Coupled neutronics, thermal-hydraulics, and fuel performance analysis of dispersion plate-type fuel assembly in a cohesive way

Abstract

The nuclear reactor is a multi-physics and multiscale coupling system. The conditions of high temperature and high neutron flux in the reactor pose a significant challenge to the thermodynamic performance of the fuel. To evaluate the plate-type fuel's performance, a coupling calculation of neutron physics code OpenMC, fuel performance analysis code BEEs-Plates, and nuclear reactor system analysis code NUSAC has been implemented based on the open-source MOOSE platform. A novel and efficient method for mesh grids generation and data mapping between external codes and MOOSE was proposed to ensure accurate data exchange. This method can simplify the mesh grids generation process of external codes and accurately transfer data across different mesh grids. Finally, the coupling calculation of the JRR-3 M research reactor fuel assembly was carried out based on the above coupling code, and the neutronics parameters, thermal–hydraulic parameters, and fuel performance-related parameters of the fuel assembly were analyzed. The calculation results show that the maximum fuel burnup will reach 16.91% FIMA after 240 days. With the increase of fuel burnup, the peak power of the fuel assembly is significantly shifted towards both ends of the height direction. The VonMises stress, volume stress, creep strain, and displacement of the fuel will be considerable. Finally, the impact of the coupling calculation on the accuracy of the computational efficiency is preliminarily investigated.