Calculation and comparison of the neutron energy flux spectra in the Syrian MNSR irradiation sites using the MCNP-4C code

A 3-D neutronic model for the Syrian Miniature Neutron Source Reactor (MNSR) was developed earlier to analyze the reactor neutronic using the MCNP-4C code. This model was used in this paper to calculate the neutron energy flux spectra in the five inner and five outer irradiation sites of the MNSR reactor. The continuous energy neutron cross sections were evaluated from ENDF/B-VI library. The neutron fluxes were calculated using 69 energy groups. The neutron energy flux for each group was calculated dividing the neutron flux by the width of each energy group. The calculations showed that the distributions of the neutron energy flux spectra in the five inner irradiation sites were in good agreements. These results were noticed in the reactor outer irradiation sites as well. Using the neutron flux spectrum in the first inner irradiation site, the thermal (0.0–0.625 eV) and fast neutron fluxes (0.5–10 MeV) were calculated and found 1.035 × 1012 and 3.022 × 1011 cm−2 s−1 respectively. The measured fluxes in the first inner irradiation site were found previously to be (0.97 ± 0.06) × 1012 and (2.89 ± 0.06) × 1011 cm−2 s−1 respectively. Good agreements were noticed between the calculated and the measured results. These agreements verify the calculated neutron flux spectra in the reactor sites.