Multiphysics coupled modeling of light water reactor fuel performance

A fuel performance code for light water reactors called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed. The CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into the COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO2 fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. Additional comparisons of beryllium doped fuel (UO2-10%volBeO) with silicon carbide, instead of Zircaloy as cladding, also indicate good agreement. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide (UO2), composite (UO2-10%volBeO), silicide (U3Si2) and mixed oxide ((Th0.9,U0.1)O2) fuel types under normal operation conditions. Compared to UO2, it was found that the UO2-10%volBeO fuel experiences lower temperatures and fission gas release while producing similar cladding strain. The U3Si2 fuel has the earliest gap closure and induces the highest cladding hoop stress. Finally, the (Th0.9,U0.1)O2 fuel is predicted to produce the lowest fission gas release and a lower fuel centerline temperature when compared with the UO2 fuel. These tests demonstrate that CAMPUS (using the COMSOL platform) is a practical tool for modeling LWR fuel performance.