A　3-D　multigroup　P-3　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.　(C)　2002　Elsevier　Science　Ltd.　All　rights　reserved.