experimental data are used for e.g. mechanical design, construction and lifetime predictions of complex components. The effective and efficient handling of large amounts of generic and detailed materials data with regard to properties related to e.g. fabrication processes, joining techniques, irradiation or aging is one of the basic elements of data management within ongoing nuclear safety and design related European research projects and networks. The paper describes the structure and functionality of Mat-DB and gives examples how these tools can be used for the management and evaluation of materials data of European (national or multi-national) R and D activities or future reactor types such as the EURATOM FP7 Generation IV reactor types or the heavy liquid metals cooled reactor

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

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

Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

International Nuclear Information System (INIS)

Scari, Maria Elizabeth

2017-01-01

Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

Effect of Proton Irradiation on the Corrosion Behaviors of Ferritic/Martensitic Steel in Liquid Metal Environment

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeonghyeon; Kim, Tae Yong; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

2016-10-15

Liquid metal fast breeder reactors (LMFBRs) such as sodium-cooled fast reactor (SFR) and lead-cooled fast reactor (LFR) are the candidates of GEN-IV nuclear energy systems. Among various liquid metals that can be used as primary coolant material, sodium is a world widely used coolant for GEN-IV reactors. In this study, as-received Gr.92 and irradiated Gr.92 specimen in the oxygen-saturated liquid sodium were examined at high temperature for 300h. The microstructure results reveal the information of the effect of irradiation and effect of the chrome concentration in specimen. From the SRIM result, penetration distance of 40 μm in stainless steel and nominal sample thickness of 30 μm was used to avoid the damage peak and any proton implantation and From the microstructural evaluation, chromium-rich zones existed under the surface of the both of non-irradiated and irradiated materials. The irradiated materials showed chromium-rich zones with larger depths than the non-irradiated specimens.

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

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)

History of fast reactor development in U.S.A.-I

International Nuclear Information System (INIS)

Ninokata, Hisashi; Sasao, Nobuyki

2007-01-01

History and present state of fast reactor was reviewed in series. As a history of fast reactor development in U.S.A. - I, this third lecture presented the dawn of the fast reactor development in the USA. The first fast reactor was the Clementine reactor with plutonium fuels and mercury coolant. The LAMPRE-1 reactor was the first sodium cooled and molten plutonium reactor. Experimental breeder reactor (EBR-1) was the first reactor to produce electricity and four kinds of fuels were loaded. Zero-power reactors were constructed to conduct reactor physics experiments on fast reactors. Today there are renewed interests in fast reactors due to their ability to fission actinides and reduce radioactive wastes. (T. Tanaka)

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

International Nuclear Information System (INIS)

1968-01-01

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

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.)

High temperature fast reactor for hydrogen production in Brazil; Reator nuclear rapido de altissima temperatura para producao de hidrogenio no Brasil

Energy Technology Data Exchange (ETDEWEB)

Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados]. E-mail: jamil@ieav.cta.br

2008-07-01

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

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

Validation of iron nuclear data for the neutron calculation of nuclear reactors

International Nuclear Information System (INIS)

Vaglio-Gaudard, C.

2010-01-01

The GEN-III and GEN-IV reactors will be equipped with heavy reflectors. However, the existing integral validation of the iron nuclear data in the latest JEFF3 European library in the frame of the neutron calculation of the heavy reflector is very partial: some results exist concerning fast reactors but there is no result corresponding to the LWR heavy reflector. No clear trend on the JEFF3 iron cross sections was brought into evidence up to now for fission reactor calculations. Iron nuclear data were completely re-evaluated in the JEFF3 library. Despite the fact that iron is widely used in the nuclear industry, large uncertainties are still associated with its nuclear data, particularly its inelastic cross section which is very important in the neutron slowing down. A validation of 56 Fe nuclear data was performed on the basis of the analysis of integral experiments. Two major critical experiments, the PERLE experiment and the Gas Benchmark, were interpreted with 3D reference Monte-Carlo calculations and the JEFF3.1.1 library. The PERLE experiment was recently performed in the EOLE zero-power facility (CEA Cadarache). This experiment is dedicated to heavy reflector physics in GEN-III light water reactors. It was especially conceived for the validation of iron nuclear data. The Gas Benchmark is representative of a Gas Fast Reactor with a stainless steel reflector (with no fertile blanket) in the MASURCA facility (CEA Cadarache). Radial traverses of reaction rates were measured to characterize flux attenuation at various energies in the reflector. The results of the analysis of both experiments show good agreement between the calculations and the measurements, which is confirmed by the analysis of complementary experiments (ZR-6M, MISTRAL4, CIRANO-ZONA2B). A process of re-estimating the 56 Fe nuclear data was implemented on the basis of feedback from these two experiments and the RDN code. This code relies on a non-linear regression method using an iterative

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

Transient Safety Analysis of Fast Spectrum TRU Burning LWRs with Internal Blankets

Energy Technology Data Exchange (ETDEWEB)

Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Zazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hill, Bob [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-01-31

The objective of this proposal was to perform a detailed transient safety analysis of the Resource-Renewable BWR (RBWR) core designs using the U.S. NRC TRACE/PARCS code system. This project involved the same joint team that has performed the RBWR design evaluation for EPRI and therefore be able to leverage that previous work. And because of their extensive experience with fast spectrum reactors and parfait core designs, ANL was also part the project team. The principal outcome of this project was the development of a state-of-the-art transient analysis capability for GEN-IV reactors based on Monte Carlo generated cross sections and the US NRC coupled code system TRACE/PARCS, and a state-of-the-art coupled code assessment of the transient safety performance of the RBWR.

