Nuclear-thermal-mechanical performance analysis of MOX fuel rod for lead-cooled fast reactor

• · Realized the development of three-dimensional fuel performance program for lead-cooled fast reactors, covering key physical phenomena such as interstitial heat transfer, fission gas release, nuclide migration, and irradiation strain. • ·A new coupling calculation method is proposed to couple the neutronics calculation software OpenMC to the COMSOL platform to study the nuclear thermal coupling of lead-cooled fast reactor fuel. • ·The effect of the burnup process on the physical properties of MOX fuel for lead-cooled fast reactors was analyzed, and the changes in burnup performance before and after coupling the burnup module were compared. Due to the harsh operating conditions of Lead-Cooled Fast Reactors (LFRs)—including elevated temperatures and intense neutron irradiation—the fuel and cladding materials exhibit complex thermomechanical and irradiation-induced behaviors that evolve throughout the burnup process. To address these challenges, a three-dimensional fuel performance analysis code was developed based on the COMSOL Multiphysics platform. The code simulates key physical phenomena such as heat conduction, mechanical deformation, fission gas release, and nuclide migration. Furthermore, it is coupled with the OpenMC burnup module to incorporate the effects of fuel depletion on thermophysical and mechanical properties. The results indicate that burnup significantly alters the behavior of MOX fuel, with the maximum fuel temperature reaching approximately 1480 °C, fission gas plenum pressure rising to 1.84 MPa, and peak cladding stress reaching 55.5 MPa. The developed tool provides a robust framework for fuel design optimization and safety assessment of LFR systems.