Burn up analysis for fuel assembly unit in a pressurized heavy water candu reactor

الملخص EN

MCNPX code has been used for modeling and simulation of an assembly of CANDU Fuel bundle.

The assembly is composed of a heterogeneous lattice of 37- element natural Uranium fuel, heavy water moderator and coolant.

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

The effective multiplication factor (Keff) of the bundle is calculated as a function of fuel burnup.

The flux and power distribution are determined.

Comparing the concentrations of both Uranium and Plutonium isotopes are analyzed in the bundle.

The results of the present model with the results of a benchmark problem, a good agreement was found.