Monte Carlo burnup analysis of measured nuclide inventories on high-burnup PWR-UO₂ and BWR-MOX fuels in the RUBUS program

To validate the nuclear data in JENDL-4.0 and obtain information to improve their accuracy, the analysis of the measured nuclide inventories of the high-burnup PWR UO₂ and BWR MOX fuels in the REBUS program was performed by using MVP-BURN with the nuclear library and burnup chain based on JENDL-4.0, and the ratios of the calculated and measured inventories (C/Es) were compared with the previous results obtained with JENDL-3.2 or JENDL-3.3. As a result, an improvement in the accuracy of the neutron cross-sections and cumulative fission yields (CFYs) in JENDL-4.0 was confirmed for the inventories of ²³⁶U, ²³⁸Pu, ²³⁹Pu, ²⁴¹Pu, ²⁴²Pu, ²⁴¹Am, ²⁴³Cm, ²⁴⁴Cm, ¹³³Cs, ¹⁴⁵Nd, ¹⁴⁸Sm, ¹⁴⁹Sm, ¹⁵¹Sm, and ¹⁵³Eu of the PWR UO₂ fuel and ²³⁹Pu, ²⁴¹Pu, ²⁴²Pu, ²⁴¹Am, ²⁴³Am, ²⁴³Cm, ²⁴⁴Cm, ²⁴⁶Cm, ¹³⁴Cs, ¹⁴⁴Ce, ¹⁴⁵Nd, ¹⁴⁷Sm, and ¹⁵⁴Eu of the BWR MOX fuel. Based on the C/Es obtained with JENDL-4.0, the directions to improve the neutron cross-sections and CFYs were tentatively proposed. They included the neutron capture cross-sections of ²³⁴U, ²³⁶U, ²³⁷Np, ²³⁸Pu, ²⁴¹Am, ²⁴²Cm, ²⁴⁵Cm, ¹⁴³Nd, ¹⁴⁷Pm, ¹⁵⁴Eu, and ¹⁵⁵Eu, the fission cross-sections of ²⁴³Cm and ²⁴⁵Cm, and the CFYs of ¹⁴⁴Nd, ¹⁴⁷Pm, and ¹⁵⁵Eu.