WorldWideScience

Sample records for cooled fast reactors

  1. Gas cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-01

    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  2. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

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

  3. Fuel development for gas-cooled fast reactors

    Science.gov (United States)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    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 [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. 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 dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US 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.

  4. Capital cost: gas cooled fast reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-15

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

  6. Shape optimization of a sodium cooled fast reactor

    Science.gov (United States)

    Schmitt, Damien; Allaire, Grégoire; Pantz, Olivier; Pozin, Nicolas

    2014-06-01

    Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth.(1, 2) Usual optimization methods for core conception are based on a parametric description of a given core design(3).(4) New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints.(5, 6) First studies show that these methods could be applied to sodium cooled core conception.(7) In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a get realistic core layout. Its caracteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas.

  7. 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",

  8. 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",

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-20

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

  11. REVIEW OF REACTOR SAFETY ANALYSES OF FAST AND LIQUID METAL COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Shaver, R. E.; Wittenbrock, N. G.

    1967-11-01

    Safety analysis reports on United States fast and liquid metal cooled reactors were reviewed to gain a better understanding of the safety philosophy applied to the design of these facilities. This information was compiled to help guide the design and safety analysis of the Fast Flux Test Facility. No attempt was made to draw conclusions concerning the relative merit of different approaches and philosophies used by different reactor design teams. The facilities reviewed were; Enrico Fermi Atomic Power Plant (FERMI) Hallam Nuclear Power Facility (HALLAM) Southwest Experimental Fast Oxide Reactor (SEFOR) Fast Reactor Test Facility (FARET) Experimental Breeder Reactor No. 1 (EBR-I) Experimental Breeder Reactor No. 2 (EBR-II) Fast Reactor Zero Power Experiment (ZPR - III). The information gathered from the safety analysis reports is tabulated under these headings: Control and Safety Systems; Reactor Protection Systems; Backup Systems; Containment or Confinement Systems; Inherent Reactivity Effects and Important Physics Parameters; Fuel and Fuel Handling; Accidents Considered and Chemical Problems; Site; Exhaust Ventilation System; and Waste Effluents.

  12. Preliminary Assessment of a Debris Bed Cooling Performance for Demonstration Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chung Ho; Park, Chang Gyu; Song, Hoon; Kim, Young Gyun; Jeong, Hae Yong; Chang, Jin Wook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    In the case of the sodium-cooled fast reactor such as KALIMER-600, Hypothetical Core Disruptive Accident (HCDA) attributed from mass nuclear fuel melting is unlikely to occur due to defense in depth concepts to meet requirements of redundancy and diversity. Multiple faults such as loss of flow, loss of heat sink, or transient overpower without scram are to lead rising the power level until cladding failure as reactivity increasing. The fact that metallic fuel melts at a lower temperature than the cladding allows significant in-pin- fuel motion to occur prior to cladding failure. Also, the combination of Doppler and axial expansion feedback and negative feedback associated with the in-pin fuel relocation prevents the reactivity from reaching prompt critical. Finally, the resulting reactivity and power reductions help prevent fuel temperatures from rising more than the fuel melting temperature. It is more difficult to occur HCDA in a metallic fueled core because reactor power and heat removal capability is maintained in balance by inherent safety characteristics However, for the future design of sodium-cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth considering due to the triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. Accordingly, evaluation of a packed debris bed cooling performance with single phase flow for demonstration sodium-cooled fast reactor was carried out for proof of the in-vessel retention of the core debris

  13. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  14. Comparison of fuel assemblies in lead cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G., E-mail: alejandria.peval@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2016-09-15

    This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

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

    OpenAIRE

    Pengcheng Zhao; Kangli Shi; Shuzhou Li; Jingchao Feng; Hongli Chen

    2016-01-01

    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 kineti...

  16. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all H

  17. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all

  18. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all H

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

    OpenAIRE

    Sungjoo Lee; Byungun Yoon; Juneseuk Shin

    2016-01-01

    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 indic...

  20. Simulation of Reactor Transient and Design Criteria of Sodium-cooled Fast Reactors

    OpenAIRE

    Gottfridsson, Filip

    2010-01-01

    The need for energy is growing in the world and the market of nuclear power is now once more expanding. Some issues of the current light-water reactors can be solved by the next generation of nuclear power, Generation IV, where sodium-cooled reactors are one of the candidates. Phénix was a French prototype sodium-cooled reactor, which is seen as a success. Although it did encounter an earlier unexperienced phenomenon, A.U.R.N., in which a negative reactivity transient followed by an oscillati...

  1. 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.

  2. 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.

  3. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

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

  5. Device for cooling the main vessel of a fast fission nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Debru, M.

    1984-10-16

    The annular space delimited by the main vessel and an internal shell is in communication with the zone of the reactor vessel, in which the cold primary liquid is located. The annular space delimited by the shell and by an internal shell is in communication with the lower part of the core via tubes. Thus, the cold primary liquid is injected into the space where it circulates from bottom to top, and flows into the space, where it circulates from top to bottom while at the same time cooling the main vessel. The invention applies, in particular, to fast fission nuclear reactors cooled by liquid sodium.

  6. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

  7. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    OpenAIRE

    Lindström, Tobias

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    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{sub 2} and B{sub 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.

  12. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  13. Preliminary Study of Lead-Oxide Cooled Fast Reactor with Natural Uranium as an Input Fuel with Reactor Shuffling Strategy

    Science.gov (United States)

    Mahmudah, Rida SN; Su’ud, Zaki

    2017-01-01

    A preliminary study of lead-oxide cooled fast reactor with natural uranium as an input fuel using reactor shuffling strategy has been conducted. In this study, reactor core is divided into four zone with the same volume, each zone use different uranium enrichment. The enrichment number is estimated so that in the end of reactor’s operation, we only need to add natural uranium as the fresh input fuel. This study used UN-PuN as the fuel and lead oxide as the coolant. Several parameter studies have been conducted to determine the most suitable input condition. It is confirmed in this study that with fuel : cladding : coolant ratio of 53 : 10 : 37, and uranium enrichment in the first to the fourth zone of 0%, 6.25%, 7.5% and 8%, respectively, the reactor can operate as long as 20 years of operation with terminal k-eff of 1.0004.

  14. Experimental investigation of a new method for advanced fast reactor shutdown cooling

    Science.gov (United States)

    Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

    2017-07-01

    We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

  15. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    Directory of Open Access Journals (Sweden)

    Lap-Yan Cheng

    2009-01-01

    Full Text Available The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR in a GEN IV direct-cycle gas-cooled fast reactor (GFR which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

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

    Science.gov (United States)

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

    2017-01-01

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

  17. FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

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

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  19. Minor actinides impact on basic safety parameters of medium-sized sodium-cooled fast reactor

    Directory of Open Access Journals (Sweden)

    Darnowski Piotr

    2015-03-01

    Full Text Available An analysis of the influence of addition of minor actinides (MA to the fast reactor fuel on the most important safety characteristics was performed. A special emphasis was given to the total control rods worth in order to describe qualitatively and quantitatively its change with MA content. All computations were performed with a homogeneous assembly model of modified BN-600 sodium-cooled fast reactor core with 0, 3 and 6% of MA. A model was prepared for the Monte Carlo neutron transport code MCNP5 for fresh fuel in the beginning-of-life (BOL state. Additionally, some other parameters, such as Doppler constant, sodium void reactivity, delayed neutron fraction, neutron fluxes and neutron spectra distribution, were computed and their change with MA content was investigated. Study indicates that the total control rods worth (CRW decreases with increasing MA inventory in the fuel and confirms that the addition of MA has a negative effect on the delayed neutron fraction.

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

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

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

  1. Development of level-1 PSA method applicable to Japan Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kurisaka, K., E-mail: kurisaka.kennichi@jaea.go.jp [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, Ibaraki (Japan); Sakai, T.; Yamano, H. [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, Ibaraki (Japan); Fujita, S.; Minagawa, K. [Department of Mechanical Engineering, School of Engineering, Tokyo Denki University, Tokyo (Japan); Yamaguchi, A.; Takata, T. [Department of Energy and Environment Engineering, Osaka University, Osaka (Japan)

    2014-04-01

    This paper describes a study to develop the level-1 probabilistic safety assessment (PSA) method that is applicable to the Japan Sodium-cooled Fast Reactor (JSFR). This study has been started since August 2010 and aims to provide a new evaluation method of (1) passive safety architectures related to internal events and (2) an advanced seismic isolation system related to a seismic event as a representative external event in Japan. Regarding the internal events evaluation, a quantitative analysis on the frequency of the core damage caused by reactor shutdown failure was conducted. A failure in passive reactor shutdown was taken into account in the event tree model. The failure rate of sodium-cooled fast reactor (SFR) specific components was evaluated based on the operating experience in existing SFRs by applying the Hierarchical Bayesian Method, which can consider a plant-to-plant variability. By conducting an uncertainty analysis, it was found that the assumption about the correlation of the probability parameters between the main and backup reactor shutdown systems (RSSs) is sensitive to the mean value of the frequency of the core damage caused by reactor shutdown failure. As for the seismic event evaluation, seismic response analysis and sensitivity analysis of a seismic isolation system were carried out. Rubber bearings have a hardening property in horizontal direction and a softening property in vertical direction in case of large deformation. Therefore the analyses considered nonlinearity of rubber bearings. Both horizontal and vertical nonlinear characteristics of rubber bearings were explained by multi-linear model. Mass point analytical models were applied. At first, seismic response analysis was executed in order to investigate influence of nonlinearity of rubber bearing upon response of building. Then sensitivity analysis was executed. Parameters of rubber bearings, oil dampers and the building were fluctuated, and influence of dispersion of these

  2. Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

    Energy Technology Data Exchange (ETDEWEB)

    Stanisz, Przemysław, E-mail: pstanisz@agh.edu.pl; Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2016-05-15

    Graphical abstract: - Highlights: • We present the Monte Carlo modeling of the LFR in the adiabatic equilibrium state. • We assess the adiabatic equilibrium fuel composition using the MCB code. • We define the self-adjusting process of breeding gain by the control rod operation. • The designed LFR can work in the adiabatic cycle with zero fuel breeding. - Abstract: Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste. We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis

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

    OpenAIRE

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

    2016-01-01

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

  4. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Moe, Wayne [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); HolbrookINL, Mark [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

  5. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  6. Neutronic assessment of liquid-metal cooled fast reactors using thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pilarski, Stevan [Electricite de France R et D, 1 Avenue du General de Gaulle, 92141 Clamart (France); Institut de Physique Nucleaire d' Orsay, 15 rue Georges Clemenceau 91406 Orsay (France)

    2009-06-15

    The long-term sustainability of atomic fission energy will require the development of new types of reactors, able to exceed the limits of the existing ones in terms of optimal use of natural resources, which clearly necessitates breeding of fissile material. In this context, fast reactors using uranium-plutonium fuel are the most mature solution from an industrial viewpoint. In addition to the obvious interest in terms of fuel resources, there is a major incentive to consider the use of the {sup 232}Th- {sup 233}U fuel cycle as an alternative to the traditional {sup 238}U-{sup 239}Pu cycle for fast reactors: it is an effective way of addressing the safety issue of the highly positive void reactivity effect, which is a well-known problem for liquid-metal cooled fast reactors of commercial size [1]. This work investigates the performance of liquid-metal cooled fast reactors in {sup 232}Th-{sup 233}U fuel cycle and draws a comparison with the traditional {sup 238}U-{sup 239}Pu cycle. Four coolants have been considered: Na, Pb, Mg(17%at.)-Pb and Li(17%at.)-Pb; a simulation of their use in cores ranging from 700 MWth to 3600 MWth has been performed in two-dimensional diffusion theory using the European system of codes ERANOS [2,3] developed at CEA. The performance parameters such as the breeding ratio have been computed for each concept, alongside safety-related parameters: the delayed neutron fraction, the cycle reactivity swing, the Doppler constant and other thermal feedbacks. More specifically, the issue of void reactivity is studied in detail using perturbation theory. These calculations are performed at equilibrium fuel composition and are complemented by the study of the initial fuel loading at start-up which is a mixture of {sup 232}Th-{sup 239}Pu. The isotopic composition of the fissile corresponds to the plutonium available from French reactors in 2035. The conclusions of this work are that near-zero to large negative void reactivity effects can be achieved in

  7. A Compact Gas-Cooled Fast Reactor with an Ultra-Long Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Hangbok Choi

    2013-01-01

    Full Text Available In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2, which is a compact gas-cooled fast reactor (GFR. The EM2 augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2 core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization.

  8. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Oktamuliani, Sri, E-mail: srioktamuliani@ymail.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id [Nuclear and Reactor Physics Laboratory, FMIPA, ITB, Physics Buildings, Jl. Ganesha 10, Bandung 40132 (Indonesia)

    2015-09-30

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.

  9. 2400MWt GAS-COOLED FAST REACTOR DHR STUDIES STATUS UPDATE.

    Energy Technology Data Exchange (ETDEWEB)

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01

    A topical report on demonstrating the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor was published in March 2006. The analysis was performed with the system code RELAP5-3D (version 2.4.1.1a) and the model included the full complement of the power conversion unit (PCU): heat exchange components (recuperator, precooler, intercooler) and rotating machines (turbine, compressor). A re-analysis of the success case in Ref is presented in this report. The case was redone to correct unexpected changes in core heat structure temperatures when the PCU model was first integrated with the reactor model as documented in Ref [1]. Additional information on the modeling of the power conversion unit and the layout of the heat exchange components is provided in Appendix A.

  10. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  11. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    Science.gov (United States)

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-01

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  12. Development of objective provision trees for Sodium-Cooled Fast Reactor Defense-in-depth evaluation

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

    KALIMER is one of sodium-cooled fast reactor and being developed by Korea Atomic Energy Research Institute (KAERI), was developed and suggested in this paper. Developed OPT is for the defense-in-depth level 3, core heat removal safety function. Using OPT method, the evaluation of defense-in-depth implementation for the design features of KALIMER reactors were tried in this study. To utilize the design information of KALIMER, challenges in OPTs which are under development in this study, were identified based on the system physical boundaries. This approach make the identification of possible and postulated challenges much clear and this will be a benefit to further identification of provisions in KALIMER design. OPTs for other levels of defense-in-depth and other safety functions are under development.

  13. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  14. Gas-Cooled Fast Reactor: A Historical Overview and Future Outlook

    Directory of Open Access Journals (Sweden)

    W. F. G. van Rooijen

    2009-01-01

    Full Text Available A review is given of developments in the area of Gas-Cooled Fast Reactors (GCFR in the period from roughly 1960 until 1980. During that period, the GCFR concept was expected to increase the breeding gain, the thermal efficiency of a nuclear power plant, and alleviate some of the problems associated with liquid metal coolants. During this period, the GCFR concept was found to be more challenging than liquid-metal-cooled reactors, and none were ever constructed. In the second part of the paper, we provide an overview of the investigations on GCFR since the year 2000, when the Generation IV Initiative rekindled interest in this reactor type. The new GCFR concepts focus primarily on sustainable nuclear power, with very efficient resource use, minimum waste, and a very strong focus on (passive safety. An overview is presented of the main design characteristics of these Gen IV GCFRs, and a literature list is provided to guide the interested reader towards more detailed publications.

  15. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  16. 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.

  17. 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.

  18. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a {sup 23}Na(n,g){sup 24}Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B{sub 4}C shielding inside the subassembly.

  19. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    Science.gov (United States)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

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

    Energy Technology Data Exchange (ETDEWEB)

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

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

    Science.gov (United States)

    Raj, Baldev; Rao, K. Bhanu Sankara

    2009-04-01

    The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

  2. Contributions to the neutronic analysis of a gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reyes-Ramirez, Ricardo, E-mail: ricarera@yahoo.com.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reinking-Cejudo, Arturo G., E-mail: reinking@servidor.unam.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico)

    2011-06-15

    Highlights: > Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. > Fuel lattice and core criticality calculations were done. > A higher Doppler coefficient than coolant density coefficient. > Zirconium carbide is a better reflector than silicon carbide. > Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained

  3. Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

    Directory of Open Access Journals (Sweden)

    Yong Kyo Lee

    2015-02-01

    Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  4. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

    2017-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

  5. Interim status report on lead-cooled fast reactor (LFR) research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G.; Li, N.; Hosemann, P.; Zhang, J.; Bolind, A.; LLNL; LANL; Univ. of Illinois

    2008-03-31

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup

  6. Impact of nuclear data on sodium-cooled fast reactor calculations

    Science.gov (United States)

    Aures, Alexander; Bostelmann, Friederike; Zwermann, Winfried; Velkov, Kiril

    2016-03-01

    Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

  7. Impact of nuclear data on sodium-cooled fast reactor calculations

    Directory of Open Access Journals (Sweden)

    Aures Alexander

    2016-01-01

    Full Text Available Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

  8. 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.

  9. Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

    Energy Technology Data Exchange (ETDEWEB)

    Tucek, Kamil [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)]. E-mail: kamil.tucek@jrc.nl; Carlsson, Johan [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands); Wider, Hartmut [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)

    2006-08-15

    A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies. In this paper, two fast reactor systems are discussed-the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries. First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600 MW{sub e}) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems. We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating

  10. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    Science.gov (United States)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  11. Preliminary Reactor Head Bolt Design of Prototype Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Han, Insu; Koo, Gyeonghoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    As structural requirements, the reactor head is designed to withstand all of the pressure, temperatures and forces which are likely to be imposed on it. The bolts that fasten the head to the vessel flange. Design of the reactor head bolts so as to withstand the loads applied should be designed. Currently, preliminary design of the PGSFR reactor bolts is progressed. So far, we have designed and evaluated example. The number and cross-sectional areas of bolts were determined using the procedure given in ASME BPVC Section III, Division 1, Appendix E. The purpose of this study is to conduct design the number and cross-sectional area of bolts attaching the PGSFR reactor head to the reactor vessel, using the ASME procedure. In this paper, preliminary bolt design for PGSFR was carried out according to the ASME procedure. Detailed calculations were carried out for bolt root diameter = 80 mm and number of bolts Nb = 45. It should be noted that the seating pressure recommended in the ASME code is only a suggested value, not mandatory appendix E. It does not guarantee a leak-tight joint. So these quantities are needed to carry out fatigue analysis of the bolts and to assure leak tightness of the joint during operation. For the future work, the fatigue and seismic analysis will be performed.

  12. Low-power lead-cooled fast reactor loaded with MOX-fuel

    Science.gov (United States)

    Sitdikov, E. R.; Terekhova, A. M.

    2017-01-01

    Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

  13. Challenges and Innovative Technologies On Fuel Handling Systems for Future Sodium-Cooled Fast Reactors

    OpenAIRE

    Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

    2011-01-01

    International audience; The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into acc...

  14. Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ponciroli, R.; Passerini, S.; Vilim, R. B.

    2016-04-17

    In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based on the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.

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

    Directory of Open Access Journals (Sweden)

    T. R. Allen

    2007-01-01

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

  16. Current design efforts for the gas-cooled fast reactor (GFR)

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3850 (United States)]. e-mail: Kevan.Weaver@inl.gov

    2005-07-01

    Current research and development on the Gas-Cooled Fast Reactor (GCFR) 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 (AFC I) 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 GCFR: a helium-cooled, direct Brayton cycle power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GCFR. These are EURATOM (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, EURATOM (including the United Kingdom), France, Japan, and Switzerland have active research activities with respect to the GCFR. The research includes GCFR design and safety, and fuels/in-core materials/fuel cycle projects. This paper outlines the current design status of the GCFR, and includes work done in the areas mentioned above. (Author)

  17. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su’ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  18. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  19. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-01-01

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

  20. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    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 radiotoxicity 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. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  1. Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

    Energy Technology Data Exchange (ETDEWEB)

    Wakai, Takashi, E-mail: wakai.takashi@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki 311 1393 (Japan); Machida, Hideo; Yoshida, Shinji [TEPCO Systems Corporation, 2-37-28 Eitai, Koto-ku, Tokyo 135 0034 (Japan); Xu, Yang [Mitsubishi FBR Systems, Inc., 2-34-17 Jingumae, Shibuya-ku, Tokyo 150 0001 (Japan); Tsukimori, Kazuyuki [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki 311 1393 (Japan)

    2014-04-01

    This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J{sub IC}, and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins.

  2. 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.

  3. Gas-cooled fast reactor fuel-cost assessment. Final report, October 1978-September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, M.L.

    1979-01-01

    This program, contracted to provide a Gas Cooled Fast Reactor (GCFR) fuel assembly fabrication cost assessment, comprised the following basic activities: establish agreement on the ground rules for cost assessment, prepare a fuel factory flow sheet, and prepare a cost assessment for fuel assembly fabrication. Two factory sizes, 250 and 25 MTHM/year, were considered for fuel assembly fabrication cost assessment. The work on this program involved utilizing GE LMFBR cost assessment and fuel factory studies experience to provide a cost assessment of GCFR fuel assembly fabrication. The recent impact of highly sensitive safety and safeguards environment policies on fuel factory containment, safety, quality assurance and safeguards costs are significantly higher than might have been expected just a few years ago. Fuel assembly fabrication costs are significant because they represent an estimated 30 to 60% of the total fuel cycle costs. In light of the relative high cost of fabrication, changes in the core and assembly design may be necessary in order to enhance the overall fuel cycle economics. Fabrication costs are based on similar operations and experience used in other fuel cycle studies. Because of extrapolation of present technology (e.g., remote fuel fabrication versus present contact fabrication) and regulatory requirements, conservative cost estimates were made.

  4. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Science.gov (United States)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  5. 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.

  6. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  7. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Science.gov (United States)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  8. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  9. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube.

  10. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R. [Inst. for Neutron Physic and Reactor Technology, Karlsruhe Inst. of Technology, Campus Nord (Germany)

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  11. Stability analysis of a natural circulation lead-cooled fast reactor

    Science.gov (United States)

    Lu, Qiyue

    This dissertation is aimed at nuclear-coupled thermal hydraulics stability analysis of a natural circulation lead cooled fast reactor design. The stability concerns arise from the fact that natural circulation operation makes the system susceptible to flow instabilities similar to those observed in boiling water reactors. In order to capture the regional effects, modal expansion method which incorporates higher azimuthal modes is used to model the neutronics part of the system. A reduced order model is used in this work for the thermal-hydraulics. Consistent with the number of heat exchangers (HXs), the reactor core is divided into four equal quadrants. Each quadrant has its corresponding external segments such as riser, plenum, pipes and HX forming an equivalent 1-D closed loop. The local pressure loss along the loop is represented by a lumped friction factor. The heat transfer process in the HX is represented by a model for the coolant temperature at the core inlet that depends on the coolant temperature at the core outlet and the coolant velocity. Additionally, time lag effects are incorporated into this HX model due to the finite coolant speed. A conventional model is used for the fuel pin heat conduction to couple the neutronics and thermal-hydraulics. The feedback mechanisms include Doppler, axial/radial thermal expansion and coolant density effects. These effects are represented by a linear variation of the macroscopic cross sections with the fuel temperature. The weighted residual method is used to convert the governing PDEs to ODEs. Retaining the first and second modes, leads to six ODEs for neutronics, and five ODEs for the thermal-hydraulics in each quadrant. Three models are developed. These are: 1) natural circulation model with a closed coolant flow path but without coupled neutronics, 2) forced circulation model with constant external pressure drop across the heated channels but without coupled neutronics, 3) coupled system including neutronics with

  12. Ferritic steels for sodium-cooled fast reactors: Design principles and challenges

    Science.gov (United States)

    Raj, Baldev; Vijayalakshmi, M.

    2010-09-01

    An overview of the current status of development of ferritic steels for emerging fast reactor technologies is presented in this paper. The creep-resistant 9-12Cr ferritic/martensitic steels are classically known for steam generator applications. The excellent void swelling resistance of ferritic steels enabled the identification of their potential for core component applications of fast reactors. Since then, an extensive knowledge base has been generated by identifying the empirical correlations between chemistry of the steels, heat treatment, structure, and properties, in addition to their in-reactor behavior. A few concerns have also been identified which pertain to high-temperature irradiation creep, embrittlement, Type IV cracking in creep-loaded weldments, and hard zone formation in dissimilar joints. The origin of these problems and the methodologies to overcome the limitations are highlighted. Finally, the suitability of the ferritic steels is re-evaluated in the emerging scenario of the fast reactor technology, with a target of achieving better breeding ratio and improved thermal efficiency.

  13. Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

    Science.gov (United States)

    Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

    2017-01-01

    In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

  14. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    Science.gov (United States)

    Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

  15. Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-29

    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

  16. Modeling of natural circulation for the inherent safety analysis of sodium cooled fast reactors

    Directory of Open Access Journals (Sweden)

    A.S. Bochkarev

    2016-12-01

    Full Text Available The paper discusses a set of developed integrated one-dimensional models of thermal-hydraulic processes that contribute to the removal of decay heat in a BN-type reactor. The assumptions and constraints involved in one-dimensional equations of unsteady natural convection in closed circuits have been analyzed. It has been shown that the calculated values of the primary circuit sodium temperature and flow rate in conditions with a loss of heat sink and with a forced circulation of the primary coolant are in a reasonable agreement with the results of a benchmark experiment in the PHENIX reactor. The model makes it possible to assess the effects general thermophysical and geometrical parameters and the selected technology have on the efficiency of passive heat removal by the natural coolant convection in the reactor tank and in the emergency heat removal system's intermediate circuit and by the heat transfer through the reactor vessel. The model is a part of an integrated algorithm used to assess the inherent safety level of advanced fast neutron reactors and is intended primarily to develop, at the early conceptual design stage, the recommendations and requirements with respect to the reactor equipment parameters leading to an increase in the reactor inherent safety. The model will be used to identify the set of quantitative thermal-hydraulic criteria that have an effect on the dynamics of emergency transients leading to a potential loss of integrity by the reactor safety barriers, and to formulate such limits for the defined criteria as would cause, if observed, the requirement for the safety barrier integrity to be met under any combination of the accident initiating events.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  18. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    Science.gov (United States)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  19. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  20. Investigation of Nuclear Data Libraries with TRIPOLI-4 Monte Carlo Code for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Lee, Y.-K.; Brun, E.

