Optimizing neutron shielding: A specialised container approach for Disused Sealed Radioactive Sources and orphan neutron emitters

Abstract

Shielding Disused Sealed Radioactive Sources (DSRSs) and orphan neutron sources, such as 252 Cf, pose challenges due to the high penetration of neutrons, induced radioactivity in shielding materials, and the limited availability of compact, effective shielding materials. This study addresses these challenges by proposing a specialized container for the safe transport, conditioning, and disposal of neutron-emitting DSRSs and orphan sources. The computational tool that was used in this study is the Monte Carlo N-Particle (MCNP). The investigation begins with an examination of moderation materials, including high-density polyethylene (HDPE), paraffin, and water-extended polyester (WEP). Subsequently, four concrete types, namely Ordinary, Barite, Portland, and Serpentine Concrete, are scrutinized. Employing a comprehensive methodology, the design of the container is optimized to effectively moderate and absorb neutron emissions. The findings of this study demonstrate that the designed compact container can safely handle activities exceeding those studied, effectively managing up to 10 μg (about 0.2 GBq) 252 Cf without exceeding 25 μSv/h at the outer container contact. This suggests a promising solution for the secure management of high-activity neutron-emitting sources. • A 200 L container was designed to safely transport and dispose of neutron-emitting sources like 252 Cf. • Four types of concrete were evaluated for neutron and gamma radiation shielding. • The optimized design features 6 cm of HDPE, 1 cm of lead, and 21.80 cm of serpentine concrete. • The container maintains a radiation dose of 25 μSv/h outside the shielding, ensuring safety compliance.