Processing of JEFF-3.3 and ENDF/B-VIII.0 and testing with critical benchmark experiments and TRIGA Mark II research reactor using MCNPX

Abstract

A comparative study has been performed with the latest evaluated nuclear data libraries JEFF-3.3 and ENDF/B-VIII.0. The study has been conducted through the benchmark calculations for 120 criticality problems and the TRIGA Mark II research reactor with the libraries processed using NJOY21 for MCNPX Monte Carlo transport code.The criticality benchmarks assemblies, taken from the ICSBEP benchmark, cover Uranium (highly enriched uranium, intermediate enriched uranium, low enriched uranium, and233U) and Plutonium fuel systems in a various metal forms, and under a various spectral conditions. The Moroccan TRIGA Mark II research reactor calculation is used to look into the predictive capability of those nuclear data libraries and then to compare the accuracy of the predicted results with the experimental data published elsewhere.Actually, the purpose of this study is to investigate some neutronic and kinetic parameters of those benchmarks for both libraries. The former consist of effective multiplication factor, heat distribution, neutron flux distribution, effective delayed neutron fraction (βeff), prompt removal lifetime (τr) and the mean neutron generation time (Λ). The results show that the calculated effective multiplication factor, heat distribution, neutron flux distribution, and the kinetic parameters are in good agreement with references. However, it is found that the computed values are strongly depending on the nuclear data set used in calculations.