    2014-04-01

    The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.

  1. Design Study of 200MWth Gas Cooled Fast Reactor with Nitride (UN-PuN Fuel Long Life without Refueling

    Directory of Open Access Journals (Sweden)

    Syarifah Ratna Dewi

    2016-01-01

    Full Text Available Design study of 200 MWth Gas Cooled Fast Reactor with UN-PuN fuel long life without refueling has been done. GFR is one type reactor in Generation IV reactor system. It uses helium coolant and fast neutron spectrum. Helium is chemical inert, single phase and low neutron moderation. In this study the calculations are performed by using SRAC code with PIJ calculation for the fuel pin cell calculation and CITATION calculation for core calculation. The data libraries use JENDL 3.2. The variation fuel fractions are 50% until 60%. The diameter active core is 150 cm and the height active core is 100 cm. The reflector radial-axial width is 50 cm. The variation of the powers are 100 MWth up to 500 MWth. The high power causes the high k-eff value. The optimum design is reached when the power is 200 MWth, variation percentage Plutonium for fuel F1:F2:F3=9%:11%:13%. The comparation of fuel:cladding:coolant fraction = 55%:10%:35%. The cooling down time of Plutonium is nine months. The optimum k-eff value is 1.0142 with excess reactivity value 1.403%. The decay of Plutonium decrease k-eff value in the beginning of burn up.

  2. 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.

  3. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors; Estudio de las propiedades termofisicas y termohidraulicas del sodio para reactores rapidos enfriados por sodio

    Energy Technology Data Exchange (ETDEWEB)

    Vega R, A. K.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: a.karen.vr@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  4. Literature review on metallic fuel source term for sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Nam Duk; Bae, Moo Hoon; Shin, An Dong; Huh, Chang Wook [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2012-10-15

    Source term is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment, following a severe accident where a significant portion of the reactor core has melted. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. Apart from assessing the radiological consequences for siting, it is also important for designing filtering systems and even reactor components. Overly conservative source term for light water reactor, TID 14844 demands for very fast closure of main steam isolation valves, rapid startup of emergency diesels, and safety systems designed to mitigate gaseous iodine. In spite of this importance, most of the knowledge we have for SFR source term comes from the research performed before 1980s. Moreover, majority of the work on metallic fuels was done during the late 1950's through the 1960's. This paper reviews and summarizes the main characteristics of SFR source terms based on the available literatures.

  5. Performance of low smeared density sodium-cooled fast reactor metal fuel

    Science.gov (United States)

    Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  6. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

    Directory of Open Access Journals (Sweden)

    Franck Dechelette

    2014-01-01

    Full Text Available The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

  7. Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-29

    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

  8. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  9. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)

    2014-06-15

    In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

  10. 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.

  11. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  12. An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, M.; Farmer, M.; Grabaskas, D.

    2017-06-26

    The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination of gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Nilay K. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamilnadu (India). Dept. of Atomic Energy (DAE); Raj, Baldev [P.S. Govindaswamy Naidu (PSG) Institutions, Coimbatore, Tamilnadu (India)

    2013-11-15

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative studies of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton {sup registered} GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or ready-made supply. Part I addresses key areas of design shortlisting, strategy, development and unification with a backdrop of international evolution. (orig.)

  14. Safety properties of sodium-cooled fast reactors%钠冷快堆及其安全特性

    Institute of Scientific and Technical Information of China (English)

    徐銤; 杨红义

    2016-01-01

    钠冷快堆是第四代核能系统国际论坛(GIF)公布的6种第四代先进反应堆中研发进展最快、最接近满足商业核电厂需要的堆型。钠冷快堆因其在固有安全性以及可增殖核燃料、嬗变长寿命放射性废物等方面的优势,得到了世界各国的重视。文章以中国第一座钠冷快堆——中国实验快堆(China Experimental Fast Reactor,CEFR)为例,介绍了钠冷快堆在设计及运行方面的安全特性。%The sodium-cooled fast reactor is the fastest prototype and the closest to com-mercialization for nuclear power plants amongst the six types of fourth generation reactors, as an-nounced at the Generation IV International Forum. Many countries are paying more and more at-tention to the research and development of these reactors, due to the inherent safety features, effi-cient utilization of uranium with the breeding of the plutonium, and transmutation of long-lived ac-tinides. The design and operational safety characteristics of the China Experimental Fast Reactor are reviewed in this paper.

  15. Passive acoustic leak detection for sodium cooled fast reactors using hidden Markov models

    Energy Technology Data Exchange (ETDEWEB)

    Riber Marklund, A. [CEA, Cadarache, DEN/DTN/STCP/LIET, Batiment 202, 13108 St Paul-lez-Durance, (France); Kishore, S. [Fast Reactor Technology Group of IGCAR, (India); Prakash, V. [Vibrations Diagnostics Division, Fast Reactor Technology Group of IGCAR, (India); Rajan, K.K. [Fast Reactor Technology Group and Engineering Services Group of IGCAR, (India)

    2015-07-01

    Acoustic leak detection for steam generators of sodium fast reactors have been an active research topic since the early 1970's and several methods have been tested over the years. Inspired by its success in the field of automatic speech recognition, we here apply hidden Markov models (HMM) in combination with Gaussian mixture models (GMM) to the problem. To achieve this, we propose a new feature calculation scheme, based on the temporal evolution of the power spectral density (PSD) of the signal. Using acoustic signals recorded during steam/water injection experiments done at the Indira Gandhi Centre for Atomic Research (IGCAR), the proposed method is tested. We perform parametric studies on the HMM+GMM model size and demonstrate that the proposed method a) performs well without a priori knowledge of injection noise, b) can incorporate several noise models and c) has an output distribution that simplifies false alarm rate control. (authors)

  16. Mitigation of corrosion and mass transfer in sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Latge, C. [CEA Cadarache, Dir. de l' Energie Nucleaire, 13 - Saint-Paul-lez-Durance (France); Feron, D. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif-sur-Yvette (France)

    2009-07-01

    Full text of publication follows: Several coolants can be used for the development of the Fast Reactors, as sodium, gas, lead or lead-bismuth eutectic, and have been selected in the Generation IV forum. The high density energy requires a coolant with a very good thermal conductivity. Liquid sodium is such a medium which is liquid between 97.8 up to 880 C at dynamic pressure below 4 bars, and with compatible neutron-physical properties. Its viscosity is comparable to that of water and its compatibility with metallic materials is fairly satisfactory. It is however necessary to keep the conditions of operation within a range such that corrosion is limited. Several materials are suitable for use in liquid sodium reactors, among ferritic and austenitic steels and high temperature alloys with up to 32% nickel contents. The designer has however to consider the mass transfer between materials of different compositions. The exchange and transfer of non-metallic elements such as carbon or nitrogen has to be taken into account. The corrosion mechanisms of austenitic steels have been extensively studied and described in the literature: surface cleaning, austenitic dissolution, formation of a ferrite layer, steady state equilibrium and several models have been proposed: main parameters include oxygen content, sodium velocity and steel temperature. Operating experience has shown that, if there are no cladding failures, the main source of radioactivity in the primary circuit is the activated corrosion products, like {sup 54}Mn, {sup 51}Cr,..., induced by the activation of core materials which are dissolved into the sodium and mainly deposited in the coldest parts of the reactor i.e. the Intermediate Heat Exchanger (IHX) and pumps. Radio-cobalt such as {sup 60}Co are also produced and a low fraction is deposited in primary components. The corrosion rates estimated and the contamination induced by activated corrosion products observed in SFR like Phenix, JOYO, BN600, PFR, EBR2 have

  17. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  18. Simulation of Radioactive Corrosion Product in Primary Cooling System of Japanese Sodium-Cooled Fast Breeder Reactor

    Science.gov (United States)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54Mn and 60Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54Mn was estimated to constitute approximately 20 % and 60Co approximately 40 % in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO.

  19. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  20. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    Energy Technology Data Exchange (ETDEWEB)

    Song, Wei [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China); Wu, Yuanyu [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lès-Durance (France); Hu, Wenjun [China Institute of Atomic Energy, P. O. Box 275(34), Beijing (China); Zuo, Jiaxu, E-mail: zuojiaxu@chinansc.cn [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China)

    2015-11-15

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  1. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  2. Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

    Science.gov (United States)

    Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

    2017-04-01

    Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

  3. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  4. Conceptual Design of Passive Safety System for Lead-Bismuth Cooled Fast Reactor

    Science.gov (United States)

    Abdullah, A. G.; Nandiyanto, A. B. D.

    2016-04-01

    This paper presents the results of the conceptual design of passive safety systems for reactor power 225 MWth using Pb-Bi coolant. Main purpose of this research is to design of heat removal system from the reactor wall. The heat from the reactor wall is removed by RVACS system using the natural circulation from the atmosphere around the reactor at steady state. The calculation is performed numerically using Newton-Raphson method. The analysis involves the heat transfer systems in a radiation, conduction and natural convection. Heat transfer calculations is performed on the elements of the reactor vessel, outer wall of guard vessel and the separator plate. The simulation results conclude that the conceptual design is able to remove heat 1.33% to 4.67% from the thermal reactor power. It’s can be hypothesized if the reactor had an accident, the system can still overcome the heat due to decay.

  5. The European JASMIN Project for the Development of a New Safety Simulation Code, ASTEC-Na, for Na-cooled Fast Neutron Reactors

    OpenAIRE

    GIRAULT N.; VAN DORSSELAERE J.p.; Jacq, F.; BRILLANT G.; KISSANE Martin; BANDINI, G; Buck,M.; CHAMPIGNY J.; Hering, W; Perez-Martin, S.; Herranz, L; RAISON Philippe; Reinke, N; TUCEK Kamil; VERWAERDE D.

    2012-01-01

    The 4-year JASMIN collaborative project, involving 9 organizations, was launched by IRSN end of 2011 within the 7th European R&D Framework Programme on the enhancement of Na-cooled Fast Neutron Reactors (SFR) safety for a higher resistance to severe accidents. The project aims at developing a new European simulation code, ASTEC-Na, with a modern architecture, sufficiently flexible to account for innovative reactor designs and eventually new types of fuel and claddings and accounting for resul...

  6. 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...

  7. Stability analysis of the Korean prototype generation-IV sodium-cooled fast reactor using linear frequency domain approach

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ji; Ha, Pham Nhu Viet; Lim, Jae Yong; Hahn, Do Hee; Kang, Chang Mu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of). Fast Reactor Development Div.

    2016-03-15

    The Korea Atomic Energy Research Institute (KAERI) has been developing the 150 MWe Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR). The design concept is highly based on passive safety mechanisms, minimizing the need for engineered safety systems. Presently, it is of primary importance to assure the reactor dynamics and stability against small reactivity disturbances under power operating conditions. KAERI has therefore developed the NuSTAB code for stability analysis of the PGSFR. In NuSTAB, the neutron-kinetic and thermal-hydraulic coupling equations are linearized to form the characteristic equation, which is solved as a generalized eigenvalue problem for determining the decay ratio, an indicator of the system stability. In this paper, the stability of the PGSFR was analyzed by applying the point kinetic and spatial kinetic options in the NuSTAB code. System responses to temperature feedbacks including the Doppler effect, thermal expansion, coolant density change, and overall feedback were studied. The results indicate that the initial U and final TRU cores of the PGSFR are both inherently stable thanks to the temperature feedbacks.

  8. Development of Advanced 9Cr Ferritic-Martensitic Steels and Austenitic Stainless Steels for Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sham, Sam [ORNL; Tan, Lizhen [ORNL; Yamamoto, Yukinori [ORNL

    2013-01-01

    Ferritic-martensitic (FM) steel Grade 92, with or without thermomechanical treatment (TMT), and austenitic stainless steels HT-UPS (high-temperature ultrafine precipitate strengthening) and NF709 were selected as potential candidate structural materials in the U.S. Sodium-cooled Fast Reactor (SFR) program. The objective is to develop advanced steels with improved properties as compared with reference materials such as Grade 91 and Type 316H steels that are currently in nuclear design codes. Composition modification and/or processing optimization (e.g., TMT and cold-work) were performed to improve properties such as resistance to thermal aging, creep, creep-fatigue, fracture, and sodium corrosion. Testings to characterize these properties for the advanced steels were conducted by the Idaho National Laboratory, the Argonne National Laboratory and the Oak Ridge National Laboratory under the U.S. SFR program. This paper focuses on the resistance to thermal aging and creep of the advanced steels. The advanced steels exhibited up to two orders of magnitude increase in creep life compared to the reference materials. Preliminary results on the weldment performance of the advanced steels are also presented. The superior performance of the advanced steels would improve reactor design flexibility, safety margins and economics.

  9. Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

    Directory of Open Access Journals (Sweden)

    C. Bassi

    2008-01-01

    Full Text Available As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA, constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs is based on uncertainties propagation into thermal-hydraulic (T-H calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary for the core damage frequency (CDF, the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

  10. Numerical Analysis on the Free Fall Motion of the Control Rod Assembly for the Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se-Hong; Choi, Choengryul; Son, Sung-Man [ELSOLTEC, Yongin (Korea, Republic of); Kim, Jae-Yong; Yoon, Kyung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. However, because of its complex shape, it is difficult to estimate the drop time by theoretical method. In this study, numerical analysis has been carried out in order to estimate drop time and falling velocity of the control rod assembly to provide the underlying data for the design optimization. Numerical analysis has been carried out to estimate the drop time and falling velocity of the control rod assembly for sodium-cooled fast reactor. Before performing the numerical analysis for the control rod assembly, sphere dropping experiment has been carried out for verification of the CFD methodology. The result of the numerical analysis for the method verification is almost same as the result of the experiment. Falling velocity and drag force increase rapidly in the beginning. And then it goes to the stable state. When the piston head of the control rod assembly is inserted into the damper, the drag force increases instantaneously and the falling velocity decreases quickly. The falling velocity is reduced about 14 % by damper. The total drop time of the control rod assembly is about 1.47s. In the next study, the experiment for the control rod assembly will be carried out, and its result is going to be compared with the CFD analysis result.

  11. Development of Pool-type Sodium-cooled Fast Reactor System Analysis Code%池式钠冷快堆系统分析程序开发

    Institute of Scientific and Technical Information of China (English)

    王晋; 张东辉; 胡文军

    2016-01-01

    针对池式钠冷快堆的特点,在对快堆系统的水力模型、热工模型和中子动力学模型进行详细分类和建模的基础上,利用 FORTRAN95语言开发了可用于池式钠冷快堆事故分析的系统分析程序(FASYS程序)。以中国实验快堆为计算对象对FASYS程序模型进行了初步验证,所获得的结果和试验值与其他系统程序计算值符合良好,证明了所开发的系统分析程序的正确性。%According to the characteristics of pool‐type sodium‐cooled fast reactor ,and with the fast reactor hydraulic model , thermal model and neutron kinetics model thoroughly classified and developed ,a fast reactor system analysis code (FASYS code) was developed by FORTRAN95 language for pool‐type sodium‐cooled fast reactor acci‐dent analysis .Transient conditions in CEFR were calculated with FASYS code and the results were used for code validation .The calculation results are consistent with the test data and other fast reactor system analysis code results , and the correctness of the FASYS code is proved .

  12. Dynamic simulation of a sodium-cooled fast reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Shinaishin, M.A.M.

    1976-08-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

  13. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

    Science.gov (United States)

    Marcille, T. F.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.; Poston, D. I.; Kapernick, R. J.

    2006-01-01

    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  14. Analysis of dashpot performance for rotating control drums of a lithium cooled fast reactor concept

    Science.gov (United States)

    Wenzler, C. J.

    1972-01-01

    A dashpot was incorporated in the design of the drive train of the rotating control drum to prevent shock damage to the control drum and drive train at the termination of a scram action. A rotating vane dashpot using reactor coolant lithium as a damping fluid appears to be the best candidate of the various damping devices explored. A performance analysis, results and discussion of vane type dashpots are presented.

  15. Implementation Plan for Qualification of Sodium-Cooled Fast Reactor Technology Information

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This document identifies and discusses implementation elements that can be used to facilitate consistent and systematic evaluation processes relating to quality attributes of technical information (with focus on SFR technology) that will be used to support licensing of advanced reactor designs. Information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The approach for determining acceptability of test data, analysis, and/or other technical information is based on guidance provided in INL/EXT-15-35805, “Guidance on Evaluating Historic Technology Information for Use in Advanced Reactor Licensing.” The implementation plan can be adopted into a working procedure at each of the national laboratories performing data qualification, or by applicants seeking future license application for advanced reactor technology.

  16. 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.

  17. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

    Directory of Open Access Journals (Sweden)

    Girardin G.

    2013-03-01

    Full Text Available PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970’s to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices – including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%.

  18. Pre-conceptual core design of a small modular fast reactor cooled by supercritical CO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Baolin; Cao, Liangzhi; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); Yuan, Xianbao, E-mail: ztsbaby@163.com [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); College of Mechanical & Power Engineering, China Three Gorges University, No 8, Daxue Road, Yichang 443002, Hubei (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China)

    2016-04-15

    Abstracts: A Small Modular fast reactor cooled by Supercritical CO{sub 2} (SMoSC) is pre-conceptually designed through three-dimensional coupled neutronics/thermal-hydraulics analysis. The power rating of the SMoSC is designed to be 300 MW{sub th} to meet the energy demand of small electrical grids. The excellent thermal properties of supercritical CO{sub 2} (S-CO{sub 2}) are employed to obtain a high thermal efficiency of about 40% with an electric output of 120 MWe. MOX fuel is utilized in the core design to improve fuel efficiency. The tube-in-duct (TID) assembly is applied to get lower coolant volume fraction and reduce the positive coolant void reactivity. According to the coupled neutronics/thermal-hydraulics calculations, the coolant void reactivity is kept negative throughout the whole core life. With a specific power density of 9.6 kW/kg and an average discharge burnup of 70.1 GWd/tHM, the SmoSC can be operated for 20 Effective Full Power Years (EFPYs) without refueling.

  19. 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.

  20. 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.

  1. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 61-Rod Bundle for a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Kim, Hyungmo; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Jeong, Ji-Young; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are crucial factor for the design code verification and validation. Wrapped wires make a cross flow in a circumference of the fuel rod, and this effect lets flow be mixed. Therefore the sub-channel analysis method is commonly used for thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. To measure flow mixing characteristics, a wire mesh sensing technique can be useful method. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, the recent reports that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In this study, mixing experiments were conducted successfully at a hexagonally arrayed 61-pin wire-wrapped fuel rod bundle test section. Wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable.

  2. Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

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

    Directory of Open Access Journals (Sweden)

    Stanisz Przemysław

    2015-09-01

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

  4. Scaling approach and thermal-hydraulic analysis in the reactor cavity cooling system of a high temperature gas -cooled reactor and thermal-jet mixing in a sodium fast reactor

    Science.gov (United States)

    Omotowa, Olumuyiwa A.

    This dissertation develops and demonstrates the application of the top-down and bottom-up scaling methodologies to thermal-hydraulic flows in the reactor cavity cooling system (RCCS) of the high temperature gas reactor (HTGR) and upper plenum of the sodium fast reactor (SFR), respectively. The need to integrate scaled separate effects and integral tests was identified. Experimental studies and computational tools (CFD) have been integrated to guide the engineering design, analysis and assessment of this scaling methods under single and two-phase flow conditions. To test this methods, two applicable case studies are considered, and original contributions are noted. Case 1: "Experimental Study of RCCS for the HTGR". Contributions include validation of scaling analysis using the top-down approach as guide to a ¼-scale integral test facility. System code, RELAP5, was developed based on the derived scaling parameters. Tests performed included system sensitivity to decay heat load and heat sink inventory variations. System behavior under steady-state and transient scenarios were predicted. Results show that the system has the capacity to protect the cavity walls from over-heating during normal operations and provide a means for decay heat removal under accident scenarios. A full width half maximum statistical method was devised to characterize the thermal-hydraulics of the non-linear two-phase oscillatory behavior. This facilitated understanding of the thermal hydraulic coupling of the loop segments of the RCCS, the heat transfer, and the two-phase flashing flow phenomena; thus the impact of scaling overall. Case 2: "Computational Studies of Thermal Jet Mixing in SFR". In the pool-type SFR, susceptible regions to thermal striping are the upper instrumentation structure and the intermediate heat exchanger (IHX). We investigated the thermal mixing above the core to UIS and the potential impact due to poor mixing. The thermal mixing of dual-jet flows at different

  5. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-02-12

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  6. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    Science.gov (United States)

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Aziz, Ferhat; Permana, Sidik; Sekimoto, Hiroshi

    2014-02-01

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  7. Generation IV nuclear energy system initiative. Large GFR core subassemblydesign for the Gas-Cooled Fast Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, E. A.; Kulak, R. F.; Therios, I. U.; Wei, T. Y. C.

    2006-07-31

    Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics

  8. Thermal hydraulic investigations on porous blockage in a prototype sodium cooled fast reactor fuel pin bundle

    Energy Technology Data Exchange (ETDEWEB)

    Raj, M.Naveen; Velusamy, K., E-mail: kvelu@igcar.gov.in; Maity, Ram Kumar

    2016-07-15

    clad temperature is found to be a strong function of porosity, with enhanced clad temperature for smaller porosity. Fuel-clad that are partly exposed to blockage are subjected to large circumferential temperature variation and the resulting huge thermal stress. Further, for a six subchannel blockage with 40% porosity and 0.5 mm mean particle diameter the critical length is 80 mm, whereas for the same blockage the critical length reduces to <7 mm when its porosity reduces to 5%. Six subchannel blockage with 60% porosity and 0.5 mm mean particle diameter, does not induce boiling even up to a blockage height of 400 mm. For a single subchannel blockage with one helical pitch length, there is no risk of sodium boiling even for porosity as low as 5%. The results of the present study would act as safety and monitoring criteria during the operation of the reactor.

  9. Reliability study of a special decay heat removal system of a gas-cooled fast reactor demonstrator

    Energy Technology Data Exchange (ETDEWEB)

    Burgazzi, Luciano, E-mail: luciano.burgazzi@enea.it

    2014-12-15

    The European roadmap toward the development of generation IV concepts addresses the safety and reliability assessment of the special system designed for decay heat removal of a gas-cooled fast reactor demonstrator (GFRD). The envisaged system includes the combination of both active and passive means to accomplish the fundamental safety function. Failure probabilities are calculated on various system configurations, according to either pressurized or depressurized accident events under investigation, and integrated with probabilities of occurrence of corresponding hardware components and natural circulation performance assessment. The analysis suggests the improvement of measures against common cause failures (CCF), in terms of an appropriate diversification among the redundant systems, to reduce the system failure risk. Particular emphasis is placed upon passive system reliability assessment, being recognized to be still an open issue, and the approach based on the functional reliability is adopted to address the point. Results highlight natural circulation as a challenging factor for the decay heat removal safety function accomplishment by means of passive devices. With the models presented here, the simplifying assumptions and the limited scenarios considered according to the level of definition of the design, where many systems are not yet established, one can conclude that attention has to be paid to the functional aspects of the passive system, i.e. the ones not pertaining to the “hardware” of the system. In this article the results of the analysis are discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The design diversity of the components undergoing CCFs can be effective for the improvement and some accident management measures are also possible by making use of the long grace period in GFRD.

  10. Fast Reactor Fuel Type and Reactor Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  11. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

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

  12. Design Improvement of Iso-Kinetic Flow Sampling Device at Subchannel in a Wire-Wrapped 37-pin Fuel Assembly for a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong-Won; Kim, Hyungmo; Ko, Yung-Joo; Chang, Seok-Kyu; Choi, Hae Seob; Euh, Dong-kin; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Securing the structural integrity of a fuel assembly during reactor operation is of utmost importance in order to prevent reactor severe accident like the Fukushima nuclear power plant through a flow characteristics tests with test assembly scaled down from a prototype reactor of a sodium-cooled fast reactor (SFR). To evaluate uncertainty is very important to ensure reliability at the results of the fuel assembly. Therefore the sub-channel analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. In KAERI, two sub-channel analysis codes (SLTHEN, MATRA-LMR) are considered to utilize for the design of the prototype reactor. In this study, design improvement of iso-Kinetic flow sampling device at sub-channel in a wire-wrapped 37-pin fuel assembly for a sodium cooled fast reactor is conducted for decreasing misalignment sensitivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In KAERI, two subchannel analysis codes are considered to be utilized for the design of the prototype reactor. In this study, the X-axis probe misalignment error is 2.5%, the Y-axis probe misalignment error is 0.9% and flowmeter and DA equipment error is 0.2%. As shown in above results, the misalignment error was the highest factor in uncertainty analysis. To solve the problem, design improvement of iso-kinetic flow sampling device at subchannel in a wire-wrapped 37-pin fuel assembly is practiced for decreasing misalignment sensitivity error.

