Implementation of ORIGEN2 code for the general-purpose reactor analysis code system, MARBLE

Sugawara, Takanori  ; Kodama, Yasuhiro*; Nishihara, Kenji  ; Hirai, Yasushi*

The general-purpose reactor analysis code system, MARBLE, has been used to calculate neutron transport and burn-up calculations for Accelerator-Driven System (ADS). In the burn-up calculation of MARBLE, fission product (FP) nuclides had been treated as lump FP in the past. It meant that MARBLE was unable to treat residual nuclides such as rare-earth ones which would be generated by the fuel exchange of the ADS. To treat residual nuclides, ORIGEN2, which was one of the most famous burn-up calculation codes was implemented to MARBLE. By the implementation of ORIGEN2 code, it was available to treat FP nuclides by each nuclide and to consider the residual nuclides in the ADS burn-up calculation.