Lead-Cooled Fast Reactor Annular UN Fuel Design and Development of Performance Analysis Program - Yuan, He; Wang, Guan; Yu, Rui; Tao, Yujie; Wang, Zhaohao; Guo, Shaoqiang; Liu, Wenbo; Yun, Di; Gu, Long (Frontiers in Energy Research, 2021)
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.