Comparison of Measured Radiation Exposure and Dose Rates with Analytical and Computational Methods

Abstract

Proper selection and utilization of survey instruments are important to fulfilling radiation survey needs. Ion chambers are typically used to survey β and γ-ray radiation, whereas REM balls containing helium-3 detectors are used to survey neutrons. While analytical calculations provide good estimates of radiation doses, it becomes challenging to estimate doses in the presence of shielding material(s) under complex geometries. γ doses are underestimated without accounting for buildup in the shielding equation. Commercially available software such as MicroShield can be used to calculate final γ radiation exposure, which relies on mathematical techniques to account for the radiation buildup in complex geometries. However, MicroShield is limited to assessing γ exposure and doses. Both γ and neutron radiation doses can be modeled using Monte Carlo particle transport methods in the computer code MCNP. Neutron and γ radiation doses from various spontaneous fission sources can also be modeled in the MCNP. However, this is dependent upon implementing proper fission models in the MCNP to calculate neutron and γ radiation doses. MicroShield in the default configuration cannot calculate γ radiation doses from spontaneous fission sources. However, the γ radiation dose from a spontaneous fission source can be correctly calculated in MicroShield if a γ spectrum is provided. MicroShield, MCNP, and analytical calculations show good agreement with measurements for γ sources used in the paper. On the other hand, MCNP and analytical calculations show good agreement with measurements for the spontaneous fission source used in the paper.