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Lead-Cooled Fast Reactor Annular UN Fuel Design and Development of Performance Analysis Program

文献类型：期刊论文

作者Yuan, He1; Wang, Guan2,5; Yu, Rui2,5; Tao, Yujie2,5; Wang, Zhaohao1; Guo, Shaoqiang1; Liu, Wenbo1; Yun, Di1,4; Gu, Long2,3,5
刊名FRONTIERS IN ENERGY RESEARCH
出版日期2021-07-27
卷号9页码:9
关键词UN fuel annular fuel fuel performance analysis COMSOL fast reactors
ISSN号2296-598X
DOI10.3389/fenrg.2021.705944
通讯作者Yun, Di(diyun1979@xjtu.edu.cn) ; Gu, Long(gulong@impcas.ac.cn)
英文摘要A kind of annular uranium nitride (UN) fuel suitable for lead-cooled fast reactor applications has been designed in this study. The design is directly targeting two main issues of UN fuel: severe swelling and thermal decomposition of UN fuel at high temperatures. A performance analysis program based on FORTRAN programming language has been developed for UN fuel in fast reactors. The program contains heat transfer, fuel stress-strain analysis, cladding stress-strain analysis, fission gas release and fuel-cladding mechanical interaction (FCMI) modules, etc. Extensive code verification has been performed by comparing simulation results obtained with the code and those obtained via the COMSOL Multiphysics platform. Preliminary code validation has been conducted as well by comparing code simulation results with experimental data. The results showed that this program could predict the fuel temperature, stress-strain, and displacement of UN fuel during reactor operation with a reasonable accuracy.
WOS关键词MATERIAL PROPERTY CORRELATIONS ; URANIUM MONONITRIDE
资助项目State Key Research and Development Program of China[2020YFB1902100] ; National Natural Science Foundation of China[11675126] ; National Natural Science Foundation of China[11705255] ; Shanghai Economic and Information Technology Commission[GYQJ-2018-2-02]
WOS研究方向Energy & Fuels
语种英语
WOS记录号WOS:000683310100001
出版者FRONTIERS MEDIA SA
资助机构State Key Research and Development Program of China ; National Natural Science Foundation of China ; Shanghai Economic and Information Technology Commission
源URL[http://119.78.100.186/handle/113462/136638]  
专题中国科学院近代物理研究所
通讯作者Yun, Di; Gu, Long
作者单位1.Xi An Jiao Tong Univ, Sch Nucl Sci & Technol, Xian, Peoples R China
2.Univ Chinese Acad Sci, Sch Nucl Sci & Technol, Beijing, Peoples R China
3.Lanzhou Univ, Sch Nucl Sci & Technol, Lanzhou, Peoples R China
4.Xi An Jiao Tong Univ, State Key Lab Multiphase Flow, Xian, Peoples R China
5.Chinese Acad Sci, Inst Modern Phys, Lanzhou, Peoples R China
推荐引用方式
GB/T 7714		 Yuan, He,Wang, Guan,Yu, Rui,et al. Lead-Cooled Fast Reactor Annular UN Fuel Design and Development of Performance Analysis Program[J]. FRONTIERS IN ENERGY RESEARCH,2021,9:9.
APA Yuan, He.,Wang, Guan.,Yu, Rui.,Tao, Yujie.,Wang, Zhaohao.,...&Gu, Long.(2021).Lead-Cooled Fast Reactor Annular UN Fuel Design and Development of Performance Analysis Program.FRONTIERS IN ENERGY RESEARCH,9,9.
MLA Yuan, He,et al."Lead-Cooled Fast Reactor Annular UN Fuel Design and Development of Performance Analysis Program".FRONTIERS IN ENERGY RESEARCH 9(2021):9.

入库方式： OAI收割

来源：近代物理研究所

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