Analysis of post-irradiation examination by SWAT4 with the JENDL-5 cross-section data

Watanabe, Tomoaki; Kikuchi, Takeo; Kasuya, Yuta*; Nomura, Takuro*; Suyama, Kenya

no journal, , 

JAEA has developed and maintained a burnup calculation code system SWAT4 to evaluate the nuclide composition of spent fuel. Due to the validations, various post-irradiation examinations (PIE) analyses have been performed using the JENDL-4.0 cross-section data. In this study, we performed the previous PIE analyses using the newly released JENDL-5 cross section data to confirm the changes in the analysis results due to the change to JENDL-5 and the differences in the results due to the reactor system and fuel type of the targeted PIE data. As a result of the analysis, it was found that the C/E values of nuclide composition by JENDL-5 did not show any significant difference in trend by reactor system or fuel type, and the results of JENDL-5 and JENDL-4.0 were similar in accuracy for all reactor systems and fuel types. As for the effect of the updated cross sections from JENDL-4.0 to JENDL-5 on C/E values, significant changes in C/E values for Pu-238, Am-241, and Cs-134 were observed due to the updated capture cross sections for Pu-238, Am-241, and Cs-133, independent of reactor system or fuel type existed. The results were highly dependent on the PIE samples as to whether the differences from the measured values were improved or not.