Validation of burnup calculation code SWAT4 by evaluation of isotopic composition data of mixed oxide fuel irradiated in pressurized water reactor
Kashima, Takao; Suyama, Kenya; Mochizuki, Hiroki*
The nuclear fuel cycle program of Japan would be delayed because of the impact of the Fukushima Daiichi NPP accident in 2011. Excessive plutonium, however, has to be utilized as mixed-oxide (MOX) fuel to reduce the quantity of plutonium possessed by Japan. Calculation codes and libraries adopted in the fuel cycle analyses of MOX fuel should be benchmarked based on comparison between calculation results and experimental data. From another viewpoint, nuclide inventory analyses of MOX fuel is important for evaluations of the Fukushima accident because MOX fuel has been loaded in the Unit 3 reactor. ARIANE is a PIE program which includes measurements of nuclide compositions of spent MOX fuels discharged from both of pressurized and boiling water reactors. In this study, the PIE data of MOX fuels irradiated in a pressurized water reactor were analyzed by the integrated burnup code system SWAT4 that combines the point burnup system ORIGEN2 and neutron transport calculation solvers, the continuous energy Monte Carlo code MVP or MCNP, and the deterministic neutronics calculation code SRAC. The calculation results of SWAT4 have generally same trends with the case of UO fuel analyses. For major uranium and plutonium isotopes, deviations less than 5% were obtained. This means that SWAT4 has the same accuracy to predict isotopic compositions of irradiated MOX fuel with the case of UO fuel. The radial distribution of isotopes in a pellet was also analyzed, whose results were compared with that measured by SIMS. SWAT4 predicted well the isotope and burnup distributions in an irradiated MOX pellet.