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

Published Online: Dec 30, 2014

Page range: 129 - 136

Received: Aug 27, 2013

Accepted: Aug 08, 2014

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

Keywords

ThO2, neutronic parameters, fuel burn-up, 233U, 235U, 239Pu fissile material

© by Zohreh Gholamzadeh

This work is licensed under the Creative Commons Attribution-NonCommercial-NoDerivatives 3.0 License.