The effect of homogenization on fuel burnup for pressure tube of CANDU-6 reactor

الملخص EN

MCNPX code has been used for modeling and simulating fuel bundle of CANDU (Canada Deuterium-Uranium) reactor.

Two-design models are used: both homogenous and heterogeneous pressure tube of CANDU reactor.

The fuel is burnt in normal operation conditions of CANDU reactors.

Natural uranium fuel is used in the model.

Neutronic parameters of the two models are compared with each other.

The multiplication factor for both homogeneous and heterogeneous models is calculated and compared during fuel burnup.

The concentration of both uranium and plutonium isotopes are analyzed in the model.

The result of the present models was compared with published references to assess the homogeneous model and good agreement was found.