  13. Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal

  14. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  15. 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

  16. Safety aspects of fuel behaviour during faults and accidents in pressurised water reactors and in liquid sodium cooled fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H. (UKAEA Information Services Branch, London); Matthews, J.R. (UKAEA Harwell Lab. (UK). Theoretical Physics Div.); Potter, P.E. (UKAEA Harwell Lab. (UK). Chemistry Div.)

    1989-07-01

    The good safety record of electrical power generating reactors in the European Community is based on a substantial effort to understand the safety characteristics of the reactors and their fuel. In this paper the present state of knowledge of oxide fuels used in current European reactors is reviewed. The main theme of the paper is the importance of the role of fission products and the chemical state of the fuel on all aspects of fuel behaviour. The paper is split into two parts. The first part deals with those aspects specific to water reactors using UO{sub 2} based fuels. The second part of the paper deals with mixed-oxide fuels and the sodium cooled reactors. In each part the following aspects are described: Chemical constitution of the fuel; fuel performance and failure limits; failed fuel behaviour; fuel behaviour in accidents; and the interactions in degraded cores after hypothetical accidents. Future directions of safety related fuel work in Europe are identified. (orig.).

  17. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    Science.gov (United States)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Monado, Fiber; Sekimoto, Hiroshi

    2012-06-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of ith region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

  18. 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.

  19. Adaptation and implementation of the TRACE code for transient analysis in designs lead cooled fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2015-07-01

    Lead-Cooled Fast Reactor (LFR) has been identified as one of promising future reactor concepts in the technology road map of the Generation IVC International Forum (GIF)as well as in the Deployment Strategy of the European Sustainable Nuclear Industrial Initiative (ESNII), both aiming at improved sustainability, enhanced safety, economic competitiveness, and proliferation resistance. This new nuclear reactor concept requires the development of computational tools to be applied in design and safety assessments to confirm improved inherent and passive safety features of this design. One approach to this issue is to modify the current computational codes developed for the simulation of Light Water Reactors towards their applicability for the new designs. This paper reports on the performed modifications of the TRACE system code to make it applicable to LFR safety assessments. The capabilities of the modified code are demonstrated on series of benchmark exercises performed versus other safety analysis codes. (Author)

  20. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  1. Liquid metal cooled reactors for space power applications

    Science.gov (United States)

    Bailey, S.; Vaidyanathan, S.; Van Hoomissen, J.

    1985-01-01

    The technology basis for evaluation of liquid metal cooled space reactors is summarized. Requirements for space nuclear power which are relevant to selection of the reactor subsystem are then reviewed. The attributes of liquid metal cooled reactors are considered in relation to these requirements in the areas of liquid metal properties, neutron spectrum characteristics, and fuel form. Key features of typical reactor designs are illustrated. It is concluded that liquid metal cooled fast spectrum reactors provide a high confidence, flexible option for meeting requirements for SP-100 and beyond.

  2. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. Pt. II. R and D necessities and development across the world. A review

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Nilay K. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamilnadu (India). Dept. of Atomic Energy (DAE); Raj, Baldev [P.S. Govindaswamy Naidu (PSG) Institutions Coimbatore, Tamilnadu (India)

    2013-12-15

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative study of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton {sup registered} GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or readymade supply. R and D necessities for inflatable seals and their development across the world are given closer look in Part II of the review in continuation of Part I. (orig.)

  3. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth

    2002-09-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, C

    2010-02-22

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

  5. FAST NEUTRONIC REACTOR

    Science.gov (United States)

    Snell, A.H.

    1957-12-01

    This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

  6. MEANS FOR COOLING REACTORS

    Science.gov (United States)

    Wheeler, J.A.

    1957-11-01

    A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

  7. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

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

  8. Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T., E-mail: murakami@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Eguchi, Y., E-mail: eguchi@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Oyama, K., E-mail: kazuhiro_oyama@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan); Watanabe, O., E-mail: osamu4_watanabe@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan)

    2015-07-15

    Highlights: • The natural circulation characteristics of a loop-type SFR are examined by a water test. • The performance of decay heat removal system is evaluated using a similarity law. • The effects of flow deviation in the parallel piping of a primary loop are clarified. • The reproducibility of the natural circulation test is confirmed. - Abstract: Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant

  9. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Science.gov (United States)

    Ariani, Menik; Satya, Octavianus Cakra; Monado, Fiber; Su'ud, Zaki; Sekimoto, Hiroshi

    2016-03-01

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on "Region-8" and "Region-10" core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  10. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Energy Technology Data Exchange (ETDEWEB)

    Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber [Department of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University, jl Palembang-Prabumulih km 32 Indralaya OganIlir, South of Sumatera (Indonesia); Su’ud, Zaki [Nuclear and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, jlGanesha 10, Bandung (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, 2-12-11N1-17 Ookayama, Meguro-Ku, Tokyo (Japan)

    2016-03-11

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  11. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    Science.gov (United States)

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-01

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

  12. Flow distribution and pressure loss in subchannels of a wire-wrapped 37-pin rod bundle for sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Euh, Dong Jin; Choi, Hae Seob; Kim, Hyung Mo; Choi, Sun Rock; Lee, Hyeong Yeon [Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and 60 degrees C (equivalent to Re ∼ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

  13. Adaptation and implementation of the TRACE code for transient analysis on designs of cooled lead fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2014-07-01

    The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)

  14. Small Liquid Metal Cooled Reactor Safety Study

    Energy Technology Data Exchange (ETDEWEB)

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N

    2005-11-02

    The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S requires negative void worth throughout

  15. 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.

  16. 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.

  17. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth

    2002-06-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  18. Dynamic model of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vaidyanathan, G., E-mail: vaidya@igcar.gov.i [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Kasinathan, N.; Velusamy, K. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2010-04-15

    Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-15

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

  20. Investigation of molten salt fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kubota, Kenichi; Konomura, Mamoru [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2002-05-01

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

  1. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) neat transport system dynamics and steam generator control

    Science.gov (United States)

    Brukx, J. F. L. M.

    1982-06-01

    Loop type LMFBR heat transport system dynamics after reactor shutdown and during subsequent decay heat removal are considered with emphasis on steam generator dynamics including the development and evaluation of various post-scram steam generator control systems, and natural circulation of the sodium coolant, including the influence of superimposed free convection on forced convection heat transfer and pressure drop. The normal operating and decay heat removal functions of the overall heat transport system are described.

  2. Nuclear data uncertainty propagation for a lead-cooled fast reactor: Combining TMC with criticality benchmarks for improved accuracy

    OpenAIRE

    Alhassan, Erwin

    2014-01-01

    For the successful deployment of advanced nuclear systems and for optimization of current reactor designs, high quality and accurate nuclear data are required. Before nuclear data can be used in applications, they are first evaluated, benchmarked against integral experiments and then converted into formats usable for applications. The evaluation process in the past was usually done by using differential experimental data which was then complimented with nuclear model calculations. This trend ...

  3. Nuclear data uncertainty quantification and data assimilation for a lead-cooled fast reactor : Using integral experiments for improved accuracy

    OpenAIRE

    Alhassan, Erwin

    2015-01-01

    For the successful deployment of advanced nuclear systems and optimization of current reactor designs, high quality nuclear data are required. Before nuclear data can be used in applications they must first be evaluated, tested and validated against a set of integral experiments, and then converted into formats usable for applications. The evaluation process in the past was usually done by using differential experimental data which was then complemented with nuclear model calculations. This t...

  4. Irradiation behavior of metallic fast reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

  5. The prospect of uranium nitride (UN-PuN) fuel for 25- 100MWe gas cooled fast reactor long life without refuelling

    Science.gov (United States)

    Syarifah, R. D.; Su'ud, Z.; Basar, K.; Irwanto, D.

    2016-11-01

    The prospect of uranium nitride (UN-PuN) fuel for 25-100MWe Gas Cooled Fast Reactor has been done. This research use helium coolant which has low neutron moderation, chemical inert and single phase. This study use natural uranium and plutonium. Plutonium taken from spent fuel of LWR (Light Water Reactor). So, it can reduced spent fuel in the world. The calculation use SRAC2006 and JENDL 4.0 for the data libraries. First, we calculate PIJ for fuel pin cell calculation and CITATION for core calculation. The reflector radial-axial width is 50 cm. The variation of fuel fraction is 40% until 65%, cladding 10%, and moderator 25% up to 50%. The variation of the power is 75-300 MWth (25-100 MWe). The calculation of survey parameter has been done. The variation of percentage plutonium is 7% up to 13%. We have optimum k-eff value in percentage of plutonium 11%. The high powers cause k-eff value high too. Second, the core configuration divided by three variation fuel (F1, F2, and F3). F1 is located in the central core, F2 middle core and F3 outer core. The variation percentage Plutonium for fuel F1:F2:F3 = 8%:10%:12%. The increasing power level make the burn up level increase. All case can reach burn up time plus than 20 years. The thermal powers increase cause the peak power density increase. The power 150 MWth, 225 MWth, and 300 MWth have excess reactivity (%Ak/k) less than 2%.

  6. 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.

  7. Measurement of the Residual Stresses and Investigation of Their Effects on a Hardfaced Grid Plate due to Thermal Cycling in a Pool Type Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    S. Balaguru

    2016-01-01

    Full Text Available In sodium-cooled fast reactors (SFR, grid plate is a critical component which is made of 316 L(N SS. It is supported on core support structure. The grid plate supports the core subassemblies and maintains their verticality. Most of the components of SFR are made of 316 L(N/304 L(N SS and they are in contact with the liquid-metal sodium which acts as a coolant. The peak operating temperature in SFR is 550°C. However, the self-welding starts at 500°C. To avoid self-welding and galling, hardfacing of the grid plate has become necessary. Nickel based cobalt-free colmonoy 5 has been identified as the hardfacing material due to its lower dose rate by Plasma Transferred Arc Welding (PTAW. This paper is concerned with the measurement and investigations of the effects of the residual stress generated due to thermal cycling on a scale-down physical model of the grid plate. Finite element analysis of the hardfaced grid plate model is performed for obtaining residual stresses using elastoplastic analysis and hence the results are validated. The effects of the residual stresses due to thermal cycling on the hardfaced grid plate model are studied.

  8. Investigation of Reactivity Feedback Mechanism of Axial and Radial Expansion Effect of Metal-Fueled Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seong, Seung-Hwan; Choi, Chi-Woong; Jeong, Tae-Kyung; Ha, Gi-Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The major inherent reactivity feedback models for a ceramic fuel used in a conventional light water reactor are Doppler feedback and moderator feedback. The metal fuel has these two reactivity feedback mechanisms previously mentioned. In addition, the metal fuel has two more reactivity feedback models related to the thermal expansion phenomena of the metal fuel. Since the metal fuel has a good capability to expand according to the temperature changes of the core, two more feedback mechanisms exist. These additional two feedback mechanism are important to the inherent safety of metal fuel and can make metal-fueled SFR safer than oxide-fueled SFR. These phenomena have already been applied to safety analysis on design extended condition. In this study, the effect of these characteristics on power control capability was examined through a simple load change operation. The axial expansion mechanism is induced from the change of the fuel temperature according to the change of the power level of PGSFR. When the power increases, the fuel temperatures in the metal fuel will increase and then the reactivity will decrease due to the axial elongation of the metal fuel. To evaluate the expansion effect, 2 cases were simulated with the same scenario by using MMS-LMR code developed at KAERI. The first simulation was to analyze the change of the reactor power according to the change of BOP power without the reactivity feedback model of the axial and radial expansion of the core during the power transient event. That is to say, the core had only two reactivity feedback mechanism of Doppler and coolant temperature.

  9. Assessment of Proliferation Resistance of Closed Nuclear Fuel Cycle System with Sodium Cooled Fast Reactors Using INPRO Evaluation Methodology

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-15

    Using the INPRO methodology, the proliferation resistance of an innovative nuclear energy system(INS) defined as a closed nuclear fuel cycle system consisting of KALIMER and pyroprocessing, has been assessed. Considering a very early development stage of the INS concept, the PR assessment is carried out based on intrinsic features, if required information and data are not available. The PR assessment of KALIMER and JSFR using the INPRO methodology affirmed that an adequate proliferation resistance has been achieved in both INSs CNFC-SFR, considering the assessor's progress and maturity of design development. KALIMER and JSFR are developed or being developed conforming to the targets and criteria defined for developing Gen IV nuclear reactor system. Based on these assessment results, proliferation resistance and physical protection(PR and PP) of KALIMER and JSFR are evaluated from the viewpoint of requirements for future nuclear fuel cycle system. The envisioned INSs CNFC-SFR rely on active plutonium management based on a closed fuel cycle, in which a fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The proliferation resistance assessment should be an ongoing analysis that keeps up with the progress and maturity of the design of Gen IV SFR.

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

    Indian Academy of Sciences (India)

    S C Chetal; P Chellapandi

    2013-10-01

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

  11. Design of alumina forming FeCrAl steels for lead or lead-bismuth cooled fast reactors

    Science.gov (United States)

    Lim, Jun; Hwang, Il Soon; Kim, Ji Hyun

    2013-10-01

    Iron-chromium-aluminum alloys containing 15-20 wt.% Cr and 4-6 wt.% Al have shown excellent corrosion resistance in the temperature range up to 600 °C or higher in liquid lead and lead-bismuth eutectic environments by the formation of protective Al2O3 layers. However, the higher Cr and Al concentrations in ferritic alloys could be problematic because of severe embrittlement in the manufacturing process as well as in service, caused by the formation of brittle phases. For this reason, efforts worldwide have so far mainly focused on the development of aluminizing surface treatments. However, aluminizing surface treatments have major disadvantages of cost, processing difficulties and reliability issues. In this study, a new FeCrAl alloy is proposed for structural materials in lead and lead-bismuth cooled nuclear applications. The alloy design relied on corrosion experiments in high temperature lead and lead-bismuth eutectic environments and computational thermodynamic calculations using the commercial software, JMatPro. The design of new alloys has focused on the optimization of Cr and Al levels for the formation of an external Al2O3 layer which can provide excellent oxidation and corrosion resistance in liquid lead alloys in the temperature range 300-600 °C while still retaining workable mechanical properties.

  12. Gas-cooled reactors: the importance of their development

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1979-06-01

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

  13. Enhancement of Irradiation Capability of the Experimental Fast Reactor Joyo

    Science.gov (United States)

    Maeda, Shigetaka; Serine, Takashi; Aoyama, Takafumi; Suzuki, Soju

    2009-08-01

    The experimental fast reactor Joyo is the first sodium-cooled fast reactor in Japan. One of its primary missions is to perform irradiation tests of fuel and structural materials to support the development of fast reactors. The MK-III high performance core upgrade to enhance the irradiation testing capabilities was completed in 2003. In order to expand Joyo's capabilities for innovative irradiation testing applications, neutron spectrum tailoring, lower irradiation temperature, movable sample devices and fast neutron beam holes are being considered. This program responds to existing irradiation needs and aims to further expand capabilities for a variety of irradiation tests.

  14. A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

    OpenAIRE

    Billings, Jay Jay; Deyton, Jordan H.; Hull, S. Forest; Lingerfelt, Eric J.; Wojtowicz, Anna

    2014-01-01

    Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) wi...

  15. Heterogeneous Transmutation Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even neutron number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both non-flattened and flattened core geometries. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of comparable size. A mass balance analysis revealed that the heterogeneous design may reduce the number of fast reactors needed to close the current once-through light water reactor fuel cycle.

  16. Sodium fast reactor evaluation: Core materials

    Science.gov (United States)

    Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon; Raison, J. P.; Mizuno, T.; Delage, F.; Carmack, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

  17. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  18. Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

    2006-09-01

    Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature

  19. Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium–Plutonium Carbide based fuel and 300 MWt Power

    Science.gov (United States)

    Clief Pattipawaej, Sandro; Su’ud, Zaki

    2017-01-01

    A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

  20. Improvement of Neutronics Calculation Methods for Fast Reactors

    OpenAIRE

    Takeda, Toshikazu

    2011-01-01

    To accurately estimate neutronics properties of fast reactors, particularly Japan Sodium-cooled Fast Reactor of1,500 MW electric, calculational methods are being improved in Japan.This paper describes the planning and the ongoing development of the neutronics calculation methods in the fieldof 1) assembly calculations including the calculations of effective cross sections, 2) core calculations and 3) uncertaintyevaluation and uncertainty reduction.

  1. Actinide management with commercial fast reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-31

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

  2. Actinide management with commercial fast reactors

    Science.gov (United States)

    Ohki, Shigeo

    2015-12-01

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

  3. Application Study on Ultrasonic Imaging Technique in Sodium-cooled Fast Reactor in Foreign Countries%国外超声成像技术在钠冷快堆中的应用研究

    Institute of Scientific and Technical Information of China (English)

    过明亮; 吕兆福; 段天英

    2014-01-01

    发展超声成像技术应用于钠冷快堆,对于提高反应堆安全具有重要意义。为此,国外多个国家开展了钠下超声成像技术的相关研究。本文通过对国外超声成像技术在钠冷快堆中的应用论证,以及钠回路实验台架的实验数据的介绍和研究,描述并分析了该技术在国外的研发进展和设计中的关键性因素,从而得出钠冷快堆中运用超声成像系统技术的可行性和一些亟待解决的问题。%Development of ultrasonic imaging technique used in sodium-cooled fast reactor is very important for improving the safety of the reactor.Therefore,many countries carried out study on under-sodium ultrasonic imaging technique.Based on the application of ultrasonic imaging technique to demonstrate the sodium-cooled fast reactor,and the introduction and study on the experimental data of sodium loop testing platform,this paper describes and analyzes the key factor of R&D progress abroad in this regard,and propose the feasibility of using ultrasonic imaging system technique in sodium-cooled fast reactor and some problems need to be solved.

  4. 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,

  5. 钠冷快堆超导钠泵建模与性能研究%Modeling and Performance Research on Sodium Cooled Fast Reactor Superconductive Sodium Pump

    Institute of Scientific and Technical Information of China (English)

    杨志达; 赵佳宁; 韩伟实

    2014-01-01

    为克服钠冷快堆机械式钠泵机械磨损、噪声大、泄漏以及普通电磁泵流量小、扬程小等问题,提出了大流量鞍型磁体的电磁泵作为驱动钠循环的主泵设计方案。对其结构进行了研究并建立了相应的数学模型,利用M atlab编写程序进行了不同电流、磁感应强度和温度条件下泵的扬程、流量和效率的性能研究。结果表明,扬程随通道宽度的增加、电流的减小、磁感应强度的减小而减小,效率随电流的减小、磁感应强度的减小而减小,温度高于400℃时由于接触电阻的降低可使效率提高,鞍型超导钠泵的流量可由电流、磁感应强度控制,但较为实用的是电流控制。该研究可为具体设计提供依据。%In order to overcome the shortcoming of sodium cooled fast reactor mechanical sodium pump including mechanical wear ,large noise ,sodium leakage and the shortcom-ing of electromagnetic pump including small flow rate and small pump head ,the large flow rate sodium pump design based on saddle superconductive magnet was put forward . The structure of sodium pump was researched and the mathematic model was estab-lished .The performance including pump head ,flow rate and efficiency were researched at different currents ,magnetic flux densities and temperatures using the computer code programmed with Matlab .The result shows that the pump head will reduce with the channel width increasing ,current reducing and magnetic flux density reducing ,and the efficiency will reduce with the current reducing and magnetic flux density reducing .If the temperature overtops 400 ℃ ,the efficiency will be high because of contact resistance disappearing .T he flow rate of the saddle type superconductive sodium pump can be controlled by current and magnetic flux density ,and the current is more suitable .This research can provide foundation for concreting design .

  6. Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

    Energy Technology Data Exchange (ETDEWEB)

    Fabbris, Olivier [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Dardour, Saied, E-mail: saied.dardour@cea.fr [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Blaise, Patrick [CEA DEN/DER/SPEX, 13108 Saint-Paul-Lez-Durance (France); Ferrasse, Jean-Henry [Aix-Marseille Université, CNRS, ECM, M2P2 UMR 7340, 13451 Marseille (France); Saez, Manuel [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France)

    2016-08-15

    Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

  7. Heterogeneous Recycling in Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-30

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

  8. 可用于小型铅铋冷快堆的核能制氢技术分析%Technical Analysis of Nuclear Hydrogen Production in Small Pb-Bi Cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    孙征; 吴晓春; 李龙; 邵静

    2016-01-01

    核能制氢作为一种有前景的大规模制氢方法,得到广泛研究。该文介绍了适用于核能制氢的反应堆堆型,以及可用于核能制氢的主要方法,并对可用于小型铅铋冷快堆的核能制氢技术进行了分析。分析结果表明,小型铅铋冷快堆制氢的潜在技术路线为热化学裂解水溴钙循环或甲烷直接裂解法。%As a promising massive way,nuclear hydrogen production is being extensively investigated across the world.In this paper,the reactor types and main methods which could be used in nuclear hydrogen production were introduced,and the techniques used in Small Pb-Bi cooled fast reactor for nuclear hydrogen production were investigated.As a result,the potential technical path in Small Pb-Bi cooled fast reactor for nuclear hydrogen production were Ca-Br-Fe thermo-chemical process and methane direct pyrolysis method.

  9. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  10. Methods for quantifying uncertainty in fast reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-07

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

  11. Fast breeder reactor protection system

    Science.gov (United States)

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  12. 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.

  13. Integrated system for temperature-controlled fast protein liquid chromatography comprising improved copolymer modified beaded agarose adsorbents and a travelling cooling zone reactor arrangement.

    Science.gov (United States)

    Müller, Tobias K H; Cao, Ping; Ewert, Stephanie; Wohlgemuth, Jonas; Liu, Haiyang; Willett, Thomas C; Theodosiou, Eirini; Thomas, Owen R T; Franzreb, Matthias

    2013-04-12

    An integrated approach to temperature-controlled chromatography, involving copolymer modified agarose adsorbents and a novel travelling cooling zone reactor (TCZR) arrangement, is described. Sepharose CL6B was transformed into a thermoresponsive cation exchange adsorbent (thermoCEX) in four synthetic steps: (i) epichlorohydrin activation; (ii) amine capping; (iii) 4,4'-azobis(4-cyanovaleric acid) immobilization; and 'graft from' polymerization of poly(N-isopropylacrylamide-co-N-tert-butylacrylamide-co-acrylic acid-co-N,N'-methylenebisacrylamide). FT-IR, (1)H NMR, gravimetry and chemical assays allowed precise determination of the adsorbent's copolymer composition and loading, and identified the initial epoxy activation step as a critical determinant of 'on-support' copolymer loading, and in turn, protein binding performance. In batch binding studies with lactoferrin, thermoCEX's binding affinity and maximum adsorption capacity rose smoothly with temperature increase from 20 to 50 °C. In temperature shifting chromatography experiments employing thermoCEX in thermally jacketed columns, 44-51% of the lactoferrin adsorbed at 42 °C could be desorbed under binding conditions by cooling the column to 22 °C, but the elution peaks exhibited strong tailing. To more fully exploit the potential of thermoresponsive chromatography adsorbents, a new column arrangement, the TCZR, was developed. In TCZR chromatography, a narrow discrete cooling zone (special assembly of copper blocks and Peltier elements) is moved along a bespoke fixed-bed separation columnfilled with stationary phase. In tests with thermoCEX, it was possible to recover 65% of the lactoferrin bound at 35 °C using 8 successive movements of the cooling zone at a velocity of 0.1mm/s; over half of the recovered protein was eluted in the first peak in more concentrated form than in the feed. Intra-particle diffusion of desorbed protein out of the support pores, and the ratio between the velocities of the cooling

  14. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  15. Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

    Science.gov (United States)

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

    2017-07-01

    Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (D

  16. 可视化钠冷快堆工程信息系统的研究和设计%Viewing Sodium Cooled Fast Reactor Research and Design Engineering Information Systems

    Institute of Scientific and Technical Information of China (English)

    林全兵

    2014-01-01

    October 24, 2012, the State Council approved the "Nuclear Security Plan (2011-2020)" and "long-term nuclear power development plan (2011-2020)." This means that since the Fukushima nuclear leak in Japan since shut down the nuclear power project in China for nearly 20 months after the last full restart. Nuclear construction and re-enter the fast track scale development, on the basis of the domestic and international nuclear power project management systems, 3D visualization techniques designed to analyze the results of relevant research studies, combined with the characteristics of the sodium cooled fast reactor construction project information system application information integration technology to build a visual sodium-cooled fast reactor engineering information systems management platform to meet the use of nuclear power construction project management unit. Establish the visual content of the paper on sodium-cooled fast reactor engineering information platform has universal reference and high practical value.%2012年10月24日,国务院常务会议讨论通过了《核电安全规划(2011-2020年)》和《核电中长期发展规划(2011-2020年)》。这意味着,自日本发生福岛核泄漏以来,停摆了近20个月之后的中国核电项目终于全面重启。核电建造又重新进入了快速规模化发展的轨道,本文在对国内外核电项目管理系统、三维可视化设计技术相关研究成果进行分析研究的基础上,结合钠冷快堆工程信息系统的施工特点,应用信息集成技术,建设一个可视化的钠冷快堆工程信息系统管理平台,满足核电建造单位项目管理使用。本文的内容对可视化钠冷快堆工程信息平台的建立有普遍的借鉴意义和较高的实用价值。

  17. Parameter analysis calculation on characteristics of portable FAST reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

  18. Effects of a Mixed Zone on TGO Displacement Instabilities of Thermal Barrier Coatings at High Temperature in Gas-Cooled Fast Reactors

    Directory of Open Access Journals (Sweden)

    Jian Wang

    2016-01-01

    Full Text Available Thermally grown oxide (TGO, commonly pure α-Al2O3, formed on protective coatings acts as an insulation barrier shielding cooled reactors from high temperatures in nuclear energy systems. Mixed zone (MZ oxide often grows at the interface between the alumina layer and top coat in thermal barrier coatings (TBCs at high temperature dwell times accompanied by the formation of alumina. The newly formed MZ destroys interface integrity and significantly affects the displacement instabilities of TGO. In this work, a finite element model based on material property changes was constructed to investigate the effects of MZ on the displacement instabilities of TGO. MZ formation was simulated by gradually changing the metal material properties into MZ upon thermal cycling. Quantitative data show that MZ formation induces an enormous stress in TGO, resulting in a sharp change of displacement compared to the alumina layer. The displacement instability increases with an increase in the MZ growth rate, growth strain, and thickness. Thus, the formation of a MZ accelerates the failure of TBCs, which is in agreement with previous experimental observations. These results provide data for the understanding of TBC failure mechanisms associated with MZ formation and of how to prolong TBC working life.