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)

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

Energy Technology Data Exchange (ETDEWEB)

Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-05-15

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

International Nuclear Information System (INIS)

Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-01-01

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

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)

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

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.)

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

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

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)

Objective Provision Trees of Reactivity Control Safety Function for Sodium-Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Kang, Bongsuk; Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-05-15

The purpose of this OPT is first to assure the DiD design during the licensing of Sf, but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). Based on the definition of Defense-in-Depth (DiD) levels and safety functions for KALIMER Sodium-Cooled Fast Reactor (SFR), suggested in the reference and, Objective Provision Trees (OPTs) of reactivity control function for level 1, 2, 3 and 4 DiD were developed and suggested in this paper. The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of Prototype Gen-IV Sf (PGSFR) is not mature yet, the OPT is developed for KALIMER design. Developed level 1 to 4 OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defense-in-depth evaluation frame for the regulatory reviews for the licensing process. In the next stage of this study, other safety function will be researched and findings can be suggested as recommendations for the safety improvement.

Objective Provision Trees of Reactivity Control Safety Function for Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Kang, Bongsuk; Yang, Huichang; Suh, Namduk

2014-01-01

The purpose of this OPT is first to assure the DiD design during the licensing of Sf, but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). Based on the definition of Defense-in-Depth (DiD) levels and safety functions for KALIMER Sodium-Cooled Fast Reactor (SFR), suggested in the reference and, Objective Provision Trees (OPTs) of reactivity control function for level 1, 2, 3 and 4 DiD were developed and suggested in this paper. The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of Prototype Gen-IV Sf (PGSFR) is not mature yet, the OPT is developed for KALIMER design. Developed level 1 to 4 OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defense-in-depth evaluation frame for the regulatory reviews for the licensing process. In the next stage of this study, other safety function will be researched and findings can be suggested as recommendations for the safety improvement

Project planning of Gen-IV sodium cooled fast reactor technology

Energy Technology Data Exchange (ETDEWEB)

Yoo, Jaewoon; Joo, H. K.; Cho, C. H.; Kim, Y. G.; Lee, D. U.; Jin, M. W.

2013-05-15

The project program will be established to shorten the design schedule by sharing the design man power and experimental facility, and by introducing the proven technology through international collaboration and the project plan including preliminary specific design, technology validation and fuel design validation plan will be more detail by reviewing the plan at the International Technical Review Meeting (ITRM). Periodic project progress review meeting will be held to find the technical issues and to resolve them. The results of the progress review meeting will be reflected into the final assessment of research project. The project progress review meeting will be held every quarter and external expert will also participate in the meeting. In parallel with the PGSFR development, innovative small modular SFR will be developed aiming to the international nuclear market. The system and component technologies of both system can be shared but innovative concept will be implemented into the design. Ultra long life core design concept and supercritical CO{sub 2} Brayton cycle will be considered as the innovative concept for enhancing the plant economy and safety.

Project planning of Gen-IV sodium cooled fast reactor technology

International Nuclear Information System (INIS)

Yoo, Jaewoon; Joo, H. K.; Cho, C. H.; Kim, Y. G.; Lee, D. U.; Jin, M. W.

2013-05-01

The project program will be established to shorten the design schedule by sharing the design man power and experimental facility, and by introducing the proven technology through international collaboration and the project plan including preliminary specific design, technology validation and fuel design validation plan will be more detail by reviewing the plan at the International Technical Review Meeting (ITRM). Periodic project progress review meeting will be held to find the technical issues and to resolve them. The results of the progress review meeting will be reflected into the final assessment of research project. The project progress review meeting will be held every quarter and external expert will also participate in the meeting. In parallel with the PGSFR development, innovative small modular SFR will be developed aiming to the international nuclear market. The system and component technologies of both system can be shared but innovative concept will be implemented into the design. Ultra long life core design concept and supercritical CO 2 Brayton cycle will be considered as the innovative concept for enhancing the plant economy and safety

Seminar on Heat-transfer fluids for fast neutron reactors

International Nuclear Information System (INIS)

Brechet, Yves; Dautray, Robert; Friedel, Jacques; Brezin, Edouard; Martin, Georges; Pineau, Andre; Carre, Francois; Gauche, Francois; Rodriguez, Guillaume; Latge, Christian; Cabet, Celine; Garnier, Jean-Claude; Bamberger, Yves; Sauvage, Jean-Francois; Buisine, Denis; Agostini, Pietro; Ulyanov, Vladimir; Auger, Thierry; Heuer, Daniel; Ghetta, Veronique; Bubelis, Evaldas; Charlaix, Elisabeth; Barrat, Jean-Louis; Boquet, Lyderic; Glickman, Evgueny; Escaravage, Claude

2014-03-01

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

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

ASTRID: Advanced Sodium Technological Reactor for Industrial Demonstration

International Nuclear Information System (INIS)

Vasile, A.

