Thermal Neutron Characterization and Dose Modeling of a (PuBe)-Pu-239 Alpha-Neutron Source

摘要

Determination of neutron dose can be challenging and requires knowledge of neutron flux as a function of energy. The goal of this project was to characterize the thermal neutron flux of a 37 GBq (PuBe)-Pu-239 alpha-neutron source and model the associated neutron dose using version MCNPX of the Monte-Carlo N-Particle transport codes. The (PuBe)-Pu-239 source was placed in a neutron howitzer, and foil activation (dysprosium foils with and without cadmium covers) was used at various distances to determine thermal neutron flux, which was then used to verify the MCNPX model representing the system. The model was then adapted for dosimetric modeling to enable future neutron dose-response studies.