  19. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-01

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

  20. History of fast reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kittel, J.H.; Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France))

    1992-01-01

    Most of the first generation of fast reactors that were operated at significant power levels employed solid metal fuels. They were constructed in the United States and United Kingdom in the 1950s and included Experimental Breeder Reactor (EBR)-I and -II operated by Argonne National Laboratory, United States, the Enrico Fermi Reactor operated by the Atomic Power Development Associates, United States and DFR operated by the U.K. Atomic Energy Authority (UKAEA). Their paper tracer pre-development of fast reactor fuel from these early days through the 1980s including ceramic fuels.

  1. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  2. 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.

  3. Direct Energy Conversion for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.; Cooper, J.; Vogt, D.; Chapline, G.; Turchi, P.; Barbee Jr., T.; Farmer, J.

    2000-07-01

    Strategic Computing Initiative (ASCI), should improve the speed and decrease the cost of developing new TEGs. The system concept to be evaluated is shown in Figure 1. Liquid metal is used to transport heat away from the nuclear heat source and to the TEG. Air or liquid (water or a liquid metal) is used to transport heat away from the cold side of the TEG. Typical reactor coolants include sodium or eutectic mixtures of lead-bismuth. These are coolants that have been used to cool fast neutron reactors. Heat from the liquid metal coolant is rejected through the thermal electric materials, thereby producing electrical power directly. The temperature gradient could extend from as high as 1300 K to 300 K, although fast reactor structural materials (including those used to clad the fuel) currently used limit the high temperature to about 825K.

  4. Study on water cooled high conversion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of study on advanced reactors for the future, conceptual design of high conversion water cooled reactors is being studied, aiming at the contribution to nuclear fuel cycle by the LWR technology, since the utilization of LWRs will extend over a long period of time . We are studying on the reactor core concepts for BWR and PWR reactor systems. As for BWR system, three types of reactor cores are investigating for three different design goals; long operation period, high conversion ratio and high applicability for the existing BWR system. In all the cases, we have obtained a fair prospect of a large core concept with a capacity of 1,000 MWe class having negative void reactivity coefficient. This study is a part of JAERI-JAPCO (Japan Atomic Power Company) cooperative studies. Various kinds of conceptual designs will be created until the end of FY 1999. The designs will be checked and reviewed at that time, then experimental studies on the realization of the concepts will start with further design works from FY 2000. (author)

  5. Modification and Validation of ATHLET Code for Sodium-cooled Fast Reactor Application%ATHLET程序的钠冷快堆应用扩展及其验证

    Institute of Scientific and Technical Information of China (English)

    周翀; Klaus Huber; 程旭

    2013-01-01

    System analysis code is important for the global simulation of the sodium-cooled fast reactor (SFR) system as well as transient and accident safety analysis .In this paper ,the best estimate system code ATHLET for light water reactors ,developed by Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) in Germany ,was modified for SFR application .Thermal-dynamic and transport properties as well as heat transfer correlations for sodium were implemented into the ATHLET code .The modified code was then applied to simulate the Phenix reactor in France ,and validation of the code was conducted with the Phenix reactor natural convection test .The calculation results were compared with the test data .The results show that the modified ATHLET code has good applicability in simulating SFR systems .%系统分析程序是对钠冷快堆的冷却剂回路系统进行全局模拟、瞬态及事故安全分析的重要工具。本工作对德国核设施与反应堆安全机构(GRS)开发的轻水堆最佳估算系统程序ATHLET 进行修改,增加了钠的物性公式和传热关系式,将其适用范围扩展到钠冷快堆。为验证修改过的ATHLET程序,对法国凤凰(Phenix )反应堆系统建模,并对其自然对流实验进行模拟,将计算结果与实验数据进行比较。结果显示,ATHLET程序的钠冷快堆应用扩展具有良好的适用性。

  6. COUPLED FAST-THERMAL POWER BREEDER REACTOR

    Science.gov (United States)

    Avery, R.

    1961-07-18

    A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

  7. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    Science.gov (United States)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    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.

  8. Liquid Metal Cooled Reactor for Space Power

    Science.gov (United States)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  9. Modelling Homogeneous Nucleation in Sodium Fast Reactors under BDBA Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, M.; Herranz, L. E.; Kissane, M.

    2014-07-01

    During postulated Beyond Design Basis Accidents (BDBAs) in Sodium-cooled Fast Reactors (SFRs), the contaminated coolant discharge at high temperature into the containment is considered as a potential scenario during the severe accident progression. In this scenario, the vaporization of sodium and its subsequent combustion (oxidation) would result in supersaturated sodium oxide vapours and formation of large quantities of contaminated aerosols by nucleation of these combustion products. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-15

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

  11. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma (Computational Engineering Analysis, Albuquerque, NM); Al Rashdan, Ahmad (Texas A& M University, College Station, TX); Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX); Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  12. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-30

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

  13. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  14. Current status of fast reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, H.H.

    1979-01-01

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

  15. Progress of China Experimental Fast Reactor in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    1 Background Fast reactor is the reactor which realized the chain fission with fast neutron.As an optional type of generation Ⅳ reactor,fast reactor has three characters:1) It can change 238U to 239Pu and raise the uranium resource utilization

  16. Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

    Science.gov (United States)

    Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

    1972-01-01

    A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

  17. Slow clean-up for fast reactor

    Science.gov (United States)

    Banks, Michael

    2008-05-01

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

  18. Application of Hastelloy X in Gas-Cooled Reactor Systems

    DEFF Research Database (Denmark)

    Brinkman, C. R.; Rittenhouse, P. L.; Corwin, W.R.

    1976-01-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data...

  19. CFD Analysis for Flow Behavior Characteristics in the Upper Plenum during low flow/low pressure transients for the Gas Cooled Fast Reactor (GCFR)

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Theron Marshall; Kevan Weaver; Hans Gougar

    2007-05-01

    Gas coolant at low pressure exhibits poor heat transfer characteristics. This is an area of concern for the passive response targeted by the Generation IV GCFR design. For the first 24 hour period, the decay heat removal for the GCFR design is dependent on an actively powered blower, which also would reduce the temperature in the fuel during transients, before depending on the passive operation. Natural circulation cooling initiates when the blower is stopped for the final phase of the decay heat removal, as under forced convection the core decay heat is adequately cooled by the running blower. The ability of the coolant to flow in the reverse direction or having recirculation, when the blowers are off, necessitates more understanding of the flow behavior characteristics in the upper plenum. The work done here focuses primarily on the period after the blower has been turned off, as the core is adequately cooled when the blowers are running, thus there was no need to carry out the analysis for the first 24 hours. In order to understand the plume behavior for the GCFR upper plenum several cases were run, with air, helium and helium-air mixture. For each case, the FLUENT was used to characterize the steady state velocity vectors and corresponding temperature in the upper plenum under passive decay heat removal conditions. This study will provide better insight into the plume interaction in the upper plenum at low flow and low pressure conditions.

  20. Boosted Fast Flux Loop Alternative Cooling Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Glen R. Longhurst; Donna Post Guillen; James R. Parry; Douglas L. Porter; Bruce W. Wallace

    2007-08-01

    The Gas Test Loop (GTL) Project was instituted to develop the means for conducting fast neutron irradiation tests in a domestic radiation facility. It made use of booster fuel to achieve the high neutron flux, a hafnium thermal neutron absorber to attain the high fast-to-thermal flux ratio, a mixed gas temperature control system for maintaining experiment temperatures, and a compressed gas cooling system to remove heat from the experiment capsules and the hafnium thermal neutron absorber. This GTL system was determined to provide a fast (E > 0.1 MeV) flux greater than 1.0E+15 n/cm2-s with a fast-to-thermal flux ratio in the vicinity of 40. However, the estimated system acquisition cost from earlier studies was deemed to be high. That cost was strongly influenced by the compressed gas cooling system for experiment heat removal. Designers were challenged to find a less expensive way to achieve the required cooling. This report documents the results of the investigation leading to an alternatively cooled configuration, referred to now as the Boosted Fast Flux Loop (BFFL). This configuration relies on a composite material comprised of hafnium aluminide (Al3Hf) in an aluminum matrix to transfer heat from the experiment to pressurized water cooling channels while at the same time providing absorption of thermal neutrons. Investigations into the performance this configuration might achieve showed that it should perform at least as well as its gas-cooled predecessor. Physics calculations indicated that the fast neutron flux averaged over the central 40 cm (16 inches) relative to ATR core mid-plane in irradiation spaces would be about 1.04E+15 n/cm2-s. The fast-to-thermal flux ratio would be in excess of 40. Further, the particular configuration of cooling channels was relatively unimportant compared with the total amount of water in the apparatus in determining performance. Thermal analyses conducted on a candidate configuration showed the design of the water coolant and

  1. Creep-fatigue Interaction Research under High Temperature Condition of Fast Reactor Sodium Pipe

    Institute of Scientific and Technical Information of China (English)

    HU; Li-na

    2015-01-01

    The working temperature of the pipe in primary loop cooling system and decay heat remove system of China Experimental Fast Reactor(CEFR)is higher than material creep temperature(427℃).The design life of the reactor is30a.The pipe works under the repeated thermal load and mechanical load at run time.In order to

  2. Safety Analysis of Pb-208 Cooled 800 MWt Modified CANDLE Reactors

    Science.gov (United States)

    Su'ud, Zaki; Widiawati, Nina; Sekimoto, H.; Artoto, A.

    2017-01-01

    Safely analysis of 800MWt Pb-208 cooled fast reactors with natural Uranium as fuel cycle input employing axial-radial combined Modiified CANDLE burnup scheme has been performed. The analysis of unprotected loss of flow(ULOF) and unprotected rod run-out transient overpower (UTOP) are discussed. Some simulations for 800 MWt Pb-208 cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations. Compared to the Pb-nat cooled long life Fast Reactors, Pb-208 cooled reactors have smaller Doppler but higher coolant density reactivity coefficient. In the UTOP accident case the analysis has been performed against external reactivity up to 0.003dk/k. And for ULOHS case it is assumed that the secondary cooling system has broken. During all accident the cladding temperature is the most critical. Especially for the case of UTOP accident. In addition the steam generator design has also consider excess power which may reach 50% extra during severe UTOP case..

  3. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  4. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  5. Optimization for Fast Zone Multilayer Fuel Assembly of Mixed Supercritical Water-Cooled Reactor%混合能谱超临界水堆快谱组件优化设计

    Institute of Scientific and Technical Information of China (English)

    杨婷; 刘晓晶; 程旭

    2011-01-01

    In order to improve the safety and sustainability of a supercritical water-cooled reactor (SCWR) core, both sub-channel and MCNP analysis were carried out to assess thermal-hydraulic and neutronic performances of the fuel assembly, which was proposed for the fast zone of a mixed-spectrum SCWR (SCWR-M). This fast zone assembly had a multilayer structure and was axially divided into several seed and blanket regions. The effects of some design parameters, I. E. Axial configuration, fuel rod diameter, pitch to diameter ratio and duct wall clearance on the thermal-hydraulic and neutronic performance of assemblies were investigated and an optimized parameter ranges were obtained.%本工作从热工水力和中子物理两方面对混合能谱超临界水堆混合谱堆芯的快谱区多层组件进行优化设计.对于轴向以再生区和裂变区交替布置的快谱组件,分别改变其轴向布置方式、燃料芯块直径、栅径比及外围燃料棒距组件盒最小距离,并分析它们对组件热工和物理性能的影响,从而得到较优的参数范围,尽可能提高混合谱超临界水堆的固有安全性和经济性.

  6. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; K. Hamman

    2009-09-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving

  7. Investigation of CO{sub 2} Recovery System Design in Supercritical Carbon Dioxide Power Cycle for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Seok; Jung, Hwa-Young; Ahn, Yoonhan; Cho, Seong Kuk; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    These are mainly possible because the S-CO{sub 2} Brayton cycle has lower compressing work than other Brayton cycles due to its high density and low compressibility near the critical point. These attributes make easier to achieve higher turbine inlet temperature. Furthermore, the coolant chemistry control and component cooling systems are relatively simple for the S-CO{sub 2} cycle unlike the steam Rankine cycle, and therefore the total plant footprint can be greatly reduced further. However, certain amount of leakage flow is inevitable in the rotating turbo-machinery since the S-CO{sub 2} power cycle is a highly pressurized system. A computational model of critical flow in turbo-machinery seal is essential to predict the leakage flow and calculate the required total mass of working fluid in S-CO{sub 2} power system. Before designing a computational model of critical flow in turbo-machinery seal, this paper will identify what the issues are in predicting leakage flow and how these issues can be successfully addressed. Also, suitability of this solution in a large scale S-CO{sub 2} power cycle will be discussed, because this solution is for the small scale. S-CO{sub 2} power cycle has gained interest especially for the SFR application as an alternative to the conventional steam Rankine cycle, since S-CO{sub 2} power cycle can provide better performance and enhance safety. This paper discussed what the problem in leakage flow is and how to deal with this problem at present. High cavity pressure causing instability of gas foil bearing and large windage losses can be reduced by booster pump used to scavenge the gas in the rotor cavity. Also, labyrinth seals can be another good solution to decrease the rotor cavity pressure. Additionally, difference between large and small scale S-CO{sub 2} power cycle in turbo-machinery leakage is addressed. It is shown that optimization of CO{sub 2} recovery system design is more important to large scale S-CO{sub 2} power cycle. For

  8. EBR-2 (Experimental Breeder Reactor-2), IFR (Integral Fast Reactor) prototype testing programs

    Energy Technology Data Exchange (ETDEWEB)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA). EBR-II Div. Argonne National Lab., IL (USA)); Planchon, H.P.; Lambert, J.D.B. (Argonne National Lab., IL (USA))

    1990-01-01

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

  9. Design of an Actinide Burning, Lead-Bismuth Cooled Reactor That Produces Low Cost Electricity

    Energy Technology Data Exchange (ETDEWEB)

    C. Davis; S. Herring; P. MacDonald; K. McCarthy; V. Shah; K. Weaver (INEEL); J. Buongiorno; R. Ballinger; K. Doyoung; M. Driscoll; P. Hejzler; M. Kazimi; N. Todreas (MIT)

    1999-07-01

    The purpose of this project is to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The choice of lead-bismuth for the reactor coolant is an actinide burning fast reactor offers enhanced safety and reliability. The advantages of lead-bismuth over sodium as a coolant are related to the following material characteristics: chemical inertness with air and water; higher atomic number; lower vapor pressure at operating temperatures; and higher boiling temperature. Given the status of the field, it was agreed that the focus of this investigation in the first two years will be on the assessment of approaches to optimize core and plant arrangements in order to provide maximum safety and economic potential in this type of reactor.

  10. 钠冷快堆非能动余热排出系统实时仿真%Real-Time Simulation of PRHRs in Sodium-Cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    张国强; 孙晓龙; 马锐; 夏庚磊

    2016-01-01

    In order to research the dynamic residual heat removal system (PRHRs ) operation characteristic of sodium cooled fast reactor,simulation model of main flow and equipment in the PRHRs were set up,using real-time two-phase multicomponent modeling tools JTopmeret,and through self-programming air heat exchanger,natural circulation module was developed,and will the coupling of thew above two,the PRHRs real-time simulation model was obtained and the running characteristics of the PRHRs simulation were researched.Simulation results show that most of the important parameters of the simulation data and design data of the relative error is within 2%,and the dynamic trend of the standby condition into the accident conditions of the system is consistent with the trend of theory.%为了研究钠冷快堆非能动余热排出系统(PRHRs)的运行特性,使用实时两相多组分建模工具 JTopmeret 搭建了 PRHRs 中主要流路与设备的仿真模型,通过自编程序开发了空气热交换器中空气自然循环模块,将二者耦合获得了 PRHRs 实时仿真模型,并对 PRHRs 的运行特性进行了仿真研究。仿真结果表明,绝大部分重要参数的仿真数据与设计数据的相对误差在2%之内,系统由备用工况进入事故工况的动态趋势与理论趋势一致。

  11. Risk Management for Sodium Fast Reactors.

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  12. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  13. 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.

  14. Selection of sodium coolant for fast reactors in the US, France and Japan

    Energy Technology Data Exchange (ETDEWEB)

    Sakamoto, Yoshihiko, E-mail: sakamoto.yoshihiko@jaea.go.jp [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan); Garnier, Jean-Claude; Rouault, Jacques [CEA, DEN, DER, Centre de Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Grandy, Christopher; Fanning, Thomas; Hill, Robert [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Chikazawa, Yoshitaka; Kotake, Shoji [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Trilateral study was conducted on coolant selection of fast reactor concept. Black-Right-Pointing-Pointer Fast reactor concepts are vital for nuclear fuel cycle sustainability goals. Black-Right-Pointing-Pointer Sodium, gas and lead cooled fast reactors are capable to achieve the goals. Black-Right-Pointing-Pointer Sodium cooled fast reactor is the most matured technology. Black-Right-Pointing-Pointer Gas and lead cooled fast reactor require long term development. - Abstract: The joint paper presents a common view of fast reactor specific missions in the development of nuclear energy and a cross-analysis of merits and demerits of several Fast Reactors concepts studied worldwide and especially in the Generation-IV International Forum (GIF) framework. The paper provides the context for fast reactors development in the United States, France and Japan and focuses on the comparison on Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), and Lead-cooled Fast Reactor (LFR), i.e. the three fast reactor concepts that have the potential to meet the nuclear fuel cycle sustainability goals. The information provided in the article permits the reader to understand each country's objectives to see that not only the objectives searched for but also the technical orientations are converging. The authors underline that SFR technology evaluation relies significantly on the substantial base technology development programs within each country which is without comparison for the other two fast reactor technologies, e.g., SFR technology has already been developed to commercial or near commercial scale in each country whereas the performance of LFR and GFR technology is still uncertain. The main GFR merits are the potential for high temperatures and the easier possibilities for inspections and repairs. The main challenges are the fuel (fabrication, in-pile behavior), materials for high temperatures, and the implementation of

  15. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  16. Comparison and validation of the results of the AZNHEX v.1.0 code with the MCNP code simulating the core of a fast reactor cooled with sodium; Comparacion y validacion de los resultados del codigo AZNHEX v.1.0 con el codigo MCNP simulando el nucleo de un reactor rapido refrigerado con sodio

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L.; Bastida O, G. E. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Esquivel E, J., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)

  17. 钠冷快堆燃料组件热工水力特性数值模拟与分析%Numerical Simulation and Analysis on Thermal-hydraulic Behavior of Fuel Assembly for Sodium-cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    刘洋; 喻宏; 周志伟

    2014-01-01

    The thermal-hydraulic behavior of triangular arranged fuel bundle with wrapped wire spacer of fuel assembly for sodium-cooled fast reactor was investigated by employing CFD code CFX ,and the results were compared with subchannel analysis code SuperEnergy .Fuel bundles composed of 7,19 ,37 and 61 fuel rods were analyzed sepa-rately .The axial velocity ,cross flow mixing effect ,and temperature rise along axial direction for different subchannels of the fuel bundle were discussed ,and the effect of wrapped wire spacer was carefully investigated .The results show that the wrapped wire spacer plays an important role on the cross flow effect and axial velocity distribution as well as the temperature rise in different subchannels .Moreover ,with the increase of fuel rods ,the flow in fuel bundle becomes more complicated ,and the non-uniformity of the axial flow also shows a tendency to enhance .%利用CFD程序CFX ,分别对7、19、37、61根棒组成的三角形排列螺旋绕丝定位的钠冷快堆燃料组件棒束通道进行了热工水力特性的分析研究,并将结果与子通道程序SuperEnergy进行了对比验证。重点考察了棒束通道轴向流动分布、横向流交混效应及子通道轴向温升,分析了定位绕丝的影响。结果表明,绕丝对棒束通道的横向流交混效应、轴向流动分布及子通道温升有着重要影响,且随棒束的增多,通道内的流动趋向复杂化,轴向流动不均匀性有升高趋势。

  18. 加速器驱动的次临界10MW气冷快堆物理方案研究%Physical Scheme Study of 10 MW Accelerator-driven Sub-critical Gas-cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    秦长平; 顾龙; 李金阳

    2013-01-01

    本文采用MCNPX与ORIGEN耦合的COUPLE2.0程序对加速器驱动的次临界气冷快堆进行了物理方案的设计和研究。计算得到了该方案350d燃耗期间的kef、质子效率、加速器束流强度、1500K温度下的多普勒系数、功率峰因子等参数,并研究了该方案的安全特性;通过计算得到了该方案燃耗并分析了该方案的嬗变能力。结果表明,该方案在350d燃耗期间的kef、加速器束流强度、功率峰因子变化较小,在假想事故下仍保持较深的次临界状态,系统安全性能较好;燃耗较浅,嬗变支持比为20.28,具有较好的嬗变效果。%A physical scheme design of a 10MW accelerator-driven sub-critical gas-cooled fast reactor was studied .The coupling program COUPLE2.0 was applied ,which couples MCNPX and ORIGEN .The physical parameters such as kef ,proton efficiency , accelerator current ,power peak factor and Doppler coefficient of 1500K varying with burnup time were obtained ,and the safety feature of the system was analyzed .By further calculation the 350 d burnup of the system was obtained and the transmutation capability was analyzed . It is found that during the 350 d burnup , kef , accelerator current ,power peak factor variations are relatively small .The system has good safety feature with deep sub-critical status in hypothetical accident . The total burnup is relatively shallow , but the system has good transmutation capability with the transmutation support ratio of 20.28 .

  19. Study of Physical Scheme for 10 MW Accelerator-driven Fast-thermal Coupled Gas-cooled Reactor%加速器驱动10 MW快热耦合气冷堆物理方案研究

    Institute of Scientific and Technical Information of China (English)

    李金阳; 顾龙; 秦长平; 王大伟; 刘璐

    2013-01-01

    The accelerator-driven sub-critical system has promising future in transmuta-tion of nuclear spent fuels .A physical design of a 10MW fast-thermal spectrum gas-cooled reactor was studied .The program COUPLE2.0 ,which couples with MCNPX and ORIGEN ,was applied to compute this scheme ,and the continuous energy neutron cross section was obtained by the nuclear data library of ENDF-7 which was modified with five different temperatures :300 ,600 ,900 ,1 500 ,and 2500K .The physical pa-rameters such as kef ,proton efficiency ,delayed neutron fraction and accelerator current varying with the burnup time were obtained .The system during 350 d burnup was obtained and the transmutation capability was analyzed by the further calculation .It is found that during the 350 d burnup ,the variations of kef and accelerator current are relatively small .The system has good transmutation capability with the transmutation support ratio of 24.86 .%加速器驱动的次临界系统(ADS )在实现嬗变核废料方面具有良好的前景。对加速器驱动10 M W次临界快热耦合能谱气冷堆的物理方案进行了设计和研究,利用 MCNPX和ORIGEN耦合的计算程序COUPLE2.0对该方案进行了计算,其中,中子截面采用ENDF-7处理后得到的5个温度300、600、900、1500、2500 K下的连续能量核数据库。得到该方案350 d燃耗期间的 kef 、质子效率、缓发中子份额以及加速器束流强度的变化。进而计算得到了该方案的燃耗信息,并分析了该方案的嬗变能力。结果表明,该方案在350 d燃耗期间的 kef 、加速器束流强度变化较小,嬗变支持比为24.86,具有较好的嬗变效果。

  20. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  1. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-07

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

  2. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    Energy Technology Data Exchange (ETDEWEB)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

    2001-07-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 {approx} 10{sup -V} at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

  3. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  4. Thermally safe operation of a semibatch reactor for liquid-liquid reactions-fast reactions

    NARCIS (Netherlands)

    Steensma, Metske; Westerterp, K.R.

    1991-01-01

    Accumulation of the reactant supplied to a cooled semibatch reactor (SBR) will occur if the mass transfer rate across the interface is insufficient to keep pace with the supply rate. Then, due to a low starting temperature or supercooling, the reaction temperature does not rise fast enough to the de

  5. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  6. 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.