2012-01-01

Conclusions: • R&D results [CEA-AREVA-EDF] obtained from 2007 to 2009 have contributed to ASTRID mid 2010 choice of options; • ASTRID has the objective to demonstrate at the industrial scale progress in the identified domains of SFR weakness (safety, operability, economy). and to perform transmutation demonstrations; • A lot of improvements are related to safety; • The first very important milestone is 2012 (June 2006 French Act on wastes management): – ASTRID pre-conceptual design studies: 2010-2012; – First investment cost evaluation; – First safety Authorities advice on the orientations for ASTRID safety; • With the ASTRID program funded by the French government, France has the opportunity to develop a GEN IV Sodium Fast Reactor

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)

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)

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)

Development of observation techniques in reactor vessel of experimental fast reactor Joyo

International Nuclear Information System (INIS)

Takamatsu, Misao; Imaizumi, Kazuyuki; Nagai, Akinori; Sekine, Takashi; Maeda, Yukimoto

2010-01-01

In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. And several IVO equipments for an SFR are developed. However, in order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. During the investigation of an incident that occurred with Joyo, IVO using a standard Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS). A simple 6 m overhead view of each S/A, through the fuel handling or inspection holes etc, was photographed using a VC for making observations of the top of S/As and IVS. About 650 photographs were required to create a composite photograph of the top of the entire S/As and IVS, and a resolution was estimated to be approximately 1 mm. In order to observe the bottom face of the UCS, a Remote Handling Device (RHD) equipped with RRFs (approximately 13 m long) was specifically developed for Joyo with a tip that could be inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. A total of about 35,000 photographs were needed for the full investigation. Regarding the resolution, the sodium flow regulating grid of 0.8 mm in thickness could be discriminated. The performance of IVO equipments under the actual reactor environment was successfully confirmed. And the results provided useful information on incident investigations. In addition, fundamental findings and the experience gained during this study, which included the design of equipment, operating procedures, resolution, lighting adjustments, photograph composition and the durability of the RRF under radiation exposure, provided valuable insights into further improvements and verifications for IVO techniques to

Safety approach and research and development presentation for the selected systems of the International forum Generation IV

International Nuclear Information System (INIS)

Fiorini, G.L.

2003-01-01

This paper deals with the six projects of the Generation IV forum: Sodium Fast reactor, lead fast reactor, gas fast reactor, very high temperature reactor, supercritical water reactor, molten salt reactor. The technical objectives of the reactor safety and the design/evaluation approach are discussed. (A.L.B.)

Creep buckling problems in fast reactor components

International Nuclear Information System (INIS)

Ramesh, R.; Damodaran, S.P.; Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

1995-01-01

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

The SGR Multipurpose - Generation IV - Transportable Cogeneration Nuclear Reactor with Innovative Shielding

International Nuclear Information System (INIS)

Pahladsingh, R.R.

2002-01-01

Deregulation and liberalization are changing the global energy-markets. At the same time innovative technologies are introduced in the electricity industry; often as a requirement from the upcoming Digital Society. Energy solutions for the future are more seen as a mix of energy-sources for generation-, transmission- and distribution energy-services. The Internet Energy-web based 'Virtual' enterprises are coming up and will gradually change our society. It the fast changing world we have to realize that there will be less time to look for the adequate solutions to anticipate on global developments and the way they will influence our own societies. Global population may reach 9 billion people by 2030; this will put tremendous pressure on energy-, water- and food supply in the global economy. It is time to think about some major issues as described below and come up with the right answers. These are needed on very short term to secure a humane global economic growth and the sustainable global environment. The DOE (Department of Energy - USA) has started the Generation IV initiative for the new generation of nuclear reactors that must lead to much better safety, economics and public acceptance the new reactors. The SGR (Simplified Gas-cooled Reactor) is being proposed as a Generation IV modular nuclear reactor, using graphite pebbles as fuel, whereby an attempt has been made to meet all the DOE requirements, to be used for future nuclear reactors. The focus in this paper is on the changing and emerging global energy-markets and shows some relevant criteria to the nuclear industry and how we can anticipate with improved and new designs towards the coming Digital Society. (author)

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

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

International Nuclear Information System (INIS)

Obara, Toru; Sekimoto, Hiroshi

1997-01-01

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

Review of fast reactor activities in India

Energy Technology Data Exchange (ETDEWEB)

Paranjpe, S R [Reactor Research Centre, Kalpakkam, Tamil Nadu (India)

1981-05-01

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

Review of fast reactor activities in India

International Nuclear Information System (INIS)

Paranjpe, S.R.

1981-01-01

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

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

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

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.

Gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, Hiroyuki

1982-07-01

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

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

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.

Prototype fast breeder reactor main options

International Nuclear Information System (INIS)

Bhoje, S.B.; Chellapandi, P.

