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Transport Calculations of photons and neutrons in different shielding materials | ||
| Al-Rafidain University College For Sciences | ||
| Article 1, Volume 0, Issue 19, 2017, Pages 149-156 | ||
| Author | ||
| Talib A. Al-sharify | ||
| 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. | ||
| Keywords | ||
| photons and neutrons | ||
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