Numerical calculation of fuel burn-up rate in a cylindrical nuclear reactor

The prediction of nuclear reactor fuel burn-up rates throughout a reactors lifetime is an important problem in reactor core design. In this study, we present a novel algorithm called BuCal code. This code was used to perform burn-up analysis for a pressurized water reactor fuel with UO₂ whose concentration is 19.5% enriched. Simulation results indicate that the total estimate of ²³⁵U consumption in 225 days with high neutron fluence is approximately 99.9% of the initial value. The study further showed that the microscopic absorption and fission cross sections decreases with increasing temperature and the concentrations of ²³⁵U and ²³⁸U decreases as the numbers of days increases while ²³⁶U build-up as the number of days increases. (author)