1996-01-01

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

Technical and management challenges associated with structural materials degradation in nuclear reactors in the future

International Nuclear Information System (INIS)

Ford, F.P.

2007-01-01

issue is compounded by the fact that some of the future GEN IV reactor designs involve fast neutron spectra, and all involve increases in temperature to the range 500 o C - 1250 o C. Comparatively little is known of the effect of, for instance, creep-fatigue interactions in high irradiation fluxes on the structural integrity of the potential materials of construction. In spite of these technical concerns there is the business management expectation that all of these reactors will experience very few materials degradation problems that might affect the economics of operation. The paper starts with a review of our present capability to predict the materials degradation modes encountered in the current BWR and PWR reactor designs. This capability is the basis for any analysis of the future degradation problems (and their mitigation) in the current reactors and in the evolutionary water-cooled reactor designs. This section concludes with an overview of assessments of future materials degradation issues that might be expected in these water-cooled reactors. These preliminary discussions are then broadened to cover some of the more obvious technical problems likely to be encountered with the more advanced GEN IV designs, such as the Very High Temperature Reactor (VHTR) and the Super Critical Water Cooled Reactor (SCWR). The article concludes with a brief discussion of some of the challenges facing the technical management/leadership, with some suggestions on how to overcome them. These challenges may become especially severe given the fact that the technical problems must be overcome in a time frame that is short compared with that taken to resolve the issues that have faced us over the last 30 years. Some specific management challenges include: The decrease in the number of experienced experimentalists and analysts over the last 10 years; The decrease in 'institutional' memory as it relates to the operation of the current reactors, and the design and construction of

Fast breeder reactor safety : a perspective

International Nuclear Information System (INIS)

Kale, R.D.

1992-01-01

Taking into consideration India's limited reserves of natural and vast reserves of thorium, the fast reactor route holds a great promise for India's energy supply in future. The fast reactor fueled with 239 Pu/ 238 U (unused or depleted) produces (breeds) more fissionable fuel material 239 Pu than it consumes. Calculations show that a fast breeder reactor (FBR) increases energy potential of natural uranium by about 60 times. As the fast reactor can also convert 232 Th into 233 U which is a fissionable material, it can make India's thorium reserves a source of almost inexhaustible energy supply for a long time to come. Significant advantage of FBR plants cooled by sodium and their world-wide operating experience are reviewed. There are two main safety issues of FBR, one nuclear and the other non-nuclear. The nuclear issue concerns core disruptive accident and the non-nuclear one concerns the high chemical energy potential of sodium. These two issues are analysed and it is pointed that they are manageable by current design, construction and operational practices. Main findings of safety research during the last six to eight years in West European Countries and United States of America (US) are summarised. Three stage engineered safety provision incorporated into the design of the sodium cooled Fast Breeder Test Reactor (FBTR) commissioned at Kalpakkam are explained. The important design safety features of FBTR such as primary system containment, emergency core cooling, plant protection system, inherent safety features achieved through reactivity coefficients, and natural convection cooling are discussed. Theoretical analysis and experimental research in fast reactor safety carried out at the Indira Gandhi Centre for Atomic Research during the past some years are reviewed. (M.G.B.)

Improvement of the decay heat removal characteristics of the generation IV gas-cooled fast reactor

International Nuclear Information System (INIS)

Epiney, A. S.

2010-09-01

The majority of NPPs worldwide are currently light water reactors, using ordinary water as both coolant and moderator. (...) For the longer-term future, viz. beyond the year 2030, Research and Development is currently ongoing on Generation IV NPPs, aimed at achieving closure of the nuclear fuel cycle, and hence both drastically improved utilization of fuel resources and minimization of long-lived radioactive wastes. Since the very beginning of the international cooperation on Generation IV, viz. the year 2000, the main research interest in Europe as regards the advanced fast-spectrum systems needed for achieving complete fuel cycle closure, has been for the Sodium-cooled Fast Reactor (SFR). However, the Gas-cooled Fast Reactor (GFR) is currently considered as the main back-up solution. Like the SFR, the GFR is an efficient breeder, also able to work as iso-breeder using simply natural uranium as feed and producing waste which is predominantly in the form of fission products. The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For depressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure conditions, need to be

General Report on the Technical Sessions

International Nuclear Information System (INIS)

Carré, Frank

2013-01-01

General Remarks: • Research & Projects on Fast Neutron Reactors & related Fuel Cycles remain at sustained level worldwide; • Active participation in Technical Oral & Poster Sessions & Sustained exchanges; • Emphasis on Safety in the aftermath of Fukushima accident: → Gen-IV initiative on “Common design/safety criteria” in relation with the IAEA; • Diversity & Complementarity of National Projects of near term large power Fast Reactors & Technology Demonstrators of Next Generation Fast Reactors: – Ambitious SFR deployment scenarios of Russia, India, China…; – Near term Demonstrators of LFR technology in Russia; – Active research, promising innovations and plans for demonstrations in all major nuclear countries on SFRs but also LFR, GFR, MSFR… • Continuing improvements & Search for breakthroughs: two approaches with their own rationale & timeline that may complement each other in a global international roadmap. Key role of operating FRs for feedback & testing; • Increasing importance of numerical simulation and basic research; • Attractiveness of Gen-IV systems for Nuclear Education & Training

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.

