Pub Date : 2020-06-01DOI: 10.11889/J.0253-3219.2020.HJS.43.060003
汪天雄, 张滕飞, 吴海成, 刘晓晶, 熊进标, 柴翔
[Background]Assembly calculation plays an important role in reactor core design.The depletion calculation accuracy of the assembly calculation is of great significance to the power distribution,refueling life and reactivity control design of nuclear reactor core.[Purpose]This study aims to evaluate the accuracy of the multigroup constant library in the neutron depletion calculation and establish a set of depletion experimental calculation model.[Methods]The depletion historical parameters and the final nuclide composition information were provided by the SFCOMPO-2.0 database.The information of the Takahama-3 reactor,H.B.Robinson-2 reactor and Beznau-1 reactor samples were modeled using the DRAGON code.Some simplified methods were applied to some unavailable data information.The numerical results were compared and analyzed with the benchmark results in the SFCOMPO-2.0 database.[Result&Conclusions]The results show that most of the nuclide productions are in good agreement with the reference value,and the errors are within 10%.Nuclides with large differences between numerical results and benchmark results are discussed simultaneously,and computational results of three reactor samples were compared and analyzed.
{"title":"Modeling and analysis of depletion experiment benchmark based on assembly calculation","authors":"汪天雄, 张滕飞, 吴海成, 刘晓晶, 熊进标, 柴翔","doi":"10.11889/J.0253-3219.2020.HJS.43.060003","DOIUrl":"https://doi.org/10.11889/J.0253-3219.2020.HJS.43.060003","url":null,"abstract":"[Background]Assembly calculation plays an important role in reactor core design.The depletion calculation accuracy of the assembly calculation is of great significance to the power distribution,refueling life and reactivity control design of nuclear reactor core.[Purpose]This study aims to evaluate the accuracy of the multigroup constant library in the neutron depletion calculation and establish a set of depletion experimental calculation model.[Methods]The depletion historical parameters and the final nuclide composition information were provided by the SFCOMPO-2.0 database.The information of the Takahama-3 reactor,H.B.Robinson-2 reactor and Beznau-1 reactor samples were modeled using the DRAGON code.Some simplified methods were applied to some unavailable data information.The numerical results were compared and analyzed with the benchmark results in the SFCOMPO-2.0 database.[Result&Conclusions]The results show that most of the nuclide productions are in good agreement with the reference value,and the errors are within 10%.Nuclides with large differences between numerical results and benchmark results are discussed simultaneously,and computational results of three reactor samples were compared and analyzed.","PeriodicalId":35563,"journal":{"name":"He Jishu/Nuclear Techniques","volume":"43 1","pages":"14-20"},"PeriodicalIF":0.0,"publicationDate":"2020-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46167514","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-06-01DOI: 10.11889/J.0253-3219.2020.HJS.43.060002
王傲, 申凤阳, 胡古, 郭键, 安伟健
[Background]With the in-depth exploration of space,the power supply requirements for energy supply are gradually increasing.Space nuclear reactor power supply stands out in the deep space exploration mission due to its advantages of passivity,long life and high reliability,hence heat pipe nuclear reactor has become the focus of research in the field of space nuclear reactor.[Purpose]This study aims to evaluate the progress and technology of heat pipe space nuclear reactor used as power supply.[Method]First of all,the conceptual design of heatpipe reactor in the early stage and its application in Kilopower was summarized.Then the details of heatpipe power system(HPS),the heatpipe-operated mars exploration reactor(HOMER),safe affordable fission engine(SAFE)and Kilopower were investigated with emphasis on the structure design,fuel selection,heatpipe arrangement and power design of each reactor.[Results]This survey provides ideas and references for future research and design of heatpipe space nuclear reactor power system.[Conclusions]The design of heatpipe reactor is highly flexible and innovative.The use of different alkali metal heat pipes and thermoelectric conversion methods has a direct impact on the system power and the quality of the entire system.The combination of heat pipe and various thermoelectric conversion methods can be attempted.
