Neutron shielding design for 241Am–Be neutron source considering different sites to achieve maximum thermal and fast neutron flux using MCNPX code

• A 241Am–Be neutron source shielding system was optimized by simulating using MCNPX code.
• The system was performed for boron compositions due to producing lower activated elements.
• A compact shield was designed in terms of both moderating power and absorbing ability.
• The arrangement was consistent with safety requirements and cost limitations.
• Proper sites were determined in order to achieve maximum thermal and fast neutron flux.

Shielding design of an isotropic 241Am–Be neutron source is investigated by Monte Carlo simulation, using MCNPX code. The 241Am–Be is an intense neutron emitter that is readily encapsulated in compact, portable, sealed sources. In the present work, a compact shield is designed with considering different materials in terms of both moderating power and absorbing ability. This arrangement is consistent with safety requirements, cost limitations and material availability. After optimizing the moderator thickness by MCNP code, different materials for attenuating neutrons are examined. Then the moderator is fixed and the best shield configuration is chosen to minimize the equivalent dose outside the shield. For this purpose, MCNPX flux to dose conversion factors reused. Finally, proper sites are determined in order to achieve maximum thermal and fast neutron flux. This configuration enables us to use neutron flux of sites with different energy ranges for irradiating samples with exposing personnel under acceptable radiation level.