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.

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

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)

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.)

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

Sensitivity analysis of the kinetic behaviour of a Gas Cooled Fast Reactor to variations of the delayed neutron parameters

International Nuclear Information System (INIS)

Van Rooijen, W. F. G.; Lathouwers, D.

2007-01-01

In advanced Generation IV (fast) reactors an integral fuel cycle is envisaged, where all Heavy Metal is recycled in the reactor. This leads to a nuclear fuel with a considerable content of Minor Actinides. For many of these isotopes the nuclear data is not very well known. In this paper the sensitivity of the kinetic behaviour of the reactor to the dynamic parameters λ k , β k and the delayed spectrum χ d,k is studied using first order perturbation theory. In the current study, feedback due to Doppler and/or thermohydraulic effects are not treated. The theoretical framework is applied to a Generation IV Gas Cooled Fast Reactor. The results indicate that the first-order approach is satisfactory for small variations of the data. Sensitivities to delayed neutron data are similar for increasing and decreasing transients. Sensitivities generally increase with reactivity for increasing transients. For decreasing transients, there are less clearly defined trends, although the sensitivity to the delayed neutron spectrum decreases with larger sub-criticality, as expected. For this research, an adjoint capable version of the time-dependent diffusion code DALTON is under development. (authors)

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)

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.

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

Objective Provision Tree (OPT) in sodium cooled fast reactors; Objective Provision Tree (OPT) en reactores rapidos refrigerados por sodio. Aplicacion a la funcion de seguridad de evacuacion de calor residual

Energy Technology Data Exchange (ETDEWEB)

Queral, C.; Montero-Mayorga, J.; Gonzalez-Cadelo, J.

2013-07-01

Application to the safety function of residual heat removal As part of the project {sup S}afety Assessment for Reactor of GEN-IV (SARGEN IV) has been implemented the methodology ISAM from the IAEA to the safety assessment of new sodium reactor designs. Within the ISAM, a new tool to facilitate this assessment is the Objective Provision Tree (OPT) which documents the provisions necessary for each of the levels of defense in depth, as well as for each critical function of security. Due to the design innovations that have sodium reactors, the evaluation of safety and licensing of these reactors requires special considerations. In this work we have analyzed the mechanisms of failure of the safety function concerning the evacuation of waste heat, and have been proposed different provisions for each of the first three levels of defense in depth. The main result of this work is reflected in the elaboration of the OPTs, one for each of the first three levels of defense in depth for the safety of evacuation of residual heat function. These trees represent in a schematic way the provisions necessary to comply with the objectives of each level which are respectively: 1) deviations from normal operation, 2) control of abnormal operation and fault detection and 3) incidental control.

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

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.

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

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

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

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)

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)

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

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

Synergies in the design and development of fusion and generation IV fission reactors

International Nuclear Information System (INIS)

Bogusch, E.; Carre, F.; Knebel, J.; Aoto, K.

2007-01-01

Future fusion reactors or systems and Generation IV fission reactors are designed and developed in worldwide programmes mostly involving the same partners to investigate and assess their potential for realisation and contribution to meet the future energy needs beyond 2030. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core face similar design issues and development needs. Therefore the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactors or systems will be designed for helium and liquid metal cooling and higher temperatures similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches might create synergistic design and development programmes. Therefore an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in support of common technologies. (orig.)

Synergies in the design and development of fusion and generation IV fission reactors

International Nuclear Information System (INIS)

Bogusch, E.; Carre, F.; Knebel, J.U.; Aoto, K.

2008-01-01

Future fusion reactor and Generation IV fission reactor systems are designed and developed in worldwide programmes to investigate and assess their potential for realisation and contribution to the future energy needs beyond 2030 mostly involving the same partners. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except for the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core, face similar design issues and development needs. Therefore, the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactor systems will be designed for high-temperature helium and liquid metal cooling but also water including supercritical water and molten salt similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches can create synergistic design and development programmes. Therefore, an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in

A Stochastic Proof of the Resonant Scattering Kernel and its Applications for Gen IV Reactors Type

International Nuclear Information System (INIS)

Becker, B.; Dagan, R.; Broeders, C.H.M.; Lohnert, G.

