A Methodology for Experimental and 3-D Computational Radiation Transport Assessments of Pu-Be Neutron Sources

Abstract

Plutonium-beryllium (Pu-Be) sources can be used as didactic source materials for special nuclear materials (SNM) detection evaluation protocols. Since limited specific information exists for many of the Pu-Be sources currently in service, before using a Pu-Be source for field studies, the leakage radiation of neutrons and gamma rays from the source must be fully assessed. Most Pu-Be sources have an outer stainless steel jacket and an inner tantalum jacket, with the Pu-Be homogeneously distributed throughout the inner jacket. To fully characterize the net leakage terms from our Pu-Be source, we applied three-dimensional radiation transport computations, including Monte Carlo (MCNP5) and deterministic (PENTRAN) methodologies. The transport model for our Pu-Be capsule is based on limited schematic and technical data. To define the decay history and resulting source spectrum, exothermic [alpha-neutron (α,n)] reactions are modeled using OrigenArp in the SCALE5 package. For transport modeling purposes, the intermetallic Pu-Be compound was treated as an intimate mixture of plutonium and beryllium, based on the manufacturer’s mass specifications. The net capsule leakage was derived using transport computations, and an iterative estimation of plutonium age was performed. Computational results for net leakage are in agreement with the manufacturer’s specification of neutron yield and dose rate. We also combined computational results with experimental measurement data to fully validate our computational methods. We have successfully achieved agreement between computational and experimental data for our Pu-Be source leakage, and we are using the results at the Florida Institute of Nuclear Detection and Security to evaluate a prototype SNM neutron detector array for parcel screening and national security applications.