Multigroup Monte Carlo Calculation Coupled of Transport and Burnup

A 3-D multigroup P₃ approximation Monte Carlo code MCMG-BURN is developed by coupling the neutron transport and burnup.MCMG-BURN code is based on the continuous-energy cross-section Monte Carlo code MCNP and the lattice homogeneous code WIMS.It uses the multigroup cross-section libraries to simulate the critical test reactors and HFETR(High Flux Engineering Test Reactor).The agreement results with the MCNP results and experiments are achieved.The MCMG-BURN code is at almost the same precision with the MCNP code while requiring considerably less computing time.