Determination of Gamma-Ray Shielding Parameters for Concretes and Dosimeters Using MCNPX

Huseyin Ozan  Tekin; V. P. Singh

doi:10.15415/jnp.2020.81009

Authors

Huseyin Ozan Tekin Department of Medical Diagnostic Imaging, University of Sharjah, P.O. Box-27272, Sharjah, United Arab Emirates
V. P. Singh Department of Physics, Karnatak University, Dharwad, Karnataka-580003, India

DOI:

https://doi.org/10.15415/jnp.2020.81009

Keywords:

Concrete, Dosimeter, Attenuation Coefficient, Gamma radiations, MCNPX

Abstract

Gamma-ray shielding parameter for some concretes and dosimeters having large scale applications in radiological protection are presented using MCNPX (version 2.4.0) at different energies. The MCNPX results are compared with experimental, MCNP and XCOM data, and good agreement is being noted. Present study indicates that MCNPX simulation method is suitable and reliable simulation tool to be used as an alternative method for the investigation of gamma-ray interaction. The present geometry can be used as standard geometry for MCNPX simulation for low- as well as high-Z materials.

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