USE OF MONTE CARLO SIMULATIONS FOR OPTIMAL GEOMETRY STUDY IN CALCULATION OF ATTENUATION COEFFICIENT FOR ELEMENT, COMPOUND AND MIXTURE
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 experi...
Main Authors: | , , , , , |
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Format: | Article |
Language: | English |
Published: |
Brazilian Radiation Protection Society (Sociedade Brasileira de Proteção Radiológica, SBPR)
2021-04-01
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Series: | Brazilian Journal of Radiation Sciences |
Subjects: | |
Online Access: | https://bjrs.org.br/revista/index.php/REVISTA/article/view/1568 |