A comparative study of the neutron flux spectra in the MNSR irradiation sites for the HEU and LEU cores using the MCNP4C code

A comparative study for fuel conversion from the HEU to LEU in the Miniature Neutron Source Reactor (MNSR) has been performed in this paper using the MCNP4C code. The neutron energy and lethargy flux spectra in the first inner and outer irradiation sites of the MNSR reactor for the existing HEU fuel (UAl₄–Al, 90% enriched) and the potential LEU fuels (U₃Si₂–Al, U₃Si–Al, U9Mo–Al, 19.75% enriched and UO_2, 12.6% enriched) were investigated using the MCNP4C code. The neutron energy flux spectra for each group was calculated by dividing the neutron flux by the width of each energy group. The neutron flux spectra per unit lethargy was calculated by multiplying the neutron energy flux spectra for each energy group by the average energy of each group. The thermal neutron flux was calculated by summing the neutron fluxes from 0.0 to 0.625 eV, the fast neutron flux was calculated by summing the neutron fluxes from 0.5 MeV to 10 MeV for the existing HEU and potential LEU fuels. Good agreements have been noticed between the flux spectra for the potential LEU fuels and the existing HEU fuels with maximum relative differences less than 10% and 8% in the inner and outer irradiation sites. - Highlights: • A comparative study for fuel conversion from the HEU to LEU fuels has been conducted. • The neutron energy and lethargy flux spectra of the MNSR reactor were investigated . • The thermal neutron flux was calculated by summing the neutron fluxes from 0.0 to 0.625 eV, • Results showed good agreements between the flux spectra for LEU fuels and the existing HEU fuels