Compact shielding design for 740 GBq 241Am-Be neutron source transport container

Abstract

For transport and storage of neutron sources, shielding materials like polymers and multi layer shields are used. A compact shielding design for 740 GBq portable 241Am-Be neutron source transport container using Monte Carlo technique is presented. In the design, a polymer host material with lead and natural boron (NB) as fillers is chosen as the shielding medium. Monte Carlo simulation (MCS) is performed for optimizing the quantity of fillers in the polymer. The total dose rate (TDR) on the surface of the container due to neutrons and gamma rays emanating from source as well as polymer shield, is considered as the optimization criterion. MCS studies indicate that a polymer material with 5% of lead and 1% of NB as fillers is observed to give optimized composition of polymer (OCP). Statistical factorial design analysis (FDA) technique is employed for the first time in the shielding design to investigate the impact of fillers in the polymer. FDA studies reveal that the quantity of lead has more significant impact compared to NB in the polymer. MCS results are validated by carrying out shielding experiments with high density polyethylene (HDPE) and composite polymer (CP) based containers. The computed and experimental dose rates are observed to be within ±12%. A shielding container made of OCP for the transport of 740 GBq 241Am-Be source provides 25% reduction in the volume as compared to HDPE and CP. The mass of the OCP container is lower by 18% and 26% compared to the containers made of HDPE and CP respectively. The surface dose rate of the OCP container adheres to the IAEA transport regulations.