S.P. Martinson and S.S. Chirayath, “Monte Carlo neutronics benchmarks on nuclear fuel depletion: A review”, Annals of Nuclear Energy, 161 (2021).
Monte Carlo (MC) neutronics codes are used widely for academic and industrial needs. Several schemes of coupling MC neutronics code with isotope generation and depletion code exist, which are used for performing nuclear fuel depletion simulations. These simulations can estimate the inventory of isotopes in neutron irradiated nuclear reactor fuel. However, the accuracy of these simulations shall be validated through experiments. MC codes are seldom validated by isotopic benchmarks compared to criticality benchmarks. This work compiles and analyzes the fuel depletion benchmarks and validations used to analyze the performance of MC-based fuel depletion neutronics codes. Analyses of these benchmarks and validations showed that the computed concentrations of 133Cs, 135Cs, 137Cs, 148Nd, 239Pu, 240Pu, and 241Pu in the irradiated fuel by the depletion codes agreed with the measured values within 10% error. However, the computed concentrations of 125Sb, 242Cm, 243Cm, 244Cm, 245Cm, and 246Cm had errors more than 15% compared to the measured values. Ventina depletion code showed the most accurate predictions for the greatest number of isotope concentrations compared to ORIGEN2 and CINDER90.