Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe op...

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Bibliographic Details
Main Authors: Gholamzadeh Zohreh, Hossein Feghhi Seyed Amir, Soltani Leila, Rezazadeh Marzieh, Tenreiro Claudio, Joharifard Mahdi
Format: Article
Language:English
Published: Sciendo 2014-12-01
Series:Nukleonika
Subjects:
tho2
neutronic parameters
fuel burn-up
233u
235u
239pu fissile material
Online Access:https://doi.org/10.2478/nuka-2014-0017

Internet

https://doi.org/10.2478/nuka-2014-0017

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