Analysis of Neutronic Parameters for Supercell of CANDU Reactor Using MCNPX Code

MCNPX code has been used for modeling and simulation of a supercell of CANDU Fuel, the supercell consists of two fuel bundle and adjuster rod. The fuel bundle are burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor of the bundle is calculated during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the bundle. The worth of the adjuster rod is calculated. Comparison of multiplication factor and worth of the adjuster rod with the previous published references showed good agreement.