Safety-related parameters calculation for OPAL reactor using MCNP v6.1.0 and ENDF/B-VII.1

Abstract

Despite the standard cell-core approach remains as workhorse for the research reactor core’s design and analysis, several specific calculations are regularly done using Monte Carlo codes. This work presents a tailored validation of the well-known MCNP6 code for calculations within compact-core research reactors using parallel-plate with low-enriched uranium fuels, cooled and moderated by light-water and reflected by heavy-water. For such purpose, MCNP v6.1.0 and nuclear data from ENDF/B-VII.1 are here used for the calculation of safety-related parameters of the state-of-the-art OPAL research reactor, providing a detailed comparison with available experimental data. This analysis covers a wide range of parameters for fresh and burned cores, where for the latter case compositions from a deterministic (i.e. cell-core) approach are inserted into MCNP models. As a result, the aptness of the MCNP6 code to reproduce key experimental data from OPAL reactor likewise the expected level of accuracy for analogous applications are here addressed.