Neutronic Evaluation of MSBR System Using MCNP Code

Neutronic Evaluation of MSBR System Using MCNP Code

The concept of Molten Salt Reactor use Th to breed fissile 233U, where an initial source of fissile material needs to be provided. However, there is no available 233U and so; the fissile fuel supply is one of the unresolved problems. Thus, it is necessary to use existing fissile materials such as 23...

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Bibliographic Details
Main Authors: Clarysson Alberto Mello da Silva, Alana Lima Vieira, Isabella Resende Magalhães, Claubia Pereira
Format: Article
Language:English
Published: Brazilian Radiation Protection Society (Sociedade Brasileira de Proteção Radiológica, SBPR) 2021-07-01
Series:Brazilian Journal of Radiation Sciences
Subjects:
msbr
nuclear fuel cycle
neutronic simulation
mcnpx
Online Access:https://bjrs.org.br/revista/index.php/REVISTA/article/view/1261
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