Chapters authored
Monte Carlo Simulations of Nuclear Fuel Burnup
In the operation of a nuclear power plant, it is very important to determine the time evolution of material composition and radionuclide inventory during the entire operation of the plant. In the experiments, the Monte Carlo N-Particle eXtended (MCNPX) code was found to be accurate in predicting the uranium fuel depletion, the plutonium production and the buildup of most of the fission products in a nuclear reactor. The goal in this chapter is to analyze the effect of different nuclear fuel grades on the total radioactivity of the reactor core by employing nuclear burnup calculations for the three different fuels: mixed oxide fuel (MOX), uranium oxide fuel (UOX) and commercially enriched uranium (CEU), utilizing simulations with MCNPX code. The calculated results indicate that there is a buildup of plutonium isotopes for UOX and CEU, whereas there is a decline in the plutonium radioisotopes for MOX fuel with burnup time. The study of reactor neutronic parameters showed UOX fuel performs better relative to MOX and CEU. Zircaloy, with low thermal neutron absorption cross-section and high thermal conductivity, produced better results for the effective multiplication factor Keff and hence proved to be a much more effective clad material.
Part of the book: Nuclear Material Performance