Author/Authors :
Deng Li، نويسنده , , Xie Zhongsheng، نويسنده , , Eytan Modiano and Li Shu.، نويسنده ,
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.