  7. Fabrication of particulate metal fuel for fast burner reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Lee, Sun Yong; Kim, Jong Hwan; Woo, Yoon Myung; Ko, Young Mo; Kim, Ki Hwan; Park, Jong Man; Lee, Chan Bok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    U Zr metallic fuel for sodium cooled fast reactors is now being developed by KAERI as a national R and D program of Korea. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. Therefore, innovative fuel concepts should be developed to address the fabrication challenges pertaining to TRU while maintaining good performances of metallic fuel. Particulate fuel concepts have already been proposed and tested at several experimental fast reactor systems and vipac ceramic fuel of RIAR, Russia is one of the examples. However, much less work has been reported for particulate metallic fuel development. Spherical uranium alloy particles with various diameters can be easily produced by the centrifugal atomization technique developed by KAERI. Using the atomized uranium and uranium zirconium alloy particles, we fabricated various kinds of powder pack, powder compacts and sintered pellets. The microstructures and properties of the powder pack and pellets are presented.

  8. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  9. Study of Natural Convection Passive Cooling System for Nuclear Reactors

    Science.gov (United States)

    Abdillah, Habibi; Saputra, Geby; Novitrian; Permana, Sidik

    2017-07-01

    Fukushima nuclear reactor accident occurred due to the reactor cooling pumps and followed by all emergencies cooling systems could not work. Therefore, the system which has a passive safety system that rely on natural laws such as natural convection passive cooling system. In natural convection, the cooling material can flow due to the different density of the material due to the temperature difference. To analyze such investigation, a simple apparatus was set up and explains the study of natural convection in a vertical closed-loop system. It was set up that, in the closed loop, there is a heater at the bottom which is representing heat source system from the reactor core and cooler at the top which is showing the cooling system performance in room temperature to make a temperature difference for convection process. The study aims to find some loop configurations and some natural convection performances that can produce an optimum flow of cooling process. The study was done and focused on experimental approach and simulation. The obtained results are showing and analyzing in temperature profile data and the speed of coolant flow at some point on the closed-loop system.

  10. Passive safety system of a super fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sutanto, E-mail: sutanto@fuji.waseda.jp [Cooperative Major in Nuclear Energy, Waseda University, Tokyo (Japan); Polytechnic Institute of Nuclear Technology—National Nuclear Energy Agency, Yogyakarta (Indonesia); Oka, Yoshiaki [The University of Tokyo, Tokyo (Japan)

    2015-08-15

    Highlights: • Passive safety system of a Super FR is proposed. • Total loss of feedwater flow and large LOCA are analyzed. • The criteria of MCST and core pressure are satisfied. - Abstract: Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

  11. Safeguards in prototype fast breeder reactor MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Deshimaru, Takehide; Tomura, Katsuji; Okuda, Yosihisa; Iwamoto, Tomonori [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1994-12-31

    MONJU is the prototype fast breeder reactor in Japan designed to have the electricity output of 280 MWe. Power Reactor and Nuclear Fuel Development Corporation (PNC) started its construction in the autumn of 1985 in Tsuruga site. The loading of the core fuel assemblies to the core have been started since October 1993 and the pre-operational test is undergoing. MONJU uses 198 MOX fuel assemblies as core fuel and 172 DU assemblies as blanket fuel. Assemblies loaded in core and stored in the ex-vessel storage tank (EVST) exist in liquid sodium. These Pu containing fuel assemblies, MOX and irradiated DU, are regarded as in the difficult-to-access area, and the flows of fuel assemblies into and out of the area are requested to be verified. The verification of the flows is designed to be made with fuel flow monitors measuring radiations, which can abridge the inspector attendance during the fuel handling. This paper describes the detailed aspects of the fuel transfers in MONJU facility and the verification of them through flow monitors together with the functions of other safeguards equipments. (author).

  12. Sodium fast reactor fuels and materials : research needs.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  13. Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

    Energy Technology Data Exchange (ETDEWEB)

    Philbin, J.S.; Nelson, W.E.; Rosenstroch, B.

    1978-10-01

    The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.

  14. Fast cooling techniques for gravitational wave antennas

    CERN Document Server

    Furtado, S R

    2002-01-01

    The resonant-mass technique for the detection of gravitational waves may involve, in the near future, the cooling of very large masses (about 100 tons) from room temperature (300 K) to extreme cryogenic temperatures (20 mK). To cool these detectors to cryogenic temperatures an exchange gas (helium) is used, and the heat is removed from the antenna to the cold reservoir by thermal conduction and natural convection. With the current technique, cooling times of about 1 month can be obtained for cylindrical bar antennas of 2.5 tons. Should this same technique be used to cool a 100 ton spherical antenna the cooling time would be about 10 months, making the operation of these antennas impracticable. In this paper, we study the above-mentioned cooling technique and others, such as thermal switching and forced convection from room temperature to liquid nitrogen temperature (77 K) using an aluminium truncated icosahedron of 19 kg weight and 25 cm diameter.

  15. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    OpenAIRE

    Vladimir Petrochenko; Georgy Toshinsky

    2012-01-01

    On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral) design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing...

  16. History of fast reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kittel, J.H. (Argonne National Lab., IL (United States)); Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France)); Bagley, K.Q. (AEA Reactor Services, Risley (United Kingdom)); Crittenden, G.C. (AEA Reactor Services, Dounreay (United Kingdom)); Dievoet, J. van (Belgonucleaire, Brussels (Belgium))

    1993-09-01

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

  17. Control rod drive for high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    DengJun-Xian; XuJi-Ming; 等

    1998-01-01

    This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.

  18. A numerical investigation of the sCO{sub 2} recompression cycle off-design behaviour, coupled to a sodium cooled fast reactor, for seasonal variation in the heat sink temperature

    Energy Technology Data Exchange (ETDEWEB)

    Floyd, J., E-mail: jeremy.floyd@cea.fr [CEA, DEN, Département d’Etudes des Réacteurs, Service d’Etudes des Systèmes Innovants, F-13108 Saint Paul Lez Durance (France); Alpy, N., E-mail: nicolas.alpy@cea.fr [CEA, DEN, Département d’Etudes des Réacteurs, Service d’Etudes des Systèmes Innovants, F-13108 Saint Paul Lez Durance (France); Moisseytsev, A., E-mail: amoissey@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Haubensack, D., E-mail: david.haubensack@cea.fr [CEA, DEN, Département d’Etudes des Réacteurs, Service d’Etudes des Systèmes Innovants, F-13108 Saint Paul Lez Durance (France); Rodriguez, G., E-mail: gilles.rodriguez@cea.fr [CEA, DEN, Département de Technologie Nucléaire, F-13108 Saint Paul Lez Durance (France); Sienicki, J., E-mail: sienicki@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Avakian, G., E-mail: gilles.avakian@cea.fr [CEA, DEN, Département d’Etudes des Réacteurs, Service d’Etudes des Systèmes Innovants, F-13108 Saint Paul Lez Durance (France)

    2013-07-15

    Highlights: • Year-round behaviour of the supercritical CO{sub 2} recompression cycle is simulated. • Behaviour of the system was uncertain due to large changes in the fluid properties. • Cycle thermodynamic optimisation and component preliminary designs were performed. • No off design cycle stability issues, compressors operate away from surge region. • Independent speed control of compressors maintains power and cycle efficiency. -- Abstract: Supercritical CO{sub 2} cycles are particularly attractive for Generation IV Sodium-Cooled Fast Reactors (SFRs) as they can be simple and compact, but still offer steam-cycle equivalent efficiency while also removing potential for Na/H{sub 2}O reactions. However, CO{sub 2} thermophysical properties are very sensitive close to the critical point which raises, in particular, questions about the compressor and so cycle off-design behaviour when subject to inevitable temperature increases that result from seasonal variations in the heat sink temperature. This publication reports the numerical investigation of such an issue that has been performed using the Plant Dynamics Code (ANL, USA), the cycle being optimised for the next French SFR, ASTRID (1500 MW{sub th}), as a test-case. On design, the net plant efficiency is 42.2% for a high pressure (25 MPa) turbine with an inlet temperature of 515 °C and considering a cycle low temperature of 35 °C. The off-design cycle behaviour is studied based on preliminary designs for the main components and assuming the use of a fixed heat sink flow rate. First results obtained using a common fixed shaft speed for all turbomachines, without any other active control, show no stability issues and roughly constant density (and volumetric flow rate) at the main compressor inlet for the range of heat sink temperature considered (21–40 °C). This occurs because the new stationary states are found without requiring a significant shift of mass to the higher pressure level, meaning the

  19. 改进型快谱超临界水冷堆增殖特性初步研究%Primary Study on Breeding Property of Improved Supercritical Water Cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    刘紫静; 于涛; 谢金森

    2012-01-01

    In this paper, the core mode of improved supercritical water cooled fast reactor is established. At first, reasonable fuel assembly design is obtained by studying the influences of seed fuel pin diameter and blanket coolant channel diameter to conversion ratio (Cr). Then, viod reactivity coefficient and CR of six different core arrangements are calculated. Finaly, the influences of fuel components to CR and void reactivity coefficient are analysed. The results show that negative void reactivity coefficient can be satisfied and Cr can be increased by reducing Hydrogen to Heavy-metal ratio (H/HM), increasing blanket assembly numbers by proper distribution. Cr is substantially increased and more negative void reactivity coefficient can be met by reducing PuO2 mass ratio in fuel, when PuO2 mass ratio reach 20.8% in MOX fuel and 235U enriched at 0.2% in UO2 fuel have been adopted as seed and blanket assmbly respectively, the sixth core program reaches CR=1.04395 and give negative void reactivity coefficient, which meets the primary requirements for SCFR breeding.%建立改进型快谱超临界水冷堆( SCFR-M)堆芯模型,探讨点火区燃料棒直径和增殖区水棒直径对堆芯转换比的影响,得到合理的燃料组件设计形式.设计计算6种不同堆芯布置下的增殖特性和空泡反应性,分析燃料组分对堆芯转换比和空泡反应性系数的影响.结果表明:减小氢原子数与重金属原子数之比(H/HM),增加堆芯增殖组件数目并采用合理布置可在满足负空泡反应系数的同时提高转换比;降低燃料中PuO2质量分数可以使转换比大幅增加,同时使堆芯的空泡反应性系数有更大负值;当点火组件采用PuO2质量分数为20.8%的MOX燃料,增殖组件采用235U富集度为0.2%的UO2燃料,方案6的设计可以使堆芯的初始转换比达到1.04395,并且空泡反应性系数为负,初步达到快谱超临界水冷堆的增殖要求.

  20. Burnup concept for a long-life fast reactor core using MCNPX.

    Energy Technology Data Exchange (ETDEWEB)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  1. On the Burning of Plutonium Originating from Light Water Reactor Use in a Fast Molten Salt Reactor—A Neutron Physical Study

    OpenAIRE

    Bruno Merk; Dzianis Litskevich

    2015-01-01

    An efficient burning of the plutonium produced during light water reactor (LWR) operation has the potential to significantly improve the sustainability indices of LWR operations. The work offers a comparison of the efficiency of Pu burning in different reactor configurations—a molten salt fast reactor, a LWR with mixed oxide (MOX) fuel, and a sodium cooled fast reactor. The calculations are performed using the HELIOS 2 code. All results are evaluated against the plutonium burning efficiency d...

  2. Research and Implementation of Sodium-Cooled Fast Reactor Engineering SOA Integration Platform Based Upon Information System%基于SOA的钠冷快堆工程信息系统集成平台的研究和实现

    Institute of Scientific and Technical Information of China (English)

    李健

    2014-01-01

    This paper analyzes the implementation of integrated technology and its limitations exist in the current construction of nuclear power construction management system, and for the sodium cooled fast reactor construction management features, proposed to build SOA-based cold sodium fast reactor engineering information system integration platform (hereinafter referred to as "integration Platform") design ideas. System integration platform for business functions, and system integration theory, the concepts of SOA, analysis and design in-depth study on the basis of the design of the overall structure of the sodium cooled fast reactor engineering information system integration platform, and discusses in detail the integrated platform-specific implementation process.%本文主要分析了当前核电建造施工管理系统集成的实现技术及其存在的局限性,并针对钠冷快堆的施工管理特点,提出了构建基于SOA的钠冷快堆工程信息系统集成平台(以下简称“集成平台”)的设计思想。对集成平台的业务功能体系、系统集成的理论、SOA的相关概念、分析与设计进行了深入研究,在此基础上设计了钠冷快堆工程信息系统集成平台的总体架构,并详细论述了集成平台的具体实现过程。

  3. Advanced gas cooled nuclear reactor materials evaluation and development program

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-01

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

  5. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  6. Neutron spectrometer for fast nuclear reactors

    CERN Document Server

    Osipenko, M; Ricco, G; Caiffi, B; Pompili, F; Pillon, M; Angelone, M; Verona-Rinati, G; Cardarelli, R; Mila, G; Argiro, S

    2015-01-01

    In this paper we describe the development and first tests of a neutron spectrometer designed for high flux environments, such as the ones found in fast nuclear reactors. The spectrometer is based on the conversion of neutrons impinging on $^6$Li into $\\alpha$ and $t$ whose total energy comprises the initial neutron energy and the reaction $Q$-value. The $^6$LiF layer is sandwiched between two CVD diamond detectors, which measure the two reaction products in coincidence. The spectrometer was calibrated at two neutron energies in well known thermal and 3 MeV neutron fluxes. The measured neutron detection efficiency varies from 4.2$\\times 10^{-4}$ to 3.5$\\times 10^{-8}$ for thermal and 3 MeV neutrons, respectively. These values are in agreement with Geant4 simulations and close to simple estimates based on the knowledge of the $^6$Li(n,$\\alpha$)$t$ cross section. The energy resolution of the spectrometer was found to be better than 100 keV when using 5 m cables between the detector and the preamplifiers.

  7. Immobilization of Fast Reactor First Cycle Raffinate

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-02-26

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

  8. System Study: Reactor Core Isolation Cooling 1998–2012

    Energy Technology Data Exchange (ETDEWEB)

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  9. System Study: Reactor Core Isolation Cooling 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  10. System Study: Reactor Core Isolation Cooling 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  12. Mechatronics of fuel handling mechanism for fast experimental reactor 'Joyo'

    Energy Technology Data Exchange (ETDEWEB)

    Fujiwara, Akikazu (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

    1984-01-01

    The outline of the fast experimental reactor ''Joyo'' is introduced, and the fuel handling mechanism peculiar to fast reactors is described. The objectives of the construction of Joyo are to obtain the techniques for the design, construction, manufacture, installation, operation and maintenance of sodium-cooled fast reactors independently, and to use it as an irradiation facility for the development of fuel and materials for fast breeder reactors. At present, the reactor is operated at 100 MW maximum thermal output for the second objective. Since liquid sodium is used as the coolant, the atmosphere of the fuel handling course changes such as liquid sodium at 250 deg C, argon gas at 200 deg C and water, in addition, the spent fuel taken out has the decay heat of 2.1 kW at maximum. The fuel handling works in the reactor and fuel transfer works, and the fuel handling mechanism of a fuel exchanger and that of a cask car for fuel handling are described. Relay sequence control system is used for the fuel handling mechanism of Joyo.

  13. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  14. Assessing reactor physics codes capabilities to simulate fast reactors on the example of the BN-600 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.

  15. A gas-cooled reactor surface power system

    Science.gov (United States)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  16. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    Science.gov (United States)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  17. Investigation of decladding via oxidation for MOX fast reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Westphal, B. R.; Wahlquist, D. L.; Sell, D. A.; Bateman, K. J.; Herrmann, S. D. [Idaho National Laboratory, Boise (United States)

    2008-08-15

    Although the oxidation of spent uranium oxide fuels has been extensively studied for its decladding and off-gassing capabilities, research on mixed oxide (MOX) fuels has not been as rigorous. A few studies have been conducted on the oxidation of MOX fuels for both thermal and fast reactor systems where the plutonium content of the MOX reflects the reactor system; generally less than 10 wt. % for thermal and more than 10 wt. % for fast. For the fast reactor fuel studies, conditions were applied during the oxidation testing of these MOX fuels that were uncharacteristic. In one case a cladding material under early development was tested and in the other a non-irradiated simulant was employed. Thus, irradiated fast reactor MOX fuel has been investigated for decladding by oxidation (DEOX) which utilizes later generation cladding material, viz. D9, an austenitic stainless steel alloy stabilized with titanium.

  18. Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

    Science.gov (United States)

    Koltashev, D. A.; Stakhanova, A. A.

    2017-01-01

    This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

  19. Minimizing the fissile inventory of the molten salt fast reactor

    OpenAIRE

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

    2009-01-01

    International audience; Molten salt reactors in the configurations presented here, called Molten Salt Fast Reactors (MSFR), have been selected for further studies by the Generation IV International Forum. These reactors may be operated in simplified and safe conditions in the Th/233U fuel cycle with fluoride salts. We present here the concept, before focusing on a possible optimization in term of minimization of the initial fissile inventory. Our studies demonstrate that an inventory of 233U ...

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

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

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

  1. Improve Design of Fuel Shear for Fast Reactor

    Institute of Scientific and Technical Information of China (English)

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

    2012-01-01

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

  2. Experimental investigation of a directionally enhanced DHX concept for high temperature Direct Reactor Auxiliary Cooling Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, Joel T.; Blandford, Edward D., E-mail: edb@unm.edu

    2016-07-15

    Highlights: • A novel directional heat exchanger design has been developed. • Hydrodynamic tests have been performed on the proposed design. • Heat transfer performance is inferred by hydrodynamic results. • Results are discussed and future work is suggested. - Abstract: The use of Direct Reactor Auxiliary Cooling Systems (DRACSs) as a safety-related decay heat removal system for advanced reactors has developed historically through the Sodium Fast Reactor (SFR) community. Beginning with the EBR-II, DRACSs have been utilized in a large number of past and current SFR designs. More recently, the DRACS has been adopted for Fluoride Salt-Cooled High-Temperature Reactors (FHRs) for similar decay heat removal functions. In this paper we introduce a novel directionally enhanced DRACS Heat Exchanger (DHX) concept. We present design options for optimizing such a heat exchanger so that shell-side heat transfer is enhanced in one primary coolant flow direction and degraded in the opposite coolant flow direction. A reduced-scale experiment investigating the hydrodynamics of a directionally enhanced DHX was built and the data collected is presented. The concept of thermal diodicity is expanded to heat exchanger technologies and used as performance criteria for evaluating design options. A heat exchanger that can perform as such would be advantageous for use in advanced reactor concepts where primary coolant flow reversal is expected during Loss-of-Forced-Circulation (LOFC) accidents where the ability to circulate coolant is compromised. The design could also find potential use in certain advanced Sodium Fast Reactor (SFR) designs utilizing fluidic diode concepts.

  3. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  4. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  5. What availability rate for a new fast sodium reactor?; Quel taux de disponibilite pour un nouveau reacteur rapide sodium?

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif sur Yvette (France)

    2009-09-15

    This article points out that 18 sodium reactors have operated in the world, prototypes to nuclear power reactors, accumulating 388 years of operation. If one discounts the prototype, only three reactors had a significant and electric power generation suitable for the analysis of availability. An analysis of availability rates for Phoenix and Superphenix is made. A comparison of availability rates of BN 600 reactor and Tricastin 1 reactor (both started in 1980) is also performed. We conclude that, since the R.E.X. (return of operating experience) of previous reactors is taken into account (mainly in material) and lack of political disturbance, can be expected for a sodium cooled fast reactor availability rates comparable to those of other existing reactors. (N.C.)

  6. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  8. Fast Thorium Molten Salt Reactors started with Plutonium

    OpenAIRE

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Mathieu, L.; Brissot, R.; Liatard, E.; Méplan, O.; Nuttin, A.

    2006-01-01

    One of the pending questions concerning Molten Salt Reactors based on the 232Th/233U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233U are examined here: dire...

  9. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    Science.gov (United States)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  10. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-30

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

  12. Fast cooling of trapped ions using the dynamical Stark shift

    Energy Technology Data Exchange (ETDEWEB)

    Retzker, A [Institute for Mathematical Sciences, Imperial College London, SW7 2PE (United Kingdom); Plenio, M B [Institute for Mathematical Sciences, Imperial College London, SW7 2PE (United Kingdom)

    2007-08-15

    A laser cooling scheme for trapped ions is presented which is based on the fast dynamical Stark shift gate, described in (Jonathan et al 2000 Phys. Rev. A 62 042307). Since this cooling method does not contain an off resonant carrier transition, low final temperatures are achieved even in a traveling wave light field. The proposed method may operate in either pulsed or continuous mode and is also suitable for ion traps using microwave addressing in strong magnetic field gradients.

  13. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

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

    2015-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-03-27

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

  15. Specific power of liquid-metal-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.

    1987-10-01

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-03-01

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

  18. Longitudinal dynamics of laser-cooled fast ion beams

    DEFF Research Database (Denmark)

    Weidemüller, M.; Eike, B.; Eisenbarth, U.

    1999-01-01

    We present recent results of our experiments on laser cooling of fast stored ion beams at the Heidelberg Test Storage Ring. The longitudinal motion of the ions is directly cooled by the light pressure force, whereas efficient transverse cooling is obtained indirectly by longitudinal......-transverse coupling mechanisms. Laser cooling in novel bunch forms consisting of square-well buckets leads to longitudinally space-charge dominated beams. The observed longitudinal ion density distributions can be well described by a self-consistent mean-field model based on a thermodynamic Debye-Huckel approach....... When applying laser cooling in square-well buckets over long time intervals, hard Coulomb collisions suddenly disappear and the longitudinal temperature drops by about a factor of three. The observed longitudinal behaviour of the beam shows strong resemblance with the transition to an Coulomb...

  19. A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

    Directory of Open Access Journals (Sweden)

    Hwanyeal Yu

    2015-12-01

    Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

  20. Alternative Fabrication of Recycling Fast Reactor Metal Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki-Hwan; Kim, Jong Hwan; Song, Hoon; Kim, Hyung-Tae; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

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

  1. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  2. High energy resolution and high count rate gamma spectrometry measurement of primary coolant of generation 4 sodium-cooled fast reactor; Spectrometrie gamma haute resolution et hauts taux de comptage sur primaire de reacteur de type generation 4 au sodium liquide

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.

    2010-11-10

    Sodium-cooled Fast Reactors are under development for the fourth generation of nuclear reactor. Breeders reactors could gives solutions for the need of energy and the preservation of uranium resources. An other purpose is the radioactive wastes production reduction by transmutation and the control of non-proliferation using a closed-cycle. These thesis shows safety and profit advantages that could be obtained by a new generation of gamma spectrometry system for SFR. Now, the high count rate abilities, allow us to study new methods of accurate power measurement and fast clad failure detection. Simulations have been done and an experimental test has been performed at the French Phenix SFR of the CEA Marcoule showing promising results for these new measurements. (author) [French] Les reacteurs a neutrons rapides refroidis au sodium sont en developpement en vue d'assurer une quatrieme generation de reacteurs repondant a la demande energetique, tout en assurant la preservation des ressources d'uranium par un fonctionnement en surgenerateur. L'objectif de la filiere est egalement d'ameliorer la gestion de la radiotoxicite des dechets produits par transmutation des actinides mineurs et de controler la non-proliferation par un fonctionnement en cycle ferme. Une instrumentation de surveillance et de controle de ce type de reacteur a ete etudiee dans cette these. La spectrometrie gamma de nouvelle generation permet, par les hauts taux de traitement aujourd'hui accessibles, d'envisager de nouvelles approches pour suivre avec une precision accrue la puissance neutronique et de detecter plus precocement des ruptures de gaine combustible. Des simulations numeriques ont ete realisees et une campagne d'essai a ete menee a bien sur le reacteur Phenix de Marcoule. Des perspectives prometteuses ont ete mises en exergue pour ces deux problematiques

  3. Investigation of V and V process for thermal fatigue issue in a sodium cooled fast reactor – Application of uncertainty quantification scheme in verification and validation with fluid-structure thermal interaction problem in T-junction piping system

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2014-11-15

    Highlights: • Outline of numerical simulation code MUGTHES for fluid-structure thermal interaction was described. • The grid convergence index (GCI) method was applied according to the ASME V and V-20 guide. • Uncertainty of MUGTHES can be successfully quantified for thermal-hydraulic problems and unsteady heat conduction problems in the structure. • Validation for fluid-structure thermal interaction problem in a T-junction piping system was well conducted. - Abstract: Thermal fatigue caused by thermal mixing phenomena is one of the most important issues in design and safety assessment of fast breeder reactors. A numerical simulation code MUGTHES consisting of two calculation modules for unsteady thermal-hydraulics analysis and unsteady heat conduction analysis in structure has been developed to predict thermal mixing phenomena and to estimate thermal response of structure under the thermal interaction between fluid and structure fields. Although verification and validation (V and V) of MUGTHES has been required, actual procedure for uncertainty quantification is not fixed yet. In order to specify an actual procedure of V and V, uncertainty quantifications with the grid convergence index (GCI) estimation according to the existing guidelines were conducted in fundamental laminar flow problems for the thermal-hydraulics analysis module, and also uncertainty for the structure heat conduction analysis module and conjugate heat transfer model was quantified in comparison with the theoretical solutions of unsteady heat conduction problems. After the verification, MUGTHES was validated for a practical fluid-structure thermal interaction problem in T-junction piping system compared with measured results of velocity and temperatures of fluid and structure. Through the numerical simulations in the verification and validation, uncertainty of the code was successfully estimated and applicability of the code to the thermal fatigue issue was confirmed.