{"title":"A survey of heatpipe space nuclear reactor power supply","authors":"王傲, 申凤阳, 胡古, 郭键, 安伟健","doi":"10.11889/J.0253-3219.2020.HJS.43.060002","DOIUrl":"https://doi.org/10.11889/J.0253-3219.2020.HJS.43.060002","url":null,"abstract":"[Background]With the in-depth exploration of space,the power supply requirements for energy supply are gradually increasing.Space nuclear reactor power supply stands out in the deep space exploration mission due to its advantages of passivity,long life and high reliability,hence heat pipe nuclear reactor has become the focus of research in the field of space nuclear reactor.[Purpose]This study aims to evaluate the progress and technology of heat pipe space nuclear reactor used as power supply.[Method]First of all,the conceptual design of heatpipe reactor in the early stage and its application in Kilopower was summarized.Then the details of heatpipe power system(HPS),the heatpipe-operated mars exploration reactor(HOMER),safe affordable fission engine(SAFE)and Kilopower were investigated with emphasis on the structure design,fuel selection,heatpipe arrangement and power design of each reactor.[Results]This survey provides ideas and references for future research and design of heatpipe space nuclear reactor power system.[Conclusions]The design of heatpipe reactor is highly flexible and innovative.The use of different alkali metal heat pipes and thermoelectric conversion methods has a direct impact on the system power and the quality of the entire system.The combination of heat pipe and various thermoelectric conversion methods can be attempted.","PeriodicalId":35563,"journal":{"name":"He Jishu/Nuclear Techniques","volume":"43 1","pages":"7-13"},"PeriodicalIF":0.0,"publicationDate":"2020-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45351982","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-06-01DOI: 10.11889/J.0253-3219.2020.HJS.43.060004
毛辉辉, 张丹, 高春天, 吴攀, 刘余, 毕树茂, 米争鹏
[Background]As a new concept reactor,direct circulation CO2 cooled reactor has different system configurations and safety characteristics compared with direct cycle boiling water reactor,indirect cycle helium gas cooled reactor and pressurized water cooled reactor,hence the CO2 cooled reactor has different initial events and safety criteria.The selection of initial events is the foundation for the safety design of reactor whilst the safety criteria provides criterion for whether the safety analysis results meet the safety requirements.[Purpose]This study aims at the initial events and acceptance criteria for the safety design of CO2 cooled reactor.[Method]The main logic diagram analysis method was employed to study the initial events of the direct circulation CO2 cooled reactor.Based on the existing engineering experience of pressurized water reactor,gas cooled reactor,the acceptance criteria for the direct circulation CO2 cooled reactor were proposed according to its own characteristics.[Results&Conclusions]The proposed initial events and acceptance criteria required for safety analysis are applicable to direct circulation reactor,providing a basis for the safety analysis of CO2 cooled nuclear power plant,and gives an important reference for the safety design of direct cycle circulation reactors.
{"title":"Study of initial events and safety criteria for CO2 cooled reactor","authors":"毛辉辉, 张丹, 高春天, 吴攀, 刘余, 毕树茂, 米争鹏","doi":"10.11889/J.0253-3219.2020.HJS.43.060004","DOIUrl":"https://doi.org/10.11889/J.0253-3219.2020.HJS.43.060004","url":null,"abstract":"[Background]As a new concept reactor,direct circulation CO2 cooled reactor has different system configurations and safety characteristics compared with direct cycle boiling water reactor,indirect cycle helium gas cooled reactor and pressurized water cooled reactor,hence the CO2 cooled reactor has different initial events and safety criteria.The selection of initial events is the foundation for the safety design of reactor whilst the safety criteria provides criterion for whether the safety analysis results meet the safety requirements.[Purpose]This study aims at the initial events and acceptance criteria for the safety design of CO2 cooled reactor.[Method]The main logic diagram analysis method was employed to study the initial events of the direct circulation CO2 cooled reactor.Based on the existing engineering experience of pressurized water reactor,gas cooled reactor,the acceptance criteria for the direct circulation CO2 cooled reactor were proposed according to its own characteristics.[Results&Conclusions]The proposed initial events and acceptance criteria required for safety analysis are applicable to direct circulation reactor,providing a basis for the safety analysis of CO2 cooled nuclear power plant,and gives an important reference for the safety design of direct cycle circulation reactors.","PeriodicalId":35563,"journal":{"name":"He Jishu/Nuclear Techniques","volume":"43 1","pages":"21-28"},"PeriodicalIF":0.0,"publicationDate":"2020-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46369886","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2017-01-01DOI: 10.11889/J.0253-3219.2017.HJS.40.100602
Zhang Pan, Xu Chao, Wen Lijing, H. Wenchao, Liu Yu-sheng, Li Cong-xin
{"title":"Numerical simulation of flow and heat transfer characteristic in passive residual heat removal heat exchanger","authors":"Zhang Pan, Xu Chao, Wen Lijing, H. Wenchao, Liu Yu-sheng, Li Cong-xin","doi":"10.11889/J.0253-3219.2017.HJS.40.100602","DOIUrl":"https://doi.org/10.11889/J.0253-3219.2017.HJS.40.100602","url":null,"abstract":"","PeriodicalId":35563,"journal":{"name":"He Jishu/Nuclear Techniques","volume":"40 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2017-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"65748927","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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