Investigation of shielding and dosimetric parameters for neutrons emitted from various types of fission sources

This study evaluates the shielding and dosimetric characteristics of neutrons emitted from various fission sources, utilizing the Monte Carlo simulations. Neutron emissions, crucial in nuclear applications and safety, are analyzed across a spectrum of fissionable isotopes. The work focuses on determining the mean fluence-to-dose conversion coefficients, average total macroscopic cross sections, as well as the flux and dose-equivalent buildup factors for various shielding materials. Results highlight that neutron flux and dose-equivalent buildup factors increase with shield thickness, influenced significantly by the shielding material’s effective mass number. Notably, concrete exhibits the highest buildup factor, contrasting with water, which shows the lowest. Furthermore, polynomial equations parameterized by source and material coefficients facilitate accurate flux and dose-equivalent buildup factor calculations. This study enhances the understanding of neutron shielding effectiveness and dosimetry, which is crucial for optimizing radiation protection measures in various nuclear contexts, from reactor operations to medical applications.