  4. Conceptual System Design of a Supercritical CO2 cooled Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong Gu; Oh, Bongseong; Baik, Seung Joon; Yu, Hwanyeal; Kim, Yonghee; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The S-CO2 Brayton cycle has many advantages for SMR's power conversion system. The S-CO2 cycle can achieve small component size and simple cycle layout as shown in Fig. 1. Therefore, a concept of one module containing the S-CO2 cooled fast reactor core and power conversion system is realizable. Thanks to the compact heat exchanger technology such as Printed Circuit Heat Exchanger (PCHE), the supercritical fluid with mediocre heat transfer performance can be utilized to a thermal cycle. This concept of fully modularized reactor is named as KAIST Micro Modular Reactor (MMR). It can achieve large economic by production in series, and transported in the land way or sea way. Based on the design results and dimensions of the reactor core and cycle components, the authors propose a conceptual layout of KAIST MMR. Based on this concept of reactor core, power conversion system, and decay heat removal system, the seasonal operation and transient analysis will be performed in the further works.

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

    Indian Academy of Sciences (India)

    K Velusamy; P Chellapandi; S C Chetal; Baldev Raj

    2010-04-01

    Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multiphysics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

  6. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    Science.gov (United States)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  7. Shape optimization of a Sodium Fast Reactor core

    Directory of Open Access Journals (Sweden)

    Dombre Emmanuel

    2013-01-01

    Full Text Available We apply in this paper a geometrical shape optimization method for the design of the core of a SFR (Sodium-cooled Fast Reactor in order to minimize a thermal counter-reaction known as the sodium void effect. In this kind of reactors, by increasing the temperature, the core may become liable to a strong increase of reactivity, a key-parameter governing the chain-reaction at quasi-static states. We first use the one group energy diffusion model and give the generalization to the two groups energy equation. We then give some numerical results in the case of the one group energy equation. Note that the application of our method leads to some designs whose interfaces can be parametrized by very smooth curves which can stand very far from realistic designs. We don’t explain here the method that it would be possible to use for recovering an operational design but there exists several penalization methods (see [2] that could be employed to this end. On applique dans cet article une méthode d’optimisation géométrique dans le cadre de la conception d’un cœur de réacteur SFR (Sodium-cooled Fast Reactor, i.e. réacteur à neutron rapide refroidi au sodium dans le but de minimiser une contre réaction thermique connue sous le nom d’effet de vidange sodium. Lorsqu’une augmentation de température survient, ce type de réacteur peut être sujet à une forte augmentation de réactivité, un paramètre clé dans le contrôle de la réaction en chaîne en régime quasi-statique. On a recours à l’équation de diffusion à un groupe puis on donne la généralisation du modèle d’optimisation pour l’équation de la diffusion à deux groupes d’énergie. On présente ensuite quelques résultats numériques obtenus dans le cas de l’équation à un groupe d’énergie. On note que l’application de cette méthode conduit à des designs de cœur présentant des interfaces très régulières qui sont loin d’un design de cœur faisable sur le

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  9. Application of Hastelloy X in gas-cooled reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, C.R.; Rittenhouse, P.L.; Corwin, W.R.; Strizak, J.P.; Lystrup, A.; DiStefano, J.R.

    1976-10-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given.

  10. MELCOR Model Development of High Temperature Gas-cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Changyong; Huh, Changwook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    The High Temperature Gas-cooled Reactor is one of the major challenging issues on the development of licensing technology for HTGR. The safety evaluation tools of HTGR can be developed in two ways - development of new HTGR-specific codes or revision of existing codes. The KINS is considering using existing analytic tools to the extent feasible, with appropriate modifications for the intended purpose. The system-level MELCOR code is traditionally used for LWR safety analysis, which is capable of performing thermal-fluid and accident analysis, including fission-product transport and release. Recently, this code is being modified for the NGNP HTGR by the NRC. In this study, the MELCOR input model for HTGR with Reactor Cavity Cooling System (RCCS) was developed and the steady state performance was analyzed to evaluate the applicability in HTGR. HTGR model with design characteristics of GT-MHR was developed using MELCOR 2.1 code to validate the applicability of MELCOR code to HTGR. In addition, the steady state of GT-MHR was analyzed with the developed model. It was evaluated to predict well the design parameters of GT-MHR. The developed model can be used as the basis for accident analysis of HTGR with further update of packages such as Radio Nuclide (RN) package.

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-05-01

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

  13. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, C.; Wachs, D.; Carmack, J.; Woolstenhulme, N.

    2017-01-01

    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

  14. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  15. Description of the magnox type of gas cooled reactor (MAGNOX)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, S.E.; Nonboel, E

    1999-05-01

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO{sub 2}) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  16. Improvement of the decay heat removal characteristics of the generation IV gas-cooled fast reactor; Amelioration des caracteristiques de la dissipation de la chaleur de decroissance pour les reacteurs a neutrons rapides de quatrieme generation refroidi au gaz

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A.S.

    2010-09-07

    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 de-pressurized 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 powered either by the power grid or by batteries for at least 24 hours. The specific contributions of the present research - aimed at achieving enhanced passivity of the DHR system for the GFR - are design and analysis related to (1) the injection of heavy gas into the primary circuit after a LOCA, to enable natural convection cooling at an intermediate-pressure level, and (2) an autonomous Brayton loop to evacuate decay heat at low primary pressure in case of a loss of the guard containment pressure. Both these developments reduce the dependence on blower power availability considerably. First, the thermal-hydraulic codes used in the study - TRACE and CATHARE - are validated for gas cooling. The validation includes benchmark comparisons between the codes, serving to identify the sensitivity of the results to the different modeling assumptions. The parameters found to be the most sensitive in this analysis, such as heat transfer and friction models, are then validated via a

  17. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  18. Simulation Research on Decay Heat Removal System in Primary Loop of Pool-type Sodium-cooled Fast Reactor%池式钠冷快堆事故余热排出系统一回路仿真研究

    Institute of Scientific and Technical Information of China (English)

    姜博; 张智刚; 于洋; 陈广亮; 张志俭

    2015-01-01

    池式钠冷快堆事故余热排出系统采用了非能动工作原理,依靠液态钠及空气的自然对流排出堆芯余热。为研究事故工况下余热排出系统一回路的换热能力,基于 FORTRAN 语言,建立堆芯单通道及盒间流模型,采用全隐二阶迎风差分格式及改进的欧拉法离散求解,对事故余热排出系统一回路系统进行数值模拟,并对全厂断电事故进行仿真计算验证。结果表明:该程序能较好地反映事故余热排出系统瞬态变化过程,并可达到超实时仿真。%T he decay heat removal system in pool‐type sodium‐cooled fast reactor (PSFR) is the passive safety system ,which depends on the natural circulation of sodium and air to keep the reactor coolant cooled .In order to verify the characteristics of the heat transfer of decay heat removal system in primary loop for accident condition ,the core single‐channel model and the flow between fuel assemblies model were established to simulate the decay heat removal system of primary loop and testify the program on station blackout accident , by using fully‐implicit second‐order upwind scheme and ameliorative Eular method to solve the equations based on FORTRAN .The calculation results show that the program could reflect the transient characteristics of the decay heat removal system ,and it could reach excess real‐time simulation .

  19. Steam-Reheat Option for Supercritical-Water-Cooled Reactors

    Science.gov (United States)

    Saltanov, Eugene

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO 2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7--20 kW/m2·K and 9.7--10 kW/m2·K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2 , and MOX may reach melting point.

  20. Development and application of modeling tools for sodium fast reactor inspection

    Science.gov (United States)

    Le Bourdais, Florian; Marchand, Benoît; Baronian, Vahan

    2014-02-01

    To support the development of in-service inspection methods for the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID) project led by the French Atomic Energy Commission (CEA), several tools that allow situations specific to Sodium cooled Fast Reactors (SFR) to be modeled have been implemented in the CIVA software and exploited. This paper details specific applications and results obtained. For instance, a new specular reflection model allows the calculation of complex echoes from scattering structures inside the reactor vessel. EMAT transducer simulation models have been implemented to develop new transducers for sodium visualization and imaging. Guided wave analysis tools have been developed to permit defect detection in the vessel shell. Application examples and comparisons with experimental data are presented.

  1. The fast breeder reactor Rapsodie (1962); Le reacteur rapide surregenerateur rapsodie (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Vautrey, L.; Zaleski, C.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1962-07-01

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors) [French] Dans ce rapport, les auteurs font le point du projet RAPSODIE (reacteur francais surregenerateur a neutrons rapides), au moment du debut effectif de sa construction. On y trouvera decrits: les principales caracteristiques neutroniques et thermiques, le bloc pile et les circuits de refroidissement, les principaux moyens de manutention des ensembles actifs ou contamines, les principes et les moyens qui regissent la conduite du reacteur, les fonctions et l'implantation des divers batiments. La divergence de RAPSODIE est prevue pour 1964. (auteurs)

  2. Conjugate heat transfer analysis of multiple enclosures in prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled reactor under design. The main vessel of the reactor serves as the primary boundary. It is surrounded by a safety vessel which in turn is surrounded by biological shield. The gaps between them are filled with nitrogen. Knowledge of temperature distribution prevailing under various operating conditions is essential for the assessment of structural integrity. Due to the presence of cover gas over sodium free level within the main vessel, there are sharp gradients in temperatures. Also cover gas height reduces during station blackout conditions due to sodium level rise in main vessel caused by temperature rise. This paper describes the model used to analyse the natural convection in nitrogen, conduction in structures and radiation interaction among them. Results obtained from parametric studies for PFBR are also presented.

  3. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    Directory of Open Access Journals (Sweden)

    Vladimir Petrochenko

    2012-09-01

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

  4. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

    Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

  5. Design Study of Small Pb-Bi Cooled Modified Candle Reactors

    Science.gov (United States)

    Su'ud, Zaki; Sekimoto, H.

    2010-06-01

    In this study application of modified CANDLE burnup scheme based long life Pb-Bi Cooled Fast Reactors for small long life reactors with natural Uranium as Fuel Cycle Input has been performed. The reactor cores are subdivided into several parts with the same volume in the axial directions. The natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2, and 10 years after that it is shifted to region 3. This concept is applied to all regions, i.e. shifted the core of I'th region into I+1 region after the end of 10 years burn-up cycle. The first region 1 is filled by fresh natural uranium fuel. Compared to the previous works, in a smaller reactor core the criticality need to be considered more carefully especially at the beginning of life. As an optimized design, a core of 85 cm radius and 150 cm height with 300 MWt power are selected. This core can be operated 10 years without refueling or fuel shuffling. The average discharge burn-up is 350 GWd/ton HM.

  6. Results of FY 2001 feasibility studies on commercialized fast reactor cycle system phase-II

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Hiroshi; Yamashita, Hidetoshi; Maeda, Fumio; Sato, Kazujiro; Ieda, Yoshiaki; Funasaka, Hideyuki [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2002-09-01

    Feasibility Studies on Commercialized Fast Reactor (FR) Cycle System Phase-II were commenced on April 1, 2001, in order to select a few promising candidate concepts for commercialization from the candidate concepts of the FR system and fuel cycle system which were screened in Phase-I, and to present an outline plan for Phase-III onward. In FY 2001, which was the first year of Phase-II, the results of Phase-I and the plan for Phase-II were evaluated as appropriate by The R and D Project Evaluation Committee. With regard to the sodium-cooled medium-scale modular reactor and lead-bismuth cooled modular reactor, economical targets are expected to be achieved. In terms of the gas-cooled reactor, the helium gas-cooled reactor (coated particle fuel type and dispersion fuel type) was screened as a candidate concept. For the reprocessing system, a feasibility of the process for the crystallization method on the advanced aqueous method was confirmed. With regard to the oxide electrowinning method, the technological feasibility of MOX electrowinning co-precipitation was confirmed. In terms of the metal electrowinning method, the possibility of system rationalization was confirmed by Pu recovery testing at liquid Cd cathode. For the fuel fabrication system, in terms of the pelletizing method, the ease of remote-controlled fabrication of low-decontamination TRU fuels was confirmed, and in terms of the vibration compaction method, the packing density is expected to be satisfied as regards the design requirement. With regard to the casting method, the operation parameters of the injection casting technology, which were satisfied to slug specification requirements, were grasped by engineering-scale testing. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1992-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1992-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1975-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-01

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

  11. Integral Fast Reactor Program. Annual progress report, FY 1992

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-06-01

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

  12. Temperature Fluctuation Characteristics Analysis for Steam Generator of Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    ZHU; Li-na; WU; Zhi-guang

    2015-01-01

    In the case of boiling heat transfer deterioration,temperature fluctuating may accelerate the corrosion of heat transfer tubes and can also lead to thermal stress on the tubes.In this paper,dryout-induced temperature fluctuation for the fast reactor steam generator is investigated.The impacts of water flow rate,sodium inlet temperature and the outlet steam

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

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

  14. Physics Characterization of a Heterogeneous Sodium Fast Reactor Transmutation System

    Energy Technology Data Exchange (ETDEWEB)

    Samuel E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even mass number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both a non-flattened and a pancake core geometry. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of the same size.

  15. Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

    Science.gov (United States)

    Nur Krisna, Dwita; Su'ud, Zaki

    2017-01-01

    Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

  16. State of the art of nuclear facilities with organic cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brede, O.; Nagel, S.; Ziegenbein, D.

    1984-06-01

    USA, Canadian, and USSR activities aimed at developing nuclear facilities with organic cooled reactors are summarized. The facilities OMRE, PNPF, WR-1, and ARBUS are described, discussing in particular the problems of the chemistry of organic coolants. Finally, problems of further development and prospects of the application of organic cooled reactors are briefly outlined.

  17. Natural Convection and Boiling for Cooling SRP Reactors During Loss of Circulation Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Buckner, M.R.

    2001-06-26

    This study investigated natural convection and boiling as a means of cooling SRP reactors in the event of a loss of circulation accident. These studies show that single phase natural convection cooling of SRP reactors in shutdown conditions with the present piping geometry is probably not feasible.

  18. Development of Observation Techniques in Reactor Vessel of Experimental Fast Reactor Joyo

    Science.gov (United States)

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

    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 1mm. 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 0.8mm 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

  19. RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Cliff Davis; Goon C. Park

    2007-09-01

    A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls.

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

    CERN Document Server

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

    2011-01-01

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

  1. Study of reactor plant disturbed cooling condition modes caused by the VVER reactor secondary circuit

    Directory of Open Access Journals (Sweden)

    V.I. Belozerov

    2016-12-01

    Based on the RELAP-5, TRAC, and TRACE software codes, reactor plant cooling condition malfunction modes caused by the VVER-1000 secondary circuit were simulated and investigated. Experimental data on the mode with the turbine-generator stop valve closing are presented. The obtained dependences made it possible to determine the maximum values of pressure and temperature in the circulation circuit as well as estimate the Minimum Critical Heat Flux Ratio (MCHFR. It has been found that, if any of the initial events occurs, safety systems are activated according to the set points; transient processes are stabilized in time; and the Critical Heat Flux (CHF limit is provided. Therefore, in the event of emergency associated with the considered modes, the reactor plant safety will be ensured.

  2. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  3. Transitioning nuclear fuel cycles with uncertain fast reactor costs

    Energy Technology Data Exchange (ETDEWEB)

    Phathanapirom, U.B., E-mail: bphathanapirom@utexas.edu; Schneider, E.A.

    2016-06-15

    This paper applies a novel decision making methodology to a case study involving choices leading to the transition from the current once-through light water reactor fuel cycle to one relying on continuous recycle of plutonium and minor actinides in fast reactors in the face of uncertain fast reactor capital costs. Unique to this work is a multi-stage treatment of a range of plausible trajectories for the evolution of fast reactor capital costs over time, characterized by first-of-a-kind penalties as well as time- and unit-based learning. The methodology explicitly incorporates uncertainties in key parameters into the decision-making process by constructing a stochastic model and embedding uncertainties as bifurcations in the decision tree. “Hedging” strategies are found by applying a choice criterion to select courses of action which mitigate “regrets”. These regrets are calculated by evaluating the performance of all possible transition strategies for every feasible outcome of the uncertain parameter. The hedging strategies are those that preserve the most flexibility for adjusting the fuel cycle strategy in response to new information as uncertainties are resolved.

  4. Instrumentation, Monitoring and NDE for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bond, Leonard J.; Doctor, Steven R.; Bunch, Kyle J.; Good, Morris S.; Waltar, Alan E.

    2007-07-28

    The Global Nuclear Energy Partnership (GNEP) has been proposed as a viable system in which to close the fuel cycle in a manner consistent with markedly expanding the global role of nuclear power while significantly reducing proliferation risks. A key part of this system relies on the development of actinide transmutation, which can only be effectively accomplished in a fast-spectrum reactor. The fundamental physics for fast reactors is well established. However, to achieve higher standards of safety and reliability, operate with longer intervals between outages, and achieve high operating capacity factors, new instrumentation and on-line monitoring capabilities will be required--during both fabrication and operation. Since the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor – II (EBR-II) reactors were operational in the USA, there have been major advances in instrumentation, not the least being the move to digital systems. Some specific capabilities have been developed outside the USA, but new or at least re-established capabilities will be required. In many cases the only available information is in reports and papers. New and improved sensors and instrumentation will be required. Advanced instrumentation has been developed for high-temperature/high-flux conditions in some cases, but most of the original researchers and manufacturers are retired or no longer in business.

  5. Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Frisani, Angelo; Hassan, Yassin A; Ugaz, Victor M

    2010-11-02

    The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the

  6. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    G Pandikumar; V Gopalakrishnan; P Mohanakrishnan

    2009-05-01

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

  7. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  8. The Effect of Duct Level on the Performance of Reactor Vault Cooling System in the PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Sujin; Ryu, Seung Ho; Kim, Dehee; Lee, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Development of the prototype gen-Ⅵ sodium-cooled fast reactor (PGSFR) has been ongoing in Korea Atomic Energy Research Institute (KAERI). A reactor vault cooling system (RVCS), one of passive decay heat removal systems (PDHRS), passively removes core decay heat by chimney effect when severe accidents occur. The air cooling path is located around containment vessel (CV). An air separator which divides the downstream air and the upstream air is installed between CV and the concrete wall. To design the RVCS, key design parameters such as stack height, gap size between the concrete wall and the air separator, gap size between the air separator and the CV, thickness and layer composition of the air separator have to be determined. A duct level is one of these design parameters. It denotes the height of the upstream air path and related to the heat transfer length from CV to air. The duct level should be optimized with considering structural reliability and heat removal performance. Thus, in this paper, the heat removal performance of RVCS is evaluated depends on the duct level using 1D system design code, that is developed by KAERI autonomously, and commercial CFD program for optimum design of RVCS In this paper, the heat removal performance of RVCS is evaluated depends on the duct level using PARS2- LMR code and commercial CFD program for optimum design of RVCS to satisfy both conflicting needs, structural reliability and cooling performance. As a result of PARS2-LMR code analysis, it was observed that the heat removal rate increases as increase of duct level and the geometrical conditions, that satisfy the design limitations, were obtained. To qualitatively observe the trends of local temperature distribution, CFD simulations were conducted and hotspots were observed at the upper region of ducts for the low duct level case.

  9. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    Science.gov (United States)

    Malathi, N.; Sahoo, P.; Ananthanarayanan, R.; Murali, N.

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  10. Level monitoring system with pulsating sensor--application to online level monitoring of dashpots in a fast breeder reactor.

    Science.gov (United States)

    Malathi, N; Sahoo, P; Ananthanarayanan, R; Murali, N

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

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

    Energy Technology Data Exchange (ETDEWEB)

    Suk, S.D.; Hahn, D. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2001-07-01

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

  12. High performance infrared fast cooled detectors for missile applications

    Science.gov (United States)

    Reibel, Yann; Espuno, Laurent; Taalat, Rachid; Sultan, Ahmad; Cassaigne, Pierre; Matallah, Noura

    2016-05-01

    SOFRADIR was selected in the late 90's for the production of 320×256 MW detectors for major European missile programs. This experience has established our company as a key player in the field of missile programs. SOFRADIR has since developed a vast portfolio of lightweight, compact and high performance JT-based solutions for missiles. ALTAN is a 384x288 Mid Wave infrared detector with 15μm pixel pitch, and is offered in a miniature ultra-fast Joule- Thomson cooled Dewar. Since Sofradir offers both Indium Antimonide (InSb) and Mercury Cadmium Telluride technologies (MCT), we are able to deliver the detectors best suited to customers' needs. In this paper we are discussing different figures of merit for very compact and innovative JT-cooled detectors and are highlighting the challenges for infrared detection technologies.

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

    CERN Document Server

    Ohki, S

    2003-01-01

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

  14. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  15. Risk-assessment methodology for fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ott, K. O.

    1976-04-01

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

  16. Computational Neutronics Methods and Transmutation Performance Analyses for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    R. Ferrer; M. Asgari; S. Bays; B. Forget

    2007-03-01

    The once-through fuel cycle strategy in the United States for the past six decades has resulted in an accumulation of Light Water Reactor (LWR) Spent Nuclear Fuel (SNF). This SNF contains considerable amounts of transuranic (TRU) elements that limit the volumetric capacity of the current planned repository strategy. A possible way of maximizing the volumetric utilization of the repository is to separate the TRU from the LWR SNF through a process such as UREX+1a, and convert it into fuel for a fast-spectrum Advanced Burner Reactor (ABR). The key advantage in this scenario is the assumption that recycling of TRU in the ABR (through pyroprocessing or some other approach), along with a low capture-to-fission probability in the fast reactor’s high-energy neutron spectrum, can effectively decrease the decay heat and toxicity of the waste being sent to the repository. The decay heat and toxicity reduction can thus minimize the need for multiple repositories. This report summarizes the work performed by the fuel cycle analysis group at the Idaho National Laboratory (INL) to establish the specific technical capability for performing fast reactor fuel cycle analysis and its application to a high-priority ABR concept. The high-priority ABR conceptual design selected is a metallic-fueled, 1000 MWth SuperPRISM (S-PRISM)-based ABR with a conversion ratio of 0.5. Results from the analysis showed excellent agreement with reference values. The independent model was subsequently used to study the effects of excluding curium from the transuranic (TRU) external feed coming from the LWR SNF and recycling the curium produced by the fast reactor itself through pyroprocessing. Current studies to be published this year focus on analyzing the effects of different separation strategies as well as heterogeneous TRU target systems.

  17. LSP simulations of fast ions slowing down in cool magnetized plasma

    Science.gov (United States)

    Evans, Eugene S.; Cohen, Samuel A.

    2015-11-01

    In MFE devices, rapid transport of fusion products, e.g., tritons and alpha particles, from the plasma core into the scrape-off layer (SOL) could perform the dual roles of energy and ash removal. Through these two processes in the SOL, the fast particle slowing-down time will have a major effect on the energy balance of a fusion reactor and its neutron emissions, topics of great importance. In small field-reversed configuration (FRC) devices, the first-orbit trajectories of most fusion products will traverse the SOL, potentially allowing those particles to deposit their energy in the SOL and eventually be exhausted along the open field lines. However, the dynamics of the fast-ion energy loss processes under conditions expected in the FRC SOL, where the Debye length is greater than the electron gyroradius, are not fully understood. What modifications to the classical slowing down rate are necessary? Will instabilities accelerate the energy loss? We use LSP, a 3D PIC code, to examine the effects of SOL plasma parameters (density, temperature and background magnetic field strength) on the slowing down time of fast ions in a cool plasma with parameters similar to those expected in the SOL of small FRC reactors. This work supported by DOE contract DE-AC02-09CH11466.

  18. The effects of spectral shift absorbers on the design and safety of fast spectrum space reactors

    Science.gov (United States)

    King, Jeffrey Charles

    Spectral Shift Absorbers (SSAs) are incorporated into space reactors to maintain them sufficiently subcritical when submerged in seawater or wet sand and subsequently flooded, following a launch abort accident. The effect of four SSAs (samarium-149, europium-151, gadolinium-155, and gadolinium-157) on the submersion criticality, operation, and temperature reactivity feedback of the thermal spectrum reactors developed in the Systems for Nuclear Auxilary Power (SNAP) program is extensively documented. Recent work on SSAs in fast spectrum space reactors, preferred for compactness and higher powers, has focused on rhenium as the primary SSA. In addition to identifying additional SSAs, the present work investigates the effects of SSAs on the overall size and mass, temperature reactivity feedback, and operational lifetime of fast spectrum space reactors. The fast spectrum S4 reactor has a sectored Mo-14%Re solid-core, loadedwith UN fuel, cooled by He-30%Xe, and designed to avoid single point failures at a steady thermal power of 550 kWth. The addition of SSAs to the reactor core increases the fuel enrichment and decreases the size and mass of the reactor and the radiation shadow shield. SSA additions of boron-10, europium-151, gadolinium-155 and iridium result in the smallest and lightest S4 reactors. The effects of SSA additions on the operational lifetime and the temperature and burnup reactivity coefficients of the S^4 reactor are studied. An increasein fuel enrichment with SSAs markedly increases the operational lifetime by decreasing the burnup reactivity coefficient with only a slight decrease in the temperature reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt%), the temperature and burnup reactivity coefficients are the highest (-0.2709 ¢/K and -1.3470 /atom%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to -0.2649 ¢/K and -1.0230 /atom%, and

  19. Preliminary safety calculations to improve the design of Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brovchenko, M.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Capellan, N.; Ghetta, V.; Laureau, A. [LPSC, CNRS/IN2P3, Grenoble INP, 53,rue des Martyrs, 38026 Grenoble Cedex (France)

    2012-07-01

    Molten salt reactors are liquid fuel reactors so that they are flexible in operation but very different in the safety approach from solid fuel reactors. This study bears on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the R and D studies. This paper presents the first step of the safety approach: the systematic description of the MSFR, limited here to the main systems surrounding the core. This systematic description is the basis on which we will be able to devise accidental scenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarios lead to reactor shut down. Because of the decay heat generated in the fuel salt, it must be cooled. After the description of the tools developed to calculate the residual heat, the different contributions are discussed in this study. The decay heat of fission products in the MSFR is evaluated to be low (3% of nominal power), mainly due to the reprocessing that transfers the fission products to the gas reprocessing unit. As a result, the contribution of the actinides is significant (0.5% of nominal power). The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable (> 1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system. (authors)

  20. Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

    Energy Technology Data Exchange (ETDEWEB)

    LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.

