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...

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Podrobná bibliografie

Hlavní autoři:	Caroline Mattos Barbosa, Cesar Raitz, Roos Sophia de Freitas Dam, William Luna Salgado, Cesar Marques Salgado, Delson Braz
Médium:	Artigo
Jazyk:	Inglês
Vydáno:	Brazilian Radiation Protection Society (Sociedade Brasileira de Proteção Radiológica, SBPR) 2021-04-01
Edice:	Brazilian Journal of Radiation Sciences
Témata:	mcnp nuclear technique gamma ray.
On-line přístup:	https://bjrs.org.br/revista/index.php/REVISTA/article/view/1568
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