Abstract: A new code MCORGS which links MCNP and ORIGENS is developed to do the transport-burnup calculation.MCORGS can be used to compute both criticality problem and subcriticality problem with an external neutron source.It has no limitation on neutron spectrum or geometry shape.MCORGS is more versatile than MCCOOR which is dedicated to compute thermal reactors with regular shape.Visual FORTRAN is used to develop MCNP,ORIGENS and MCORGS,so it can be run in any WINDOWS operation system.MCORGS is easy to use for it automatically produces lots of information in input files.There are different treatments in absolute flux calculation and cross section update between MCORGS and MCCOOR.Results of a VVER-1000 benchmark show that MCORGS can get more accurate results in less time than MCCOOR.ADS benchmark is also calculated to validate its ability to deal with fast sub critical reactor with external source.