    1980-08-01

    The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets).

  1. Experimental investigation of sodium boiling heat exchange in fuel subassembly mockup for perspective fast reactor safety substantiation

    Directory of Open Access Journals (Sweden)

    R.R. Khafizov

    2015-10-01

    Full Text Available Numerical modeling of ULOF-type accident development in sodium-cooled fast reactor carried out using the COREMELT code indicate the development and spreading of sodium boiling in the core accompanied with fluctuations of reactor technological parameters lasting over a period of several tens of a seconds. Significant influence on the calculation results is produced by two-phase coolant flow regime so the code boiling models requiring experimental confirmation. Design solution that includes the “sodium cavity” above the reactor core was suggested in order to exclude reactor accidents resulting in the destruction of reactor core elements. As the result of experimental studies on heat exchange during sodium boiling in the fast reactor fuel subassembly mockup with “sodium cavity” conducted on the AR-1 test facility under natural circulation conditions it was demonstrated possibility of long-term fuel pins simulators stable cooling. Schematic map of two-phase liquid metal flow regimes in fuel pin bundles is presented, data on the heat transfer during liquid metal coolant boiling in the fuel assembly are presented and analyzed. The obtained experimental data are used for further elaboration of the calculation model of sodium boiling in the fuel assembly and for COREMELT computer code verification.

  2. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  3. GRS Method for Uncertainties Evaluation of Parameters in a Prospective Fast Reactor

    Science.gov (United States)

    Peregudov, A.; Andrianova, O.; Raskach, K.; Tsibulya, A.

    2014-04-01

    A number of recent studies have been devoted to the uncertainty estimation of reactor calculation parameters by the GRS (Generation Random Sampled) method. This method is based on direct sampling input data resulting in formation of random sets of input parameters which are used for multiple calculations. Once these calculations are performed, statistical processing of the calculation results is carried out to determine the mean value and the variance of each calculation parameter of interest. In our study this method is used to estimate the uncertainty of calculation parameters (keff, power density, dose rate) of a prospective sodium-cooled fast reactor. Neutron transport calculations were performed by the nodal diffusion code TRIGEX and Monte Carlo code MMK.

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

    Indian Academy of Sciences (India)

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

    2002-10-01

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

  5. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

    Directory of Open Access Journals (Sweden)

    Zhu Guifeng

    2016-01-01

    Full Text Available Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2 salt Temperature Reactivity Coefficient (TRC. Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern and two kinds of reflector materials (SiC and graphite. This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9Be(n,2n reaction and low neutron absorption of 6Li (even at 1000 ppm in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.

  6. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  7. High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

    Energy Technology Data Exchange (ETDEWEB)

    Pudelek, R. E.; Gilbert, W. C.

    2002-02-26

    This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste

  8. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    OpenAIRE

    Galvez, Cristhian

    2011-01-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the pa...

  9. Technical Progress of 600 MW Demonstration Fast Reactor(CFR600)

    Institute of Scientific and Technical Information of China (English)

    YANG; Hong-yi; LIU; Yi-zhe; YANG; Yong; LIU; Zhao-yang; LI; Hai-sheng; WU; Qiang; SUN; Xiao-fu; YANG; Xiao-yan; MA; Jian-ming; LIU; Chen; GUO; Ming-liang

    2015-01-01

    In the year 2015,600 MW Demonstration Fast Reactor(CFR600)is the key technology research and development project in CNNC,the staged achievements have been obtained by Department of Reactor Engineering Technology(Fast Reactor Research and Design),after the great quantity work for the main system.During the whole work,the

  10. Fast Pyrolysis of Lignin Using a Pyrolysis Centrifuge Reactor

    DEFF Research Database (Denmark)

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

    2013-01-01

    Fast pyrolysis of lignin from an ethanol plant was investigated on a lab scale pyrolysis centrifuge reactor (PCR) with respect to pyrolysis temperature, reactor gas residence time, and feed rate. A maximal organic oil yield of 34 wt % dry basis (db) (bio-oil yield of 43 wt % db) is obtained...... at temperatures of 500−550 °C, reactor gas residence time of 0.8 s, and feed rate of 5.6 g/min. Gas chromatography mass spectrometry and size-exclusion chromatography were used to characterize the Chemical properties of the lignin oils. Acetic acid, levoglucosan, guaiacol, syringols, and p-vinylguaiacol are found...... to be major chemical components in the lignin oil. The maximal yields of 0.62, 0.67, and 0.38 wt % db were obtained for syringol, p-vinylguaiacol, and guaiacol, respectively. The reactor temperature effect was investigated in a range of 450−600 °C and has a considerable effect on the observed chemical...

  11. Behavior of actinides in the Integral Fast Reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    1994-06-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  12. Model biases in high-burnup fast reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)

    2012-07-01

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  13. Behavior of actinides in the Integral Fast Reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    1994-06-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  14. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    Energy Technology Data Exchange (ETDEWEB)

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  15. Numerical studies of fast ion slowing down rates in cool magnetized plasma using LSP

    Science.gov (United States)

    Evans, Eugene S.; Kolmes, Elijah; Cohen, Samuel A.; Rognlien, Tom; Cohen, Bruce; Meier, Eric; Welch, Dale R.

    2016-10-01

    In MFE devices, rapid transport of fusion products from the core into the scrape-off layer (SOL) could perform the dual roles of energy and ash removal. The first-orbit trajectories of most fusion products from small field-reversed configuration (FRC) devices will traverse the SOL, allowing those particles to deposit their energy in the SOL and be exhausted along the open field lines. Thus, the fast ion slowing-down time should affect the energy balance of an FRC reactor and its neutron emissions. However, the dynamics of fast ion energy loss processes under the conditions expected in the FRC SOL (with ρe fast ions in a cool plasma. As we use explicit algorithms, these simulations must spatially resolve both ρe and λDe, as well as temporally resolve both Ωe and ωpe, increasing computation time. Scaling studies of the fast ion charge (Z) and background plasma density are in good agreement with unmagnetized slowing down theory. Notably, Z-scaling represents a viable way to dramatically reduce the required CPU time for each simulation. This work was supported, in part, by DOE Contract Number DE-AC02-09CH11466.

  16. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2016-01-01

    Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

  17. Safeguards in the prototype fast breeder reactor MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Usami, S.; Deshimaru, T.; Tomura, K. [Power Reactor and Nuclear Fuels Development Corporation, Ibaraki-ken (Japan)

    1995-12-31

    MONJU is a prototype fast breeder reactor in Japan designed to have a 280-MW(electric) output. The Power Reactor and Nuclear Fuel Development Corporation (PNC) started its construction in the autumn of 1985 in Tsuruga. The loading of the core fuel assemblies was started in October 1993, and the preoperational test is ongoing. MONJU uses 198 mixed-oxide (MOX) fuel assemblies as core fuel and 172 depleted uranium assemblies as blanket fuel. Assemblies loaded in-core and stored in the ex-vessel storage tank (EVST) reside in liquid sodium. These plutonium-containing fuel assemblies, MOX, and irradiated depleted uranium are regarded as in the difficult-to-access area, and the flows of fuel assemblies into and out of the area must be verified. Flow is verified by fuel flow monitors measuring radiation, which can limit inspector attendance during fuel handling.

  18. Transient analyses for a molten salt fast reactor with optimized core geometry

    Energy Technology Data Exchange (ETDEWEB)

    Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

    2015-10-15

    Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

  19. A Study of Reactor Neutrino Monitoring at Experimental Fast Reactor JOYO

    CERN Document Server

    Furuta, H; Hara, T; Haruna, T; Ishihara, N; Ishitsuka, M; Ito, C; Katsumata, M; Kawasaki, T; Konno, T; Kuze, M; Maeda, J; Matsubara, T; Miyata, H; Nagasaka, Y; Nitta, K; Sakamoto, Y; Suekane, F; Sumiyoshi, T; Tabata, H; Takamatsu, M; Tamura, N

    2011-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3m from the JOYO reactor core of 140MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11\\pm1.24(stat.)\\pm0.46(syst.)events/day. Although the statistical significance of the measurement was not enough, the background in such a compact detector at the ground level was studied in detail and MC simulation was found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

  20. Experimental Progress in Fast Cooling in the ESR

    CERN Document Server

    Steck, Markus; Beller, Peter; Franzke, Bernhard; Nolden, Fritz

    2005-01-01

    The ESR storage ring at GSI is operated with highly charged heavy ions. Due to the high electric charge the ions interact much stronger with electromagnetic fields. Therefore both cooling methods which are applied to stored ions in the ESR, stochastic cooling and electron cooling, are more powerful than for singly charged particles. The experimental results exhibit cooling times for stochastic cooling of a few seconds. For cold ion beams, electron cooling provides cooling times which are one to two orders of magnitude smaller. The beams are cooled to beam parameters which are limited by intrabeam scattering. At small ion numbers, however, intrabeam scattering is suppressed by electron cooling, clear evidence was found that the ion beam forms a one-dimensional ordered structure, a linear chain of ions. The strengths of stochastic cooling and electron cooling are complementary and can be combined favorably. Stochastic cooling is employed for pre-cooling of hot secondary beams followed by electron cooling to pro...

  1. Final report-passive safety optimization in liquid sodium-cooled reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Cahalana, J. E.; Hahn, D.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2007-08-13

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  2. Example Work Domain Analysis for a Reference Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    The nuclear industry is currently designing and building a new generation of reactors that will include different structural, functional, and environmental aspects, all of which are likely to have a significant impact on the way these plants are operated. In order to meet economic and safety objectives, these new reactors will all use advanced technologies to some extent, including new materials and advanced digital instrumentation and control systems. New technologies will affect not only operational strategies, but will also require a new approach to how functions are allocated to humans or machines to ensure optimal performance. Uncertainty about the effect of large scale changes in plant design will remain until sound technical bases are developed for new operational concepts and strategies. Up-to-date models and guidance are required for the development of operational concepts for complex socio-technical systems. This report describes how the classical Work Domain Analysis method was adapted to develop operational concept frameworks for new plants. This adaptation of the method is better able to deal with the uncertainty and incomplete information typical of first-of-a-kind designs. Practical examples are provided of the systematic application of the method in the operational analysis of sodium-cooled reactors. Insights from this application and its utility are reviewed and arguments for the formal adoption of Work Domain Analysis as a value-added part of the Systems Engineering process are presented.

  3. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  4. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  5. Fast reactor core concepts to improve transmutation efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Fujimura, Koji; Kawashima, Katsuyuki [Hitachi Research Laboratory, Hitachi, Ltd., 7-1-1, Omika-cho, Hitachi-shi, Ibaraki, 319-1221 Japan (Japan); Itooka, Satoshi [Hitachi-GE Nuclear Energy, Ltd., 3-1-1, Saiwai-cho, Hitachi-shi, Ibaraki, 317-0073 Japan (Japan)

    2015-12-31

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

  6. Evaluation of the breed/burn fast reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Atefi, B.; Driscoll, M.J.; Lanning, D.D.

    1979-12-01

    A core design concept and fuel management strategy, designated breed/burn, has been evaluated for heterogeneous fast breeder reactors. In this concept internal blanket assemblies after fissile material is bred in over several incore cycles, are shuffled into a moderated radial blanket and/or central island. The most promising materials combination identified used thorium in the internal blankets (due to the superior performance of epithermal Th-U233 systems) and zirconium hydride (ZrH/sub 16/) as the moderator (because of the compact assembly and core designs it permitted).

  7. Accelerated Irradiations for High Dose Microstructures in Fast Reactor Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhijie [Univ. of Michigan, Ann Arbor, MI (United States)

    2017-03-31

    The objective of this project is to determine the extent to which high dose rate, self-ion irradiation can be used as an accelerated irradiation tool to understand microstructure evolution at high doses and temperatures relevant to advanced fast reactors. We will accomplish the goal by evaluating phase stability and swelling of F-M alloys relevant to SFR systems at very high dose by combining experiment and modeling in an effort to obtain a quantitative description of the processes at high and low damage rates.

  8. Study of guided wave transmission through complex junction in sodium cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Elie, Q.; Le Bourdais, F.; Jezzine, K.; Baronian, V. [Non Destructive Testing Department at the French Atomic Energy Commission (CEA), Saclay, 91191 Gif sur Yvette CEDEX, (France)

    2015-07-01

    Ultrasonic guided wave techniques are seen as suitable candidates for the inspection of welded structures within sodium cooled fast reactors (SFR), as the long range propagation of guided waves without amplitude attenuation can overcome the accessibility problem due to the liquid sodium. In the context of the development of the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID), the French Atomic Commission (CEA) investigates non-destructive testing techniques based on guided wave propagation. In this work, guided wave NDT methods are applied to control the integrity of welds located in a junction-type structure welded to the main vessel. The method presented in this paper is based on the analysis of scattering matrices peculiar to each expected defect, and takes advantage of the multi-modal and dispersive characteristics of guided wave generation. In a simulation study, an algorithm developed using the CIVA software is presented. It permits selecting appropriate incident modes to optimize detection and identification of expected flawed configurations. In the second part of this paper, experimental results corresponding to a first validation step of the simulation results are presented. The goal of the experiments is to estimate the effectiveness of the incident mode selection in plates. The results show good agreement between experience and simulation. (authors)

  9. Steady Thermal Field Simulation of Forced Air-cooled Column-type Air-core Reactor

    Institute of Scientific and Technical Information of China (English)

    DENG Qiu; LI Zhenbiao; YIN Xiaogen; YUAN Zhao

    2013-01-01

    Modeling the steady thermal field of the column-type air-core reactor,and further analyzing its distribution regularity,will help optimizing reactor design as well as improving its quality.The operation mechanism and inner insulation structure of a novel current limiting column-type air-core reactor is introduced in this paper.The finite element model of five encapsulation forced air-cooled column type air-core reactor is constructed using Fluent.Most importantly,this paper present a new method that,the steady thermal field of reactor working under forced air-cooled condition is simulated without arbitrarily defining the convection heat transfer coefficient for the initial condition; The result of the thermal field distribution shows that,the maximum steady temperature rise of forced air-cooled columntype air-core reactor happens approximately 5% to its top.The law of temperature distribution indicates:In the 1/3part of the reactor to its bottom,the temperature will rise rapidly to the increasing of height,yet the gradient rate is gradually decreasing; In the 5 % part of the reactor to its top,the temperature will drop rapidly to the increasing of height; In the part between,the temperature will rise slowly to the increasing of height.The conclusion draws that more thermal withstand capacity should be considered at the 5 % part of the reactor to its top to achieve optimal design solution.

  10. An autonomous long-term fast reactor system and the principal design limitations of the concept

    Science.gov (United States)

    Tsvetkova, Galina Valeryevna

    The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National

  11. Fast Traveling-Wave Reactor of the Channel Type

    CERN Document Server

    Rusov, Vitaliy D; Vashchenko, Volodymyr N; Chernezhenko, Sergei A; Kakaev, Andrei A; Pantak, Oksana I

    2015-01-01

    The main aim of this paper is to solve the technological problems of the TWR based on the technical concept described in our priority of invention reference, which makes it impossible, in particular, for the fuel claddings damaging doses of fast neutrons to excess the ~200 dpa limit. Thus the essence of the technical concept is to provide a given neutron flux at the fuel claddings by setting the appropriate speed of the fuel motion relative to the nuclear burning wave. The basic design of the fast uranium-plutonium nuclear traveling-wave reactor with a softened neutron spectrum is developed, which solves the problem of the radiation resistance of the fuel claddings material.

  12. Time dependence of corrosion in steels for use in lead-alloy cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Machut, McLean [Department of Nuclear Engineering and Engineering Physics, University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)], E-mail: mtmachut@wisc.edu; Sridharan, Kumar [Department of Nuclear Engineering and Engineering Physics, University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States); Li Ning [Materials Physics and Application Division, AFCI, Los Alamos National Laboratory, NM (United States); Ukai, Shigeharu [Division of Materials Science and Engineering, Hokaido University (Japan); Allen, Todd [Department of Nuclear Engineering and Engineering Physics, University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)

    2007-09-15

    Stability of the protective oxide layer is critical for the long-term performance of cladding and structural components in lead-alloy cooled nuclear systems. Measurements have shown that removal of the outer magnetite layer is a significant effect at higher temperatures in flowing lead-bismuth. Developing a predictive capability for oxide thickness and material removal is therefore needed. A model for the corrosion of steels in liquid lead-alloys has been employed to assist in materials development for application in the Generation IV Lead-cooled Fast Reactor (LFR). Data from corrosion tests of steels in Los Alamos National Laboratory's DELTA Loop is used to benchmark the model and to obtain predictions of long-term material's corrosion performance. The model is based on modifications of Wagner's diffusion based oxidation theory and Tedmon's equation for high-temperature oxidation with scale removal. Theoretically and experimentally obtained values for parabolic oxide growth rate, mass transfer corrosion rate, and long-term material thinning rates are presented and compared to the literature.

  13. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  14. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  15. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff considers acceptable for demonstrating the operability of emergency core cooling systems (ECCSs) for boiling...

  16. Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems: Annual and Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Kevan D. Weaver; Theron Marshall; James Parry

    2005-10-01

    The Idaho National Laboratory (INL) contribution to the Nuclear Energy Research Initiative (NERI) project number 2002-005 was divided into reactor physics, and thermal-hydraulics and plant design. The research targeted credible physics and thermal-hydraulics models for a gas-cooled fast reactor, analyzing various fuel and in-core fuel cycle options to achieve a true breed and burn core, and performing a design basis Loss of Coolant Accident (LOCA) analysis on that design. For the physics analysis, a 1/8 core model was created using different enrichments and simulated equilibrium fuel loadings. The model was used to locate the hot spot of the reactor, and the peak to average energy deposition at that location. The model was also used to create contour plots of the flux and energy deposition over the volume of the reactor. The eigenvalue over time was evaluated using three different fuel configurations with the same core geometry. The breeding capabilities of this configuration were excellent for a 7% U-235 model and good in both a plutonium model and a 14% U-235 model. Changing the fuel composition from the Pu fuel which provided about 78% U-238 for breeding to the 14% U-235 fuel with about 86% U-238 slowed the rate of decrease in the eigenvalue a noticeable amount. Switching to the 7% U-235 fuel with about 93% U-238 showed an increase in the eigenvalue over time. For the thermal-hydraulic analysis, the reactor design used was the one forwarded by the MIT team. This reactor design uses helium coolant, a Brayton cycle, and has a thermal power of 600 MW. The core design parameters were supplied by MIT; however, the other key reactor components that were necessary for a plausible simulation of a LOCA were not defined. The thermal-hydraulic and plant design research concentrated on determining reasonable values for those undefined components. The LOCA simulation was intended to provide insights on the influence of the Reactor Cavity Cooling System (RCCS), the

  17. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

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

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

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

  19. Improved safety fast reactor with “reservoir” for delayed neutrons generating

    Science.gov (United States)

    Kulikov, G. G.; Apse, V. A.; Shmelev, A. N.; Kulikov, E. G.

    2017-01-01

    The paper considers the possibility to improve safety of fast reactors by using weak neutron absorber with large atomic weight as a material for external neutron reflector and for internal cavity in the reactor core (the neutron “reservoir”) where generation of some additional “delayed” neutron takes place. The effects produced by the external neutron reflector and the internal neutron “reservoir” on kinetic behavior of fast reactors are inter-compared. It is demonstrated that neutron kinetics of fast reactors with such external and internal zones becomes the quieter as compared with neutron kinetics of thermal reactors.

  20. Uranium self-shielding in fast reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-03-01

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

  1. The value of helium-cooled reactor technologies for transmutation of nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, C.; Baxter, A. [General Atomics, Los Alamos, NM (United States)

    2001-07-01

    Helium-cooled reactor technologies offer significant advantages in accomplishing the waste transmutation process. They are ideally suited for use with thermal, epithermal, or fast neutron energy spectra. They can provide a relatively hard thermal neutron spectrum for transmutation of fissionable materials such as Pu-239 using ceramic-coated transmutation fuel particles, a graphite moderator, and a non-fertile burnable poison. These features (1) allow deep levels of transmutation with minimal or no intermediate reprocessing, (2) enhance passive decay heat removal via heat conduction and radiation, (3) allow operation at relatively high temperatures for a highly efficient generation of electricity, and (4) discharge the transmuted waste in a form that is highly resistant to corrosion for long times. They also offer the possibility for the use of epithermal neutrons that can interact with transmutable materials more effectively because of the large atomic cross sections in this energy domain. A fast spectrum may be useful for deep burnup of certain minor actinides. For this application, helium is essentially transparent to neutrons, does not degrade neutron energies, and offers the hardest possible neutron energy environment. In this paper, we report results from recent work on materials transmutation balances, safety, value to a geological repository, and economic considerations. (authors)

  2. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

  3. Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

    Science.gov (United States)

    Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

    2014-08-01

    The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

  4. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  5. Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor.

    Science.gov (United States)

    Yavar, A R; Sarmani, S B; Wood, A K; Fadzil, S M; Radir, M H; Khoo, K S

    2011-05-01

    Determination of thermal to fast neutron flux ratio (f(fast)) and fast neutron flux (ϕ(fast)) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f(fast) and subsequently ϕ(fast) were determined using the absolute method. The f(fast) ranged from 48 to 155, and the ϕ(fast) was found in the range 1.03×10(10)-4.89×10(10) n cm(-2) s(-1). These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

  6. Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

    2005-09-01

    As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong

  7. The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.R.

    1986-01-01

    This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

  8. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Science.gov (United States)

    Sulaiman, S. A.; Dominguez-Ontiveros, E. E.; Alhashimi, T.; Budd, J. L.; Matos, M. D.; Hassan, Y. A.

    2015-04-01

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A&M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  9. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  10. Data on test results of vessel cooling system of high temperature engineering test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Saikusa, Akio [Secretariat of Nuclear Safety Commission, Tokyo (Japan); Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-02-01

    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

  11. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1989-08-01

    We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

  12. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  13. Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior

    Directory of Open Access Journals (Sweden)

    Ordóñez Ródenas José

    2016-01-01

    Full Text Available One of the goals of Generation IV reactors is to increase safety from those of previous generations. Different research platforms have been identified the need to improve the reliability of the simulation tools to ensure the capability of the plant to accommodate the design basis transients established in preliminary safety studies. The paper describes the modelling of primary pumps in advanced sodium cooled reactors using the TRACE code. Following the implementation of the models, the results obtained in the analysis of different design basis transients are compared with the simplifying approximations used in reference models. The paper shows the process to obtain a consistent pump model of the ESFR (European Sodium Fast Reactor design and the analysis of loss of flow transients triggered by pumps coast–down analyzing the thermal hydraulic neutronic coupled system response. A sensitivity analysis of the system pressure drops effect and the other relevant parameters that influence the natural convection after the pumps coast–down is also included.

  14. Optimization of a heterogeneous fast breeder reactor core with improved behavior during unprotected transients

    Energy Technology Data Exchange (ETDEWEB)

    Poumerouly, S.; Schmitt, D.; Massara, S.; Maliverney, B. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)

    2012-07-01

    Innovative Sodium-cooled Fast Reactors (SFRs) are currently being investigated by CEA, AREVA and EDF in the framework of a joint French collaboration, and the construction of a GEN IV prototype, ASTRID (Advanced Sodium Technical Reactor for Industrial Demonstration), is scheduled in the years 2020. Significant improvements are expected so as to improve the reactor safety: the goal is to achieve a robust safety demonstration of the mastering of the consequences of a Core Disruptive Accident (CDA), whether by means of prevention or mitigation features. In this framework, an innovative design was proposed by CEA in 2010. It aims at strongly reducing the sodium void effect, thereby improving the core behavior during unprotected loss of coolant transients. This design is strongly heterogeneous and includes, amongst others, a fertile plate, a sodium plenum associated with a B{sub 4}C upper blanket and a stepwise modulation of the fissile height of the core (onwards referred to as the 'diabolo shape'). In this paper, studies which were entirely carried out at EDF are presented: the full potential of this heterogeneous concept is thoroughly investigated using the SDDS methodology. (authors)

  15. In vessel detection of delayed neutron emitters from clad failure in sodium cooled nuclear reactors: An estimation of the signal

    Science.gov (United States)

    Filliatre, P.; Jammes, C.; Chapoutier, N.; Jeannot, J.-P.; Jadot, F.; Batail, R.; Verrier, D.

