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MCS multi group cross sections generation for fast reactor analysis
MCS multi group cross sections generation for fast reactor analysis
2020
Tung Dong Cao Nguyen
Hyunsuk Lee
Xianan Du
Vutheam Dos
Tuan Quoc Tran
Deokjung Lee
Keywords:
reactor safety
Burnup
radiation transport
group
Materials science
Nuclear engineering
Neutron transport
neutron physics
Thermal hydraulics
calculation methods
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