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: | Caroline Mattos Barbosa, Cesar Raitz, Roos Sophia de Freitas Dam, William Luna Salgado, Cesar Marques Salgado, Delson Braz |
---|---|
Format: | Article |
Language: | English |
Published: |
Brazilian Radiation Protection Society (Sociedade Brasileira de Proteção Radiológica, SBPR)
2021-04-01
|
Series: | Brazilian Journal of Radiation Sciences |
Subjects: | |
Online Access: | https://bjrs.org.br/revista/index.php/REVISTA/article/view/1568 |
Similar Items
-
Volume fractions calculation of a biphasic system on cylindrical tube using gamma ray and MCNP6 code
by: William Luna Salgado, et al.
Published: (2021-04-01) -
Study of radioactive particle tracking using MCNP-X code and artificial neural network
by: Roos Sophia de Freitas Dam, et al.
Published: (2021-07-01) -
Mass Attenuation Coefficients of Human Body Organs using MCNPX Monte Carlo Code
by: Huseyin Tekin, et al.
Published: (2017-12-01) -
Calculation of Mass Attenuation Coefficients for Pedicle Screw by Theoretical and Monte Carlo Simulation Methods
by: Hasan Özdoğan, et al.
Published: (2021-11-01) -
Calculation of scale thickness in oil pipelines using transmission gamma
by: Tâmara Porfíro Teixeira, et al.
Published: (2021-02-01)