This study presents a benchmark evaluation of the new ENDF/B-VIII.0 nuclear data library for the Organization for Economic Co-operation and Development/Nuclear Energy Agency Medium 1000 MWth sodium-cooled fast reactor...This study presents a benchmark evaluation of the new ENDF/B-VIII.0 nuclear data library for the Organization for Economic Co-operation and Development/Nuclear Energy Agency Medium 1000 MWth sodium-cooled fast reactor(SFR).The study presented herein covers both SFR core types,i.e.,metallic fueled(MET-1000)and oxide fueled(MOX-1000),simulated using the continuous-energy Monte Carlo Serpent2 code.The neutronics performances of the ENDF/B-VIII.0-based simulations were compared mainly to two libraries:ENDF/B-VII.1 and JENDL-4.0.The comparison includes several neutronics parameters evaluated for the beginning and end of the cycle conditions.These parameters include the effective multiplication factor keff,total effective delayed neutron fraction beff,sodium void reactivity(DqNa),Doppler constant(DqDoppler),and control rod worth(DqCR).In addition,a sensitivity study was used to reveal the major isotope/reaction pairs contributing to the discrepancy observed in the performance of the three libraries using 33 and 44-energy-group structures.展开更多
The study gives a brief introduction on development of innovated nuclear system in China,mainly focus on the materials R&D status for the sodium cooled fast reactor.With the high speed development of economy,China...The study gives a brief introduction on development of innovated nuclear system in China,mainly focus on the materials R&D status for the sodium cooled fast reactor.With the high speed development of economy,China needs a huge energy supply;at same time a more cleaning energy to reduce the carbon release is demanded.The nuclear energy is the most cleaning energy at present time,especially the innovated nuclear system which is so-called GenerationⅣpower plants has got its prior development due to its safety, economical and little fission production produced.Fast breeder reactor,as the priority development reactor type in the Gen-Ⅳnuclear system,is the key to the advanced closed fuel cycle technologies.China experimental fast reactor(CEFR ) has been completed the design,construction the synthesis system commissioning and reached its physical criticality on July 21,2010.At China Institute of Atomic Energy,the CEFR and other research facilities have been established,and extensive studies are planning to carry out in the areas of fuel and materials development.This will laid the foundation for the design and development of the future's CFR—900(China Demonstration Fast Reactor) and CCFR(China Commercial Fast Reactor). Highlights of some of materials R&D studies are discussed in this paper.展开更多
A water leakage on the surface of heat transfer tube in a steam generator of sodium-cooled fast reactor causes SWR （sodium-water reaction）. The SWR damages the leak surface and gives rise to the leak enlargement. Mo...A water leakage on the surface of heat transfer tube in a steam generator of sodium-cooled fast reactor causes SWR （sodium-water reaction）. The SWR damages the leak surface and gives rise to the leak enlargement. Most of initial leakage starts from micro leak （less than 0.5 g/s）. However, the leak rate increases more than two orders of magnitude and the resultant leak damages surrounding heat transfer tubes and it brings secondary failure of the heat transfer tube. Evaluation of the leak enlargement is necessary to assess the leak rate increase, so that evaluate the possibility of secondary failure. In this study, a simulant experiment, which uses neutralization reaction, is proposed to reproduce the leak enlargement. To examine the feasibility of the experiment, numerical simulations are carried out. From the result, a funnel-shaped nozzle enlargement is observed and the shape similar to the shape of the enlarged nozzle from the SWAT （sodium-water reaction test loop） experiment.展开更多
The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT （control-rod guide tube） during CDAs （core-disruptive accidents） of SFRs （sodium-cooled fast reactors）, si...The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT （control-rod guide tube） during CDAs （core-disruptive accidents） of SFRs （sodium-cooled fast reactors）, since fuel discharge will decrease the core reactivity and CRGTs have a potential to provide an effective discharge path. Fuel discharge contains multi-component fluid dynamics with phase changes, and, in the present study, the SFR safety analysis code SIMMER （Sn, implicit, multifield, multicomponent, Eulerian recriticality） was utilized as a technical basis. First, dominant phenomena affecting fuel discharge through the CRGT are identified based on parametric calculations by the SIMMER code. Next, validations on the code models closely relating to these phenomena were carried out based on experimental data. It was shown that the SIMMER code with some model modifications could reproduce the experimental results appropriately. Through the present study, the evaluation methodology for the molten-fuel discharge through the CRGT was successfully developed.展开更多
The supercritical carbon dioxide(SCO_(2))Brayton cycle has become an ideal power conversion system for sodium-cooled fast reactors(SFR)due to its high efficiency,compactness,and avoidance of sodiumwater reaction.In th...The supercritical carbon dioxide(SCO_(2))Brayton cycle has become an ideal power conversion system for sodium-cooled fast reactors(SFR)due to its high efficiency,compactness,and avoidance of sodiumwater reaction.In this paper,the 1200 MWe large pool SFR(CFR1200)is used as the heat source of the system,and the sodium circuit temperature and the heat load are the operating boundaries of the cycle system.The performance of different SCO_(2) Brayton cycle systems and changes in key equipment performance are compared.The study indicates that the inter-stage cooling and recompression cycle has the best match with the heat source characteristics of the SFR,and the cycle efficiency is the highest(40.7%).Then,based on the developed system transient analysis program(FR-Sdaso),a pool-type SFR power plant system analysis model based on the inter-stage cooling and recompression cycle is established.In addition,the matching between the inter-stage cooling recompression cycle and the SFR during the load cycle of the power plant is studied.The analysis shows that when the nuclear island adopts the flow-advanced operation strategy and the carbon dioxide flowrate in the SCO_(2) power conversion system is adjusted with the goal of maintaining the sodium-carbon dioxide heat exchanger sodium side outlet temperature unchanged,the inter-stage cooling recompression cycle can match the operation of the SFR very well.展开更多
基金the Research Institute of Science and Engineering at the University of Sharjah(No.1802040790-P).
