USE OF MONTE CARLO SIMULATIONS FOR OPTIMAL GEOMETRY STUDY IN CALCULATION OF ATTENUATION COEFFICIENT FOR ELEMENT, COMPOUND AND MIXTURE

Abstract

MCNP is a code extensively used to simulate experiments involving transport of radiation using the Monte Carlo method. This code allows the study of different geometries, materials, and radiation types (e.g. gamma, neutron, and electron), enabling the building of approximate models before the experimental implementation. The objective of this study is to develop an optimal geometry for the calculation of the mass attenuation coefficient for different materials using the MCNP code. Several measurement geometries were tested with different radiation energies, and the best results were obtained using lead collimators on both detector and radiation source. The considered geometries were isotropic source without any collimation, isotropic source with detector and/or source collimation, and a point source collimated into a cone of directions. The last case was proposed as a replacement for the computationally time expensive simulation of the two-collimator geometry. The energies 59.54 keV, 81 keV, 356 keV, and 662 keV were used to model 241Am, 133Ba, and 137Cs radiation sources, respectively. The materials were, NaI for the detector, aluminum, water, and sea water (3.5% NaCl) for the target sample, and lead for the collimators. The values of mass attenuation coefficient obtained from the simulations were compared with the theoretical NIST XCOM values for validation of the geometries.

Keywords

• MCNP
• Nuclear Technique
• gamma ray.