Impact of detailed depletion chains and cross-section treatments on nuclide inventory using SWAT-X

渡邉 友章  ; 多田 健一   ; 山本 章夫*

Watanabe, Tomoaki; Tada, Kenichi; Yamamoto, Akio*

This study investigates the impact of detailed depletion-chain and cross-section treatments on burnup calculation results using SWAT-X, which is a Monte Carlo burnup code currently under development by the authors. In this work, a new capability to construct arbitral depletion chains based on ENDF-6 formatted nuclear data libraries has been implemented in SWAT-X. Detailed depletion chains consisting of 4,070 nuclides were constructed using decay and fission yield data from JENDL-5. Using these depletion chains, burnup calculations for a UO2 pin-cell model assuming a PWR were performed with different cross-section treatments. Cross sections for 226 major nuclides were treated in continuous energy in the Monte Carlo calculations for all cases, whereas those for the remaining nuclides in the depletion chain were treated differently, either in continuous energy or multi-group with different structures, to evaluate their impact on the calculated nuclide inventories. The results indicate that the effect of the cross-section treatment is small, suggesting that a simplified approach with multi-group cross sections is sufficient. In this study, the impact of cross-section coverage in the depletion chain was also examined by extending the data to include a total of 2,697 nuclides using TENDL-2023. As a result, some changes and new generations of nuclides were observed, although their overall impact on the nuclide inventory was limited.