Power Coefficient Calculation of a CANDU Reactor

Abstract

A coupling calculation between the neutronic and thermal-hydraulic codes has been established for the estimation of the power coefficient using the lattice parameters generated by WIMS-AECL code. In order to utilize the existing capability of RFSP code, a few modifications were made to the RFSP and an interface program between RFSP and NUCIRC was written. The power coefficients were calculated for natural uranium and DUPIC fuel cores. The simulation has shown that the power coefficient of the time-average DUPIC fuel core is more negative compared with that of the natural uranium core, which could be attributed to more fuel temperature and less coolant density feedback effects. However, this study has also shown that the fuel temperature feedback effect predicted by lattice codes needs to be validated, especially for the irradiated natural uranium fuel.