Analysis of safety parameters for the MINERVE reactor

The present study aims to show the effect of different cross section libraries on MINERVE reactor safety parameters. The MCNP5 calculation model of the MINERVE reactor facility is used to determine core and safety parameters such as axial and radial fission rate distributions, control rod worth and spectral indices. Different neutron spectra were achieved by changing the experimental lattice within the MINERVE reactor. MINERVE provides a large experimental basis for the improvement of the cross section databases. The current study calculates these parameters by MCNP5 code using three different cross section libraries (the continuous energy cross sections of the ENDFB-VI, T16-2003, and ENDF/B-VII.1 libraries). Neutronic calculations were performed for R1UO₂, R1MOX, R2UO₂, and R2MOX core configurations, representative of a LWR loaded with UO₂, mixed oxide matrix, over moderated UO₂ and over moderated MOX, respectively. The study aims to determine the most suitable cross section library to be used with MCNP5 reactor calculation code for light water reactor fuel lattice. The MCNP5 results were compared with the experimental results.