Development and verification of code IMPC-Depletion for nuclide depletion calculation
Zelong Zhao, Yongwei Yang, Qingyu Gao
Topics & Concepts
BurnupNuclideNuclear transmutationNuclear engineeringNeutron fluxMonte Carlo methodMOX fuelNuclear physicsComputer sciencePhysicsNeutronMathematicsEngineeringPlutoniumStatisticsNuclear reactor physics and engineeringNuclear Materials and PropertiesNuclear Physics and Applications