2008-01-01

Monte Carlo codes such as MCNP are widely accepted as almost-reference for reactor analysis. The Monte Carlo Code should therefore use as few as possible approximations in order to produce 'experimental-level' calculations. In this study we deal with one of the most problematic approximations done in MCNP in which the resonances are ignored for the secondary neutron energy distribution, namely the change of the energy and angular direction of the neutron after interaction with a heavy isotope with pronounced resonances. The endeavour of exploiting the influence of the resonances on the scattering kernel goes back to 1944 where E. Wigner and J. Wilkins developed the first temperature dependent scattering kernel. However only in 1998, the full analytical solution for the double differential resonant dependent scattering kernel was suggested by W. Rothenstein and R. Dagan. An independent stochastic approach is presented for the first time to confirm the above analytical kernel with a complete different methodology. Moreover, by manipulating in a subtle manner the scattering subroutine COLIDN of MCNP, it is proven that this very subroutine is, to some extent, inappropriate as well as the relevant explanation in the MCNP manual. The impact of this improved resonance dependent scattering kernel on diverse types of reactors, in particular for the Generation IV innovative core design HTR, is shown to be significant. (authors)

Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyeong-Yeon, E-mail: hylee@kaeri.re.kr

2016-11-15

Highlights: • Comparison of elevated temperature design (ETD) codes was made. • Material properties and evaluation procedures were compared. • Two heat-resistant materials of Grade 91 steel and austenitic stainless steel 316 are the target materials in the present study. • Application of the ETD codes to Generation IV reactor components and a comparison of the conservatism was conducted. - Abstract: The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted.

Application of FORSS sensitivity and uncertainty methodology to fast reactor benchmark analysis

International Nuclear Information System (INIS)

Weisbin, C.R.; Marable, J.H.; Lucius, J.L.; Oblow, E.M.; Mynatt, F.R.; Peelle, R.W.; Perey, F.G.

1976-12-01

FORSS is a code system used to study relationships between nuclear reaction cross sections, integral experiments, reactor performance parameter predictions, and associated uncertainties. This paper presents the theory and code description as well as the first results of applying FORSS to fast reactor benchmarks. Specifically, for various assemblies and reactor performance parameters, the nuclear data sensitivities were computed by nuclide, reaction type, and energy. Comprehensive libraries of energy-dependent coefficients have been developed in a computer retrievable format and released for distribution by RSIC and NNCSC. Uncertainties induced by nuclear data were quantified using preliminary, energy-dependent relative covariance matrices evaluated with ENDF/B-IV expectation values and processed for 238 U(n,f), 238 U(n,γ), 239 Pu(n,f), and 239 Pu(ν). Nuclear data accuracy requirements to meet specified performance criteria at minimum experimental cost were determined

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

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.)

Strategies of development of reactor types

International Nuclear Information System (INIS)

Bacher, P.

2004-01-01

The development of nuclear energy in the coming decades will depend on the goals followed, on the available technologies and on the strategies implemented in the world in agreement with public acceptation. This article is limited to the technical aspects of the strategies of development of reactor types: 1 - objectives; 2 - common constraints to all reactor types: safety and terrorism risks, wastes, non-proliferation, economics; 3 - different reactor types: general considerations, proven technologies (PWR, BWR, Candu), non-proven technologies but having an important experience, technologies at the design stage; 4 - energy systems and 'Generation IV forum': systems based on thermal neutron reactors and low enrichment, systems for the valorization of 238 U, systems for Pu burning, systems allowing the destruction of minor actinides, thorium-based systems, the Gen IV international forum; 5 - conclusion. (J.S.)

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

International Nuclear Information System (INIS)

Koch, M.; Kazimi, M.S.

1991-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Koch, M.; Kazimi, M.S.

1991-04-01

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

The story of fission reactors: from Chicago Pile to advanced energy systems

International Nuclear Information System (INIS)

Kannan, Umasankari

2017-01-01

Nuclear reactors have been designed which cater to different applications from small research reactors of a few watts to power reactors of several Giga Watts. Based on the neutron energy, there are thermal, intermediate and fast reactors operating are being designed. On the fuel utilization front, there are designs ranging from reactors using natural uranium fuel to enriched uranium to more efficient thorium based reactors. Reactors have also been designed which are neutron eaters, minor actinide burners and breeders. There have been variety of coolant and moderating materials used for different applications from water, gas cooled, liquid sodium cooled to molten salt cooled reactors. Several new reactor designs have been developed using innovative concepts in high temperature reactors, nuclear power packs and compact reactors for special purposes. The design challenges are many from modest designs to complicated hybrid reactors. The GEN-IV forum of IAEA has selected a few of these reactor designs for commercial power production in the coming years based on several quantified indicators. The evolutionary and revolutionary design approaches have been made over the years catering to different need of energy generation. A glimpse of some of the reactors being currently developed and the design modifications done in existing reactors have been given in this paper

Pool type liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Guthrie, B.M.

1978-08-01

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

Sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Hokkyo, N; Inoue, K; Maeda, H

1968-11-21

In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

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)

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.)

Millimeter-Wave Thermal Analysis Development and Application to GEN IV Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Wosko, Paul; Sundram, S. K.

