Design of boron carbide-shielded irradiation site of the outer irradiation site of the Syrian MNSR using MCNP

Abstract

The MCNP4C model for the Syrian Miniature Neutron Source Reactor (MNSR) was enhanced to incorporate a boron carbide-shielded irradiation site in the first outer irradiation site for a permanent epithermal neutron irradiation site by using boron carbide. The concept of re-designing a new MCNP4C model includes the calculation of the excess core reactivity, control rod worth, shutdown margin and thermal neutron flux in the inner and outer irradiation sites before and after shielding with boron carbide, as well as the calculation of the thermal and epithermal neutron fluxes in the first outer irradiation site. Distribution of the thermal neutron flux in the first outer irradiation site has been calculated using the MCNP4C code and experimentally irradiating five copper wires using the outer irradiation capsule. Good agreement was noticed between the calculated and the measured values. To compensate for the reactivity losses due to the neutrons absorption in the cylindrical boron carbide shell a beryllium shim was added to the top tray of the reactor.