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

He Yuan; Guan Wang; Guan Wang; Rui Yu; Rui Yu; Yujie Tao; Yujie Tao; Zhaohao Wang; Shaoqiang Guo; Wenbo Liu; Di Yun; Di Yun; Long Gu; Long Gu; Long Gu

doi:10.3389/fenrg.2021.705944

Frontiers in Energy Research (Jul 2021)

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

He Yuan,
Guan Wang,
Guan Wang,
Rui Yu,
Rui Yu,
Yujie Tao,
Yujie Tao,
Zhaohao Wang,
Shaoqiang Guo,
Wenbo Liu,
Di Yun,
Di Yun,
Long Gu,
Long Gu,
Long Gu

Affiliations

He Yuan: School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, China
Guan Wang: Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou, China
Guan Wang: School of Nuclear Science and Technology, University of Chinese Academy of Sciences, Beijing, China
Rui Yu: Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou, China
Rui Yu: School of Nuclear Science and Technology, University of Chinese Academy of Sciences, Beijing, China
Yujie Tao: Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou, China
Yujie Tao: School of Nuclear Science and Technology, University of Chinese Academy of Sciences, Beijing, China
Zhaohao Wang: School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, China
Shaoqiang Guo: School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, China
Wenbo Liu: School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, China
Di Yun: School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, China
Di Yun: State Key Laboratory of Multi-phase Flow, Xi’an Jiaotong University, Xi’an, China
Long Gu: Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou, China
Long Gu: School of Nuclear Science and Technology, University of Chinese Academy of Sciences, Beijing, China
Long Gu: School of Nuclear Science and Technology, Lanzhou University, Lanzhou, China

DOI: https://doi.org/10.3389/fenrg.2021.705944
Journal volume & issue: Vol. 9

Abstract

Read online

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.

Published in Frontiers in Energy Research

ISSN: 2296-598X (Online)
Publisher: Frontiers Media S.A.
Country of publisher: Switzerland
LCC subjects: General Works
Website: https://www.frontiersin.org/journals/energy-research

About the journal

Abstract

Keywords