2012-10-16

New millimeter-wave thermal analysis instrumentation has been developed and studied for characterization of materials required for diverse fuel and structural needs in high temperature reactor environments such as the Next Generation Nuclear Plant (NGNP). A two-receiver 137 GHz system with orthogonal polarizations for anisotropic resolution of material properties has been implemented at MIT. The system was tested with graphite and silicon carbide specimens at temperatures up to 1300 ºC inside an electric furnace. The analytic and hardware basis for active millimeter-wave radiometry of reactor materials at high temperature has been established. Real-time, non contact measurement sensitivity to anisotropic surface emissivity and submillimeter surface displacement was demonstrated. The 137 GHz emissivity of reactor grade graphite (NBG17) from SGL Group was found to be low, ~ 5 %, in the 500 – 1200 °C range and increases by a factor of 2 to 4 with small linear grooves simulating fracturing. The low graphite emissivity would make millimeter-wave active radiometry a sensitive diagnostic of graphite changes due to environmentally induced stress fracturing, swelling, or corrosion. The silicon carbide tested from Ortek, Inc. was found to have a much higher emissivity at 137 GHz of ~90% Thin coatings of silicon carbide on reactor grade graphite supplied by SGL Group were found to be mostly transparent to millimeter-waves, increasing the 137 GHz emissivity of the coated reactor grade graphite to about ~14% at 1250 ºC.

A review of fast reactor programme in Japan

International Nuclear Information System (INIS)

Matsuno, Y.; Bando, S.

1981-03-01

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

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

General remarks on fast neutron reactor physics

International Nuclear Information System (INIS)

Barre, J.Y.

1980-01-01

The main aspects of fast reactor physics, presented in these lecture notes, are restricted to LMFBR's. The emphasis is placed on the core neutronic balance and the burn-up problems. After a brief description of the power reactor main components and of the fast reactor chronology, the fundamental parameters of the one-group neutronic balance are briefly reviewed. Then the neutronic burn-up problems related to the Pu production and to the doubling time are considered

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)

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.)

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.

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)

Gen-III/III+ reactors. Solving the future energy supply shortfall. The SWR-1000 option

International Nuclear Information System (INIS)

Stosic, Z.V.

2006-01-01

Deficiency of non-renewable energy sources, growing demand for electricity and primary energy, increase in population, raised concentration of greenhouse gases in the atmosphere and global warming are the facts which make nuclear energy currently the most realistic option to replace fossil fuels and satisfy global demand. The nuclear power industry has been developing and improving reactor technology for almost five decades and is now ready for the next generation of reactors which should solve the future energy supply shortfall. The advanced Gen-III/III+ (Generation III and/or III+) reactor designs incorporate passive or inherent safety features which require no active controls or operational intervention to manage accidents in the event of system malfunction. The passive safety equipment functions according to basic laws of physics such as gravity and natural convection and is automatically initiated. By combining these passive systems with proven active safety systems, the advanced reactors can be considered to be amongst the safest equipment ever made. Since the beginning of the 90's AREVA NP has been intensively engaged in the design of two advanced Gen-III+ reactors: (i) PWR (Pressurized Water Reactor) EPR (Evolutionary Power Reactor) and (ii) BWR (Boiling Water Reactor) SWR-1000. The SWR-1000 reactor design marks a new era in the successful tradition of BWR technology. It meets the highest safety standards, including control of a core melt accident. This is achieved by supplementing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation. A short construction period, flexible fuel cycle lengths and a high fuel discharge burn-up contribute towards meeting economic goals. The SWR-1000 completely fulfils international nuclear regulatory requirements. (author)

Philosophy of safety evaluation on fast breeder reactor

International Nuclear Information System (INIS)

1981-01-01

This is the report submitted from the special subcommittee on reactor safety standard to the Nuclear Safety Commission on October 14, 1980, and it was decided to temporarily apply this concept to the safety examination on fast breeder reactors. The examination and discussion of this report were performed by taking the prototype reactor ''Monju'' into consideration, which is to be the present target, referring to the philosophy of the safety evaluation on fast breeder reactors in foreign countries and based on the experiences in the fast experimental reactor ''Joyo''. The items applicable to the safety evaluation for liquid metal-cooled fast breeder reactors (LMFBR) as they are among the existing safety examination guidelines are applied. In addition to the existing guidelines, the report describes the matters to be considered specifically for core, fuel, sodium, sodium void, reactor shut-down system, reactor coolant boundary, cover gas boundary and others, intermediate cooling system, removal of decay heat, containment vessels, high temperature structures, and aseismatic property in the safety design of LMFBR's. For the safety evaluation for LMFBR's, the abnormal transient changes in operation and the phenomena to be evaluated as accidents are enumerated. In order to judge the propriety of the criteria of locating LMFBR facilities, the serious and hypothetical accidents are decided to be evaluated in accordance with the guideline for reactor location investigation. (Wakatsuki, Y.)

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

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.

Status of liquid metal cooled fast breeder reactors

International Nuclear Information System (INIS)

1985-01-01

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

European lead fast reactor (ELSY and LEADER projects)

International Nuclear Information System (INIS)

Alemberti, Alessandro; Carlsson, Johan; Malambu, Edouard; Orden, Alfredo; Cinotti, Luciano; Struwe, Dankward; Agostini, Pietro; Monti, Stefano

2010-01-01

The conceptual design of the European Lead Fast Reactor is being developed starting from September 2006, in the frame of the ELSY project. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative DHR systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; it has been demonstrated that the high density of lead can be mitigated by compact solutions and adoption of seismic isolators. Safety has been one of the major focuses all over the ELSY development. In addition to the inherent safety advantages of lead coolant (high boiling point and no exothermic reactions with air or water) a high safety grade of the overall system has been reached. In fact the overall primary system has been