    2014-04-01

    The detection of clad failures is mandatory in sodium-cooled fast neutron reactors in compliance with the "clean sodium" concept. An in-vessel detection system, sensitive to delayed neutrons from fission products released into the primary coolant by failures, partially tested in SUPERPHENIX, is foreseen in current SFR projects in order to reduce significantly the delay before an alarm is issued. In this paper, an estimation of the signal received by such a system in case of a failure is derived, taking the French project ASTRID as a working example. This failure induced signal is compared to that of the contribution of the neutrons from the core itself. The sensitivity of the system is defined in terms of minimal detectable surface of clad failure. Possible solutions to improve this sensitivity are discussed, involving either the sensor itself, or the hydraulic design of the vessel in the early stage of the reactor conception.

  16. Feasibility study of fuel cladding performance for application in ultra-long cycle fast reactor

    Science.gov (United States)

    Jung, Ju Ang; Kim, Seung Hyun; Shin, Sang Hun; Bang, In Cheol; Kim, Ji Hyun

    2013-09-01

    As a part of the research and development activities for long-life core sodium-cooled fast reactors, the cladding performance of the ultra-long cycle fast reactor (UCFR) is evaluated with two design power levels (1000 MWe and 100 MWe) and cladding peak temperatures (873 K and 923 K). The key design concept of the UCFR is that it is non-refueling during its 30-60 years of operation. This concept may require a maximum peak cladding temperature of 923 K and a cladding radiation damage of over 200 dpa (displacements per atom). Therefore, for the design of the UCFR, deformation due to thermal creep, irradiation creep, and swelling must be taken into consideration through quantitative evaluations. As candidate cladding materials for use in UCFRs, ferritic-martensitic (FM) steels, oxide dispersion strengthened (ODS) steels, and SiC-based composite materials are studied using deformation behavior modeling for a feasibility evaluation. The results of this study indicate that SiC is a potential UCFR cladding material, with the exception of irradiation creep due to high neutron fluence stemming from its long operating time of about 30-60 years.

  17. Evaluation of eddy-current probe signals due to cracks in ferromagnetic parts of fast reactor

    Science.gov (United States)

    Wu, Tao; Bowler, John R.

    2017-02-01

    Eddy current testing to evaluate the condition of metallic parts in a sodium cooled fast reactor under standby conditions is challenging due to the presence of liquid sodium at 250 °C. The eddy current test system should be sensitive enough to capture small signal changes and hence an advanced inspection systems is needed. We have developed new hardware and improved numerical models to predict the eddy current probe signal due to cracks in metallic fast reactor parts by using volume integral equation method. The analytical expressions are derived for the quasi-static time-harmonic electromagnetic fields of a circular eddy current coil which interacts with conductive plate. Naturally, the method of moment is used to approximate the integral equation and obtain the discrete approximation of the field in the crack domain. A simple and accurate analytical method for dealing with the hyper-singularity element evaluation is also provided. An accurate controlled experiment is carried out on the ferromagnetic stainless steel plate with precision made notch to obtain reference impedance changes for comparison with the theoretical model predictions. Good agreement between predictions and experiment is obtained.

  18. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  19. Fuel burn analysis of a sodium fast reactor with KANEXT and Serpent; Analisis de quemado de combustible de un reactor rapido de sodio con KANEXT y SERPENT

    Energy Technology Data Exchange (ETDEWEB)

    Lopez S, R. C.; Francois L, J. L., E-mail: rcarlos.lope@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2015-09-15

    The fast reactors cooled by sodium are one of the options considered in the Generation IV. Since most of the reactors of Fourth Generation are still in development stage, is necessary to have efficient and reliable computational tools, this in order to obtain accurate results in reasonable computational times. In this paper is introduced and describes the deterministic code KANEXT (KArlsruhe Neutronic EXtended Tool) and is compared against a Monte Carlo code of more diffusion: Serpent. KANEXT, being a modular code requires the interaction of different modules to perform a job, this interaction of modules is described in this article. The parameters to be compared are the results of the neutron multiplication effective factor and the evolution of isotopes during the burning. The mentioned comparison is carried out for a fast reactor cooled by sodium of relatively small size compared to commercial size reactors. In this paper the particularities of the reactor are described, important for the analysis such as geometry, enrichments, reflector, etc. The considerations in the implementation in both codes are also described, as are simplifications, length of the burning steps, possible solutions of the Bateman equations for the burning fuel in Serpent and the solution options for transport (P3) and diffusion (P1) in KANEXT. The results show good correspondence between Serpent and KANEXT, which give confidence to continue using KANEXT as the main tool. Respect to computation time, time saving is evident with the use of deterministic codes instead of Monte Carlo codes, in this particular case, the time savings using KANEXT is about 98.5% of the time used by Serpent. (Author)

  20. Designing a SCADA system simulator for fast breeder reactor

    Science.gov (United States)

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

    2016-04-01

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

  1. Cellular convection in vertical annuli of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hemanath, M.G. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)], E-mail: hemanath@igcar.gov.in; Meikandamurthy, C.; Ramakrishnan, V.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2007-08-15

    In the pool type fast reactors the roof structure is penetrated by a number of pumps and heat exchangers that are cylindrical in shape. Sandwiched between the free surface of sodium and the roof structure, is stagnant argon gas, which can flow in the annular space between the components and roof structure, as a thermosyphon. These thermosyphons not only transport heat from sodium to roof structure, but also result in cellular convection in vertical annuli resulting in circumferential temperature asymmetry of the penetrating components. There is need to know the temperature asymmetry as it can cause tilting of the components. Experiments were carried out in an annulus model to predict the circumferential temperature difference with and without sodium in the test vessel. Three-dimensional analysis was also carried out using PHOENICS CFD code and compared with the experiment. This paper describes the experimental details, the theoretical analysis and their comparison.

  2. Limitations of eddy current testing in a fast reactor environment

    Science.gov (United States)

    Wu, Tao; Bowler, John R.

    2016-02-01

    The feasibility of using eddy current probes for detecting flaws in fast nuclear reactor structures has been investigated with the aim of detecting defects immersed in electrically conductive coolant including under liquid sodium during standby. For the inspections to be viable, there is a need to use an encapsulated sensor system that can be move into position with the aid of visualization tools. The initial objective being to locate the surface to be investigated using, for example, a combination of electromagnetic sensors and sonar. Here we focus on one feature of the task in which eddy current probe impedance variations due to interaction with the external surface of a tube are evaluated in order to monitor the probe location and orientation during inspection.

  3. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  4. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  5. Microbial fouling community analysis of the cooling water system of a nuclear test reactor with emphasis on sulphate reducing bacteria.

    Science.gov (United States)

    Balamurugan, P; Joshi, M Hiren; Rao, T S

    2011-10-01

    Culture and molecular-based techniques were used to characterize bacterial diversity in the cooling water system of a fast breeder test reactor (FBTR). Techniques were selected for special emphasis on sulphate-reducing bacteria (SRB). Water samples from different locations of the FBTR cooling water system, in addition to biofilm scrapings from carbon steel coupons and a control SRB sample were characterized. Whole genome extraction of the water samples and SRB diversity by group specific primers were analysed using nested PCR and denaturing gradient gel electrophoresis (DGGE). The results of the bacterial assay in the cooling water showed that the total culturable bacteria (TCB) ranged from 10(3) to 10(5) cfu ml(-1); iron-reducing bacteria, 10(3) to 10(5) cfu ml(-1); iron oxidizing bacteria, 10(2) to 10(3) cfu ml(-1) and SRB, 2-29 cfu ml(-1). However, the counts of the various bacterial types in the biofilm sample were 2-3 orders of magnitude higher. SRB diversity by the nested PCR-DGGE approach showed the presence of groups 1, 5 and 6 in the FBTR cooling water system; however, groups 2, 3 and 4 were not detected. The study demonstrated that the PCR protocol influenced the results of the diversity analysis. The paper further discusses the microbiota of the cooling water system and its relevance in biofouling.

  6. Simulation of hydrocarbons pyrolysis in a fast-mixing reactor

    Institute of Scientific and Technical Information of China (English)

    MG Ktalkherman; IG Namyatov

    2015-01-01

    Currently, thermal decomposition of hydrocarbons for the production of basic petrochemicals (ethylene, propyl-ene) is carried out in steam-cracking processes. Aside from the conventional method, under consideration are alternative ways purposed for process intensification. In the context of these activities, the method of high-temperature pyrolysis of hydrocarbons in a heat-carrier flow is studied, which differs from previous ones and is based on the ability of an ultra-short time of feedstock/heat-carrier mixing. This enables to study the pyrolysis process at high temperature (up to 1500 K) at the reactor inlet. A set of model experiments is conducted on the lab scale facility. Liquefied petroleum gas (LPG) and naphtha are used as a feedstock. The detailed data are obtain-ed on temperature and product distributions within a wide range of the residence time. A theoretical model based on the detailed kinetics of the process is developed, too. The effect of governing parameters on the pyrolysis process is analyzed by the results of the simulation and experiments. In particular, the optimal temperature is detected which corresponds to the maximum ethylene yield. Product yields in our experiments are compared with the similar ones in the conventional pyrolysis method. In both cases (LPG and naphtha), ethylene selectivity in the fast-mixing reactor is substantial y higher than in current technology.

  7. Preparation of U–Zr–Mn, a Surrogate Alloy for Recycling Fast Reactor Fuel

    Directory of Open Access Journals (Sweden)

    Jong-Hwan Kim

    2015-01-01

    Full Text Available Metallic fuel slugs of U–10Zr–5Mn (wt%, a surrogate alloy for the U–TRU–Zr (TRU: a transuranic element alloys proposed for sodium-cooled fast reactors, were prepared by injection casting in a laboratory-scale furnace, and their characteristics were evaluated. As-cast U–Zr–Mn fuel rods were generally sound, without cracks or thin sections. Approximately 68% of the original Mn content was lost under dynamic vacuum and the resulting slug was denser than those prepared under Ar pressure. The concentration of volatile Mn was as per the target composition along the entire length of the rods prepared under 400 and 600 Torr. Impurities, namely, oxygen, carbon, silicon, and nitrogen, totaled less than 2,000 ppm, satisfying fuel criteria.

  8. U.S. Sodium Fast Reactor Codes and Methods: Current Capabilities and Path Forward

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, A. J.; Fanning, T. H.

    2017-06-26

    The United States has extensive experience with the design, construction, and operation of sodium cooled fast reactors (SFRs) over the last six decades. Despite the closure of various facilities, the U.S. continues to dedicate research and development (R&D) efforts to the design of innovative experimental, prototype, and commercial facilities. Accordingly, in support of the rich operating history and ongoing design efforts, the U.S. has been developing and maintaining a series of tools with capabilities that envelope all facets of SFR design and safety analyses. This paper provides an overview of the current U.S. SFR analysis toolset, including codes such as SAS4A/SASSYS-1, MC2-3, SE2-ANL, PERSENT, NUBOW-3D, and LIFE-METAL, as well as the higher-fidelity tools (e.g. PROTEUS) being integrated into the toolset. Current capabilities of the codes are described and key ongoing development efforts are highlighted for some codes.

  9. Cryogenic Cooling System for 5 kA, 200 μH Class HTS DC Reactor

    Science.gov (United States)

    Park, Heecheol; Kim, Seokho; Kim, Kwangmin; Park, Minwon; Park, Taejun; Kim, A.-rong; Lee, Sangjin

    DC reactors, made by aluminum busbar, are used to stabilize the arc of an electric furnace. In the conventional arc furnace, the transport current is several tens of kilo-amperes and enormous resistive loss is generated. To reduce the resistive loss at the DC reactor, a HTS DC reactor can be considered. It can dramatically improve the electric efficiency as well as reduce the installation space. Similar with other superconducting devices, the HTS DC reactor requires current leads from a power source in room temperature to the HTS coil in cryogenic environment. The heat loss at the metal current leads can be minimized through optimization process considering the geometry and the transport current. However, the transport current of the HTS DC reactor for the arc furnace is much larger than most of HTS magnets and the enormous heat penetration through the current lead should be effectively removed to keep the temperature around 70∼77 K. Current leads are cooled down by circulation of liquid nitrogen from the cooling system with a stirling cryocooler. The operating temperature of HTS coil is 30∼40 K and circulation of gaseous helium is used to remove the heat generation at the HTS coil. Gaseous helium is transported through the cryogenic helium blower and a single stage GM cryocooler. This paper describes design and experimental results on the cooling system for current leads and the HTS coil of 5 kA, 200 μH class DC reactor as a prototype. The results are used to verify the design values of the cooling systems and it will be applied to the design of scale-up cooling system for 50 kA, 200 μH class DC reactor.

  10. Circulating and plateout activity program for gas-cooled reactors with arbitrary radioactive chains

    Energy Technology Data Exchange (ETDEWEB)

    Apperson, C.E. Jr.

    1978-03-01

    A time-dependent method for estimating the fuel body, circulating, plateout, and filter inventory of a high temperature gas-cooled reactor (HTGR) during normal operation is discussed. The primary coolant model accounts for the source, buildup, decay, and cleanup of isotopes that are gas borne inside the prestressed concrete reactor vessel (PCRV). This method has been implemented in the SUVIUS computer program that is described in detail.

  11. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E.M.J.; Bogaard, J.P.A. van den

    1995-06-01

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

  12. Measurement and evaluation of Corrosion Products deposition distribution in the Experimental Fast Reactor JOYO

    Energy Technology Data Exchange (ETDEWEB)

    Aoyama, Takafumi; Sumino, Kozo [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Masui, Tomohiko; Saikawa, Takuya

    1997-12-01

    The Corrosion Product (CP) is the major radiation source in the primary cooling system of an LMFBR plant. It is important to characterize and predict the CP behavior to reduce the personnel exposure dose due to CP deposition. The CP measurement was carried out in the Experimental Fast Reactor JOYO during the 11th annual inspection period when the accumulated reactor thermal power reached about 143 GWd. The CP deposition density was measured using a pure germanium detector. The plastic scintillation fiber (PSF) was applied for the gamma-ray dose rate distribution measurement and compared with the thermoluminescence dosimeter (TLD). The major results obtained by the CP measurements in JOYO are the follows: (1) The major CP nuclides deposited in the primary cooling system are {sup 54}Mn and {sup 60}Co. {sup 54}Mn is the dominant isotope and it tends to deposit in the cold leg region. On the other hand, {sup 60}Co deposits mainly in the hot leg region. The deposition density of {sup 54}Mn is about seven times as much as that of {sup 60}Co in the cold leg region and twice in the hot leg region. (2) The deposition densities of {sup 54}Mn and {sup 60}Co, and the gamma-dose rate were decreased from the last data in the previous annual inspection period mainly due to the short operation time and the longer cooling time. (3) The continuous gamma-ray dose rate distribution up to 10m can be measured by using the PSF in a few minutes. The PSF is suitable to measure the gamma-ray dose rate distribution in the maintenance work area where it is narrow and the mixture of gamma-ray sources from primary pipings and components. The data base of detailed gamma-ray dose rate distribution was greatly extended by the PSF. (author)

  13. Fast optical cooling of a nanomechanical cantilever by a dynamical Stark-shift gate

    Science.gov (United States)

    Yan, Leilei; Zhang, Jian-Qi; Zhang, Shuo; Feng, Mang

    2015-10-01

    The efficient cooling of nanomechanical resonators is essential to exploration of quantum properties of the macroscopic or mesoscopic systems. We propose such a laser-cooling scheme for a nanomechanical cantilever, which works even for the low-frequency mechanical mode and under weak cooling lasers. The cantilever is coupled by a diamond nitrogen-vacancy center under a strong magnetic field gradient and the cooling is assisted by a dynamical Stark-shift gate. Our scheme can effectively enhance the desired cooling efficiency by avoiding the off-resonant and undesired carrier transitions, and thereby cool the cantilever down to the vicinity of the vibrational ground state in a fast fashion.

  14. Fast optical cooling of a nanomechanical cantilever by a dynamical Stark-shift gate

    CERN Document Server

    Yan, Leilei; Zhang, Shuo; Feng, Mang

    2014-01-01

    The efficient cooling of the nanomechanical resonators is essential to exploration of quantum properties of the macroscopic or mesoscopic systems. We propose such a laser-cooling scheme for a nanomechanical cantilever, which works even for the low-frequency mechanical mode and under weak cooling lasers. The cantilever is attached by a diamond nitrogen-vacancy center under a strong magnetic field gradient and the cooling is assisted by a dynamical Stark-shift gate. Our scheme can effectively enhance the desired cooling efficiency by avoiding the off-resonant and unexpected carrier transitions, and thereby cool the cantilever down to the vicinity of the vibrational ground state in a fast fashion.

  15. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or

  16. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  17. Summarized compatibility review of reactor materials for CO2-cooled graphite-moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Seddon, B.J.

    1964-09-23

    This report, which is a revised edition of TRG-Report-267, summarises an internal document and collates information on the compatibility of a range of materials used in CO{sub 2}-cooled graphite-moderated reactors. Information is presented in the form of six tables based on compatibilities of materials with carbon dioxide, beryllium, Magnox, magnesium, uranium and compatibilities of pairs of other relevant materials.

  18. Heat pipe cooled reactors for multi-kilowatt space power supplies

    Energy Technology Data Exchange (ETDEWEB)

    Ranken, W.A.; Houts, M.G.

    1995-01-01

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFEs) to radiator heat pipes.

  19. Heat pipe cooled reactors for multi-kilowatt space power supplies

    Science.gov (United States)

    Ranken, W. A.; Houts, M. G.

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFE's) to radiator heat pipes.

  20. Gas Cooled, Natural Uranium, D20 Moderated Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dahlberg, R.C.; Beasley, E.G.; DeBoer, T.K.; Evans, T.C.; Molino, D.F.; Rothwell, W.S.; Slivka, W.R.

    1956-08-01

    The attractiveness of a helium cooled, heavy water moderated, natural uranium central station power plant has been investigated. A fuel element has been devised which allows the D20 to be kept at a low pressure while the exit gas temperature is high. A preliminary cost analysis indicates that, using currently available materials, competitive nuclear power in foreign countries is possible.

  1. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  2. Determination of the reactivity coefficient of a sodium cooled reactor with metallic fuel; Bestimmung der Reaktivitaetskoeffizienten eines schnellen natriumgekuehlten Reaktors mit metallischem Brennstoff

    Energy Technology Data Exchange (ETDEWEB)

    Guilliard, N. [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE)

    2013-07-01

    Fast sodium cooled breeding reactors are of interest in the frame of the Generation IV reactor design. Die to the experience in France and Japan the concept seems to be realizable in the near future. Due to the new design concepts the accident scenarios and the safety analyses based on modern simulation codes have to be adjusted in the frame of the European JASMIN project. The project is aimed to develop a European accident code for fast breeder reactors based on the modular LWR code ASTEC. Extensions with respect to sodium as coolant, improved physical models and the different design are necessary. Besides this a point kinetic model shall be implemented. The coupling of point kinetic neutronics to a thermal hydraulic code requires the determination of the reactivity coefficients of the respective system. Using the core design of a benchmark specification OECD/NEA SFR task force the reactivity coefficients are determined as an example.

  3. Scaled Facility Design Approach for Pool-Type Lead-Bismuth Eutectic Cooled Small Modular Reactor Utilizing Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sangrok; Shin, Yong-Hoon; Lee, Jueun; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2015-10-15

    In low carbon era, nuclear energy is the most prominent energy source of electricity. For steady ecofriendly nuclear energy supply, Generation IV reactors which are future nuclear reactor require safety, sustainability, economics and non-proliferation as four criteria. Lead cooled fast reactor (LFR) is one of these reactor type and Generation IV international forum (GIF) adapted three reference LFR systems which are a small and movable systems with long life without refueling, intermediate size and huge electricity generation system for power grid. NUTRECK (Nuclear Transmutation Energy Center of Korea) has been designed reactor called URANUS (Ubiquitous, Rugged, Accident-forgiving, Non-proliferating, and Ultra-lasting Sustainer) which is small modular reactor and using lead-bismuth eutectic coolant. To prove natural circulation capability of URANUS and analyze design based accidents, scaling mock-up experiment facility will be constructed. In this paper, simple specifications of URANUS will be presented. Then based on this feature, scaling law and scaled facility design results are presented. To validate safety feature and thermodynamics characteristic of URANUS, scaled mockup facility of URANUS is designed based on the scaling law. This mockup adapts two area scale factors, core and lower parts of mock-up are scaled for 3D flow experiment. Upper parts are scaled different size to reduce electricity power and LBE tonnage. This hybrid scaling method could distort some thermal-hydraulic parameters, however, key parameters for experiment will be matched for up-scaling. Detailed design of mock-up will be determined through iteration for design optimization.

  4. MR-6 type fuel elements cooling in natural convection conditions after the reactor shut down

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Bykowski, W.; Moldysz, A. [Institute of Atomic Energy, Otwock Swierk (Poland)

    2002-07-01

    Natural cooling conditions of the nuclear fuel in the channel type reactor after its shut down are commonly determined with relatively high uncertainty. This is not only to he lack of adequate measurements of thermal parameters i.e. the residual power generation, the coolant flow and temperatures, but also due to indeterminate model of convection mechanism. The numerical simulation of natural convection in multitube fuel assembly in the fuel channel leads to various convection modes including evidently chaotic behaviour. To determine the real cooling conditions in the MARIA research reactor a series of experiments has been performed with fuel assembly equipped with a set of thermocouples. After some forced cooling period (the shortest was half an hour after the reactor shut down) the reactor was left with the only natural convection. Two completely different cooling modes have been observed. The MARIA core consists of series of individual fuel channel and so called bypasses, maintaining the hydraulic properties of the fuel channel, connected in parallel. Initially, the convection cells were established trough few so-called bypasses providing a very effective mode of cooling. In this mode the flow charts were identical to those existing in forced cooling mode. After certain period the system switched on the second cooling mode with natural circulation within the individual fuel cells. Higher temperatures and temperature fluctuations were characteristic for this mode approaching 30 deg in amplitude. In almost all the cases the system was switching few times between modes, but eventually remained in the second mode. The switching times were not regular and the process has a chaotic behaviour. (author)

  5. Hydraulic Experiment for Simulative Assemblies of Blanket Assembly and Np Transmutation Assembly of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    CHENG; Dao-xi; QI; Xiao-guang; ZHAI; Wei-ming; YANG; Bing; ZHOU; Ping

    2013-01-01

    The out-of reactor hydraulic experiment of fast reactor assembly is one of the important experiments in the process of the development of the fast reactor assembly.In this experiment,the size of the throttling element in the foot of the assembly is decided which is fit for the flow division in the reactor and the

  6. Safe design of cooled tubular reactors for exothermic multiple reactions: Multiple-reaction networks

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1988-01-01

    The model of the pseudo-homogeneous, one-dimensional cooled tubular reactor is applied to a multiple-reaction network. It is demonstrated for a network which consists of two parallel and two consecutive reactions. Three criteria are developed to obtain an integral yield which does not deviate more t

  7. Safe design of cooled tubular reactors for exothermic multiple reactions: Multiple-reaction networks

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1988-01-01

    The model of the pseudo-homogeneous, one-dimensional cooled tubular reactor is applied to a multiple-reaction network. It is demonstrated for a network which consists of two parallel and two consecutive reactions. Three criteria are developed to obtain an integral yield which does not deviate more

  8. Safe design of cooled tubular reactors for exothermic, multiple reactions. Consecutive reactions

    NARCIS (Netherlands)

    Westerterp, K.R.; Overtoom, R.R.M.

    1985-01-01

    The model of the pseudo-homogeneous, one-dimensional, cooled tubular reactor is applied to two consecutive, irreversible first order reactions. A criterion is derived to obtain a desired integral yield. Based on this criterion three requirements are formulated, which enable us to choose the relevant

  9. Thermally safe operation of a cooled semi-batch reactor: slow liquid-liquid reactions

    NARCIS (Netherlands)

    Steensma, M.; Westerterp, K.R.

    1988-01-01

    Thermally safe operation of a semi-batch reactor (SBR) implies that conditions leading to strong accumulation of unreacted reactants must be avoided. All thermal responses of a SBR, in which a slow liquid-liquid reaction takes place, can be represented in a diagram with the kinetics, cooling capacit

  10. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  11. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  12. Safe design of cooled tubular reactors for exothermic, multiple reactions; parallel reactions—I: Development of criteria

    NARCIS (Netherlands)

    Westerterp, K.R.; Ptasiński, K.J.

    1984-01-01

    Previously reported design criteria for cooled tubular reactors are based on the prevention of reactor temperature run away and were developed for single reactions only. In this paper it is argued that such criteri a should be based on the reactor selectivity, from which eventually a maximum

  13. On the Burning of Plutonium Originating from Light Water Reactor Use in a Fast Molten Salt Reactor—A Neutron Physical Study

    Directory of Open Access Journals (Sweden)

    Bruno Merk

    2015-11-01

    Full Text Available An efficient burning of the plutonium produced during light water reactor (LWR operation has the potential to significantly improve the sustainability indices of LWR operations. The work offers a comparison of the efficiency of Pu burning in different reactor configurations—a molten salt fast reactor, a LWR with mixed oxide (MOX fuel, and a sodium cooled fast reactor. The calculations are performed using the HELIOS 2 code. All results are evaluated against the plutonium burning efficiency determined in the Consommation Accrue de Plutonium dans les Réacteurs à Neutrons RApides (CAPRA project. The results are discussed with special view on the increased sustainability of LWR use in the case of successful avoidance of an accumulation of Pu which otherwise would have to be forwarded to a final disposal. A strategic discussion is given about the unavoidable plutonium production, the possibility to burn the plutonium to avoid a burden for the future generations which would have to be controlled.