Deterministic evaluation of safety parameters and neutron flux spectra in the MNSR research reactor using DRAGON-4 code

Comparative studies for conversion of the fuel from high enriched uranium (HEU) to low enriched uranium (LEU) in the reactor (MNSR) were performed using DRAGON code deterministic. In this work the HEU fuel (UAl 4 -Al, 90% enriched with Al-clad) and LEU (12.6% UO 2 enriched with zircaloy-4 alloy clad) cores were analyzed. This model was utilized in this work to calculate the neutron energy flux spectrum in the first inner and outer irradiation sites of the Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections has been evaluated from ENDF/B-VII library. The neutron fluxes has been calculated using 69 energy groups. The neutron energy flux for every group has been calculated dividing the neutron flux by the width of every energy group. As well, the neutron flux spectrum per unit lethargy has been calculated by multiplying the neutron energy flux spectra for every energy group by the average energy of every group. The thermal neutron flux was calculated by summing the neutron fluxes from (0.0–0.625) eV, the fast neutron flux was calculated by summing the neutron fluxes from (0.5–10) MeV for the existing HEU and potential LEU fuels. In this work, the effective multiplication factor ( k e f f ), excess reactivity ( P e x ), Shut- Down Margin (SDM), Control Rod Worth (CRW), Safety Reactivity Factor (SRF) and neutron energy flux spectrum for every group was calculated by dividing the neutron flux by the width of every energy group. In final, the results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels have been good agreements with the HEU results, compare with Monte Carlo MCNP4C code measured. Therefore, the LEU fuel has been validated as an appropriate option for fuel conversion of the MNSR in the future.