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

doi:10.3389/fenrg.2021.705944

	Lead-Cooled Fast Reactor Annular UN Fuel Design and Development of Performance Analysis Program
	Yuan, He 1; Wang, Guan 2,5; Yu, Rui 2,5; Tao, Yujie 2,5; Wang, Zhaohao 1; Guo, Shaoqiang 1; Liu, Wenbo 1; Yun, Di 1,4; Gu, Long 2,3,5
刊名	FRONTIERS IN ENERGY RESEARCH
	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
DOI	10.3389/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.
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关键词	MATERIAL PROPERTY CORRELATIONS ; URANIUM MONONITRIDE
WOS研究方向	Energy & Fuels
语种	英语
出版者	FRONTIERS MEDIA SA
WOS记录号	WOS:000683310100001
资助机构	State Key Research and Development Program of China ; National Natural Science Foundation of China ; Shanghai Economic and Information Technology Commission
内容类型	期刊论文
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.