Comparison of Slightly Enriched Uranium Fuel and MOX Fuel in CANDU-6 Reactor

Abstract

A comparison study has been performed between natural uranium (NU), 0.9% and 1.2% slightly enriched uranium (SEU), and mixed-oxide (MOX) fuels in a CANDU 6 reactor for various axial refueling schemes with and without the adjuster rods present in the core. The few group cross section databases including local parameter effects are generated by the multi-group transport code DRAGON using a Winfiith WIMS 69-group library. The 3D CANDU fuel management optimization code OPTEX-4 is then used to compare optimized time-average equilibrium core performance. Finally instantaneous calculations are performed by the 3D diffusion code DONJON, from which both the channel power peaking factor (CPPF) and local parameters effect were estimated.