Experimental validation of ambient dose equivalent rate calculations for a ²⁴¹Am-Be neutron source using OpenMC code

This work presents the experimental validation of ambient neutron dose equivalent rate calculations for a ²⁴¹Am-Be neutron source using Monte Carlo simulations performed with OpenMC. The model reproduces the cylindrical geometry of the source and accounts for its finite spatial extent by uniformly sampling neutron emission points throughout the active volume. Since the actual distribution of the ²⁴¹AmO₂-Be powder within the active region is not available, a uniform spatial emission distribution was assumed. The ²⁴¹Am-Be neutron emission spectrum was imposed as the fixed-source energy distribution. The structural components supporting the source were also modeled to account for neutron scattering contributions, while the neutron detector was represented by a spherical scoring region with dimensions equivalent to those of the calibrated detector used in the experimental measurements. The simulated ambient neutron dose equivalent rates were compared with experimental measurements obtained at different source-to-detector distances and with analytical estimates based on ICRP/ICRU fluence-to-H*₍₁₀₎ conversion coefficients. The results show good agreement among the Monte Carlo calculations and experimental measurements, with the agreement improving at larger source-to-detector distances. The results support the use of OpenMC for neutron dosimetry and radiation protection applications involving isotopic neutron sources, while highlighting the importance of accurately representing the finite source geometry and experimental configuration, particularly in the near field.