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

Economic assessment of R and D with real options in the field of fast reactors taking into account uncertainty on their competitiveness: the case of France

International Nuclear Information System (INIS)

Taverdet-Popiolek, Nathalie; Shoai Tehrani, Bianka

2013-01-01

In a context of potential worldwide nuclear development, this paper aims at assessing the economic value of pursuing research in Generation IV fast reactors today, given that it would allow industrial deployment around 2040 in case of high uranium prices. Two key variables shall be considered as inputs for the assessment: the price of uranium and the over-cost of Generation IV reactors compared to the previous generation. Our model based on real options theory demonstrates that this value is positive and outweighs the risks associated with the competitiveness of Generation IV. (authors)

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)

Calculation of the neutron parameters of fast thermal reactor

International Nuclear Information System (INIS)

Kukuleanu, V.; Mocioiu, D.; Drutse, E.; Konstantinesku, E.

1975-01-01

The system of neutron calculation for fast reactors is given. This system was used for estimation of physical parameters of fast thermal reactors investigated. The results obtained and different specific problems of the reactors of this type are described. (author)

The ARIES-II and ARIES-IV second-stability tokamak reactors

International Nuclear Information System (INIS)

Najmabadi, F.; Conn, R.W.; Hasan, M.Z.; Mau, T.-K.; Sharafat, S.; Baxi, C.B.; Leuer, J.A.; McQuillan, B.W.; Puhn, F.A.; Schultz, K.R.; Wong, C.P.C.; Brooks, J.; Ehst, D.A.; Hassanein, A.; Hua, T.; Hull, A.; Mattis, R.; Picologlou, B.; Sze, D.-K.; Dolan, T.J.; Herring, J.S.; Bathke, C.G.; Krakowski, R.A.; Werley, K.A.; Bromberg, L.; Schultz, J.; Davis, F.; Holmes, J.A.; Lousteau, D.C.; Strickler, D.J.; Jardin, S.C.; Kessel, C.; Snead, L.; Steiner, D.; Valenti, M.; El-Guebaly, L.A.; Emmert, G.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.; Sviatoslavsky, I.N.; Cheng, E.T.

1992-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. The ARIES-III study focuses on the potential of tokamaks to operate with D- 3 He fuel system as an alternative to deuterium and tritium. The ARIES-II and ARIES-IV designs have the same fusion plasma but different fusion-power-core designs. The ARIES-II reactor uses liquid lithium as the coolant and tritium breeder and vanadium alloy as the structural material in order to study the potential of low-activation metallic blankets. The ARIES-IV reactor uses helium as the coolant, a solid tritium-breeding material, and silicon carbide composite as the structural material in order to achieve the safety and environmental characteristic of fusion. In this paper the authors describe the trade-off leading to the optimum regime of operation for the ARIES-II and ARIES-IV second-stability reactors and review the engineering design of the fusion power cores

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)

Application of FORSS sensitivity and uncertainty methodology to fast reactor benchmark analysis

Energy Technology Data Exchange (ETDEWEB)

Weisbin, C.R.; Marable, J.H.; Lucius, J.L.; Oblow, E.M.; Mynatt, F.R.; Peelle, R.W.; Perey, F.G.

1976-12-01

FORSS is a code system used to study relationships between nuclear reaction cross sections, integral experiments, reactor performance parameter predictions, and associated uncertainties. This paper presents the theory and code description as well as the first results of applying FORSS to fast reactor benchmarks. Specifically, for various assemblies and reactor performance parameters, the nuclear data sensitivities were computed by nuclide, reaction type, and energy. Comprehensive libraries of energy-dependent coefficients have been developed in a computer retrievable format and released for distribution by RSIC and NNCSC. Uncertainties induced by nuclear data were quantified using preliminary, energy-dependent relative covariance matrices evaluated with ENDF/B-IV expectation values and processed for /sup 238/U(n,f), /sup 238/U(n,..gamma..), /sup 239/Pu(n,f), and /sup 239/Pu(..nu..). Nuclear data accuracy requirements to meet specified performance criteria at minimum experimental cost were determined.

Comparison of KANEXT and SERPENT for fuel depletion calculations of a sodium fast reactor

International Nuclear Information System (INIS)

Lopez-Solis, R.C.; Francois, J.L.; Becker, M.; Sanchez-Espinoza, V.H.

2014-01-01

As most of Generation-IV systems are in development, efficient and reliable computational tools are needed to obtain accurate results in reasonably computer time. In this study, KANEXT code system is presented and validated against the well-known Monte Carlo SERPENT code, for fuel depletion calculations of a sodium fast reactor (SFR). The KArlsruhe Neutronic EXtended Tool (KANEXT) is a modular code system for deterministic reactor calculations, consisting of one kernel and several modules. Results obtained with KANEXT for the SFR core are in good agreement with the ones of SERPENT, e.g. the neutron multiplication factor and the isotopes evolution with burn-up. (author)

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

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.)

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.

A review of the UK fast reactor programme, March 1979