Transport Calculations of photons and neutrons in different shielding materials

Abstract

Transport calculations of photon and neutron fluxes and doses were made. The Monte Carlo multigroup experimental transport code MORSE-CG [1] was used for these calculations with its modified version [2].Distributions of gamma ray fluxes through the shield of a pressurized water reactor were obtained. The shield was represented by layers of water, steel and lead; it was a semi realistic representation of a working reactor.A simulation was made for the total dose that could be received by a person living near a very active radiating source of neutrons or photons. The obtained results were compared with experimental referenced data which shows a very good agreement.