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.
Highlights
Nuclear techniques are applied in different areas of industry, medicine, and environmental control
MCNP Monte Carlo N-Particle is a code developed based upon the statistical Monte Carlo method, the MCNP simulates the transport of radiation and particles, neutrons, photons, and electrons, and the processes of interaction of radiation with matter
MCNP allows the user to develop the model of an experiment and change the parameters to reach the ideal model desired [2]
Summary
Nuclear techniques are applied in different areas of industry, medicine, and environmental control. It is advantageous to make use of these techniques for they are non-invasive and offer a variety of analysis methods for each problem. The interaction of radiation with materials, with different composition and densities, can be simulated with the Monte Carlo N-Particle (MCNP) code, and the simulation results can be used to calculate the mass attenuation coefficient of those materials. MCNP Monte Carlo N-Particle is a code developed based upon the statistical Monte Carlo method, the MCNP simulates the transport of radiation and particles, neutrons, photons, and electrons, and the processes of interaction of radiation with matter. The applications are in several areas that use nuclear techniques, such as radiological protection, medical physics, nuclear safety and criticality, analysis, and design of detectors, among others [1]. MCNP allows the user to develop the model of an experiment and change the parameters to reach the ideal model desired [2]
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