Title of article
Monte Carlo transport and burnup calculation
Author/Authors
Deng Li، نويسنده , , Xie Zhongsheng، نويسنده , , Eytan Modiano and Li Shu.، نويسنده ,
Issue Information
روزنامه با شماره پیاپی سال 2003
Pages
6
From page
127
To page
132
Abstract
A 3-D multigroup P3 neutron transport Monte Carlo code MCMG-BURN is developed for coupling neutron transport with burnup. MCMG-BURN code is based on Monte Carlo code MCNP with the continuous energy cross-section and the reactor lattice code WIMS. It uses the up-front multigroup macroscopic cross-section library based on burnup for Monte Carlo calculations. The almost consistent results with the experiments have been achieved.
Journal title
Annals of Nuclear Energy
Serial Year
2003
Journal title
Annals of Nuclear Energy
Record number
405757
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