MCODE, Version 2.2: An MCNP-ORIGEN DEpletion Program
MCODE Version 2.2 is a linkage program, which combines the continuous-energy Monte Carlo code, MCNP-4C, and the one-group depletion code, ORIGEN2, to perform burnup calculations for nuclear fission reactor systems. MCNP is used as the advanced physics modeling tool providing the neutron flux solu...
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| Format: | Technical Report |
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Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program
2012
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| Online Access: | http://hdl.handle.net/1721.1/75242 |