文摘This study presents a benchmark evaluation of the new ENDF/B-VIII.0 nuclear data library for the Organization for Economic Co-operation and Development/Nuclear Energy Agency Medium 1000 MWth sodium-cooled fast reactor(SFR).The study presented herein covers both SFR core types,i.e.,metallic fueled(MET-1000)and oxide fueled(MOX-1000),simulated using the continuous-energy Monte Carlo Serpent2 code.The neutronics performances of the ENDF/B-VIII.0-based simulations were compared mainly to two libraries:ENDF/B-VII.1 and JENDL-4.0.The comparison includes several neutronics parameters evaluated for the beginning and end of the cycle conditions.These parameters include the effective multiplication factor keff,total effective delayed neutron fraction beff,sodium void reactivity(DqNa),Doppler constant(DqDoppler),and control rod worth(DqCR).In addition,a sensitivity study was used to reveal the major isotope/reaction pairs contributing to the discrepancy observed in the performance of the three libraries using 33 and 44-energy-group structures.
文摘The study gives a brief introduction on development of innovated nuclear system in China,mainly focus on the materials R&D status for the sodium cooled fast reactor.With the high speed development of economy,China needs a huge energy supply;at same time a more cleaning energy to reduce the carbon release is demanded.The nuclear energy is the most cleaning energy at present time,especially the innovated nuclear system which is so-called GenerationⅣpower plants has got its prior development due to its safety, economical and little fission production produced.Fast breeder reactor,as the priority development reactor type in the Gen-Ⅳnuclear system,is the key to the advanced closed fuel cycle technologies.China experimental fast reactor(CEFR ) has been completed the design,construction the synthesis system commissioning and reached its physical criticality on July 21,2010.At China Institute of Atomic Energy,the CEFR and other research facilities have been established,and extensive studies are planning to carry out in the areas of fuel and materials development.This will laid the foundation for the design and development of the future's CFR—900(China Demonstration Fast Reactor) and CCFR(China Commercial Fast Reactor). Highlights of some of materials R&D studies are discussed in this paper.
文摘A water leakage on the surface of heat transfer tube in a steam generator of sodium-cooled fast reactor causes SWR （sodium-water reaction）. The SWR damages the leak surface and gives rise to the leak enlargement. Most of initial leakage starts from micro leak （less than 0.5 g/s）. However, the leak rate increases more than two orders of magnitude and the resultant leak damages surrounding heat transfer tubes and it brings secondary failure of the heat transfer tube. Evaluation of the leak enlargement is necessary to assess the leak rate increase, so that evaluate the possibility of secondary failure. In this study, a simulant experiment, which uses neutralization reaction, is proposed to reproduce the leak enlargement. To examine the feasibility of the experiment, numerical simulations are carried out. From the result, a funnel-shaped nozzle enlargement is observed and the shape similar to the shape of the enlarged nozzle from the SWAT （sodium-water reaction test loop） experiment.
文摘The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT （control-rod guide tube） during CDAs （core-disruptive accidents） of SFRs （sodium-cooled fast reactors）, since fuel discharge will decrease the core reactivity and CRGTs have a potential to provide an effective discharge path. Fuel discharge contains multi-component fluid dynamics with phase changes, and, in the present study, the SFR safety analysis code SIMMER （Sn, implicit, multifield, multicomponent, Eulerian recriticality） was utilized as a technical basis. First, dominant phenomena affecting fuel discharge through the CRGT are identified based on parametric calculations by the SIMMER code. Next, validations on the code models closely relating to these phenomena were carried out based on experimental data. It was shown that the SIMMER code with some model modifications could reproduce the experimental results appropriately. Through the present study, the evaluation methodology for the molten-fuel discharge through the CRGT was successfully developed.
基金the International Cooperative Research and Development Project on Key Technologies of the Fourth Generation Nuclear Energy System Sodium-cooled Fast Reactor(2016YFE0100800).
文摘The supercritical carbon dioxide(SCO_(2))Brayton cycle has become an ideal power conversion system for sodium-cooled fast reactors(SFR)due to its high efficiency,compactness,and avoidance of sodiumwater reaction.In this paper,the 1200 MWe large pool SFR(CFR1200)is used as the heat source of the system,and the sodium circuit temperature and the heat load are the operating boundaries of the cycle system.The performance of different SCO_(2) Brayton cycle systems and changes in key equipment performance are compared.The study indicates that the inter-stage cooling and recompression cycle has the best match with the heat source characteristics of the SFR,and the cycle efficiency is the highest(40.7%).Then,based on the developed system transient analysis program(FR-Sdaso),a pool-type SFR power plant system analysis model based on the inter-stage cooling and recompression cycle is established.In addition,the matching between the inter-stage cooling recompression cycle and the SFR during the load cycle of the power plant is studied.The analysis shows that when the nuclear island adopts the flow-advanced operation strategy and the carbon dioxide flowrate in the SCO_(2) power conversion system is adjusted with the goal of maintaining the sodium-carbon dioxide heat exchanger sodium side outlet temperature unchanged,the inter-stage cooling recompression cycle can match the operation of the SFR very well.
