Verifying Dose Rates Outside of Polytechnique Montréal Subcritical Assembly Using MCNP5.1

Abstract

The Monte Carlo Neutral Particle Code (MCNP) was used to calculate the dose at a distance of one meter from a subcritical assembly. The assembly is a roughly 2m x 2m x 2m parallelepiped graphite pile. The fuel is natural uranium metal, and the cladding is aluminum. A neutron source of americium-beryllium was used. In order to verify the code, a previous experimental laboratory conducted at Polytechnique Montréal was replicated. Once validated, the dose values were found to be below 2.5 μSv/h as per the safety requirements of Ontario Tech University [1].