Feedback on neutron capture cross sections of ²³⁸Pu and ²⁴¹Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

Abstract

The atomic fractions of 238 Pu and 241 Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for 238 Pu and 241 Am. The thermal neutron capture cross sections of 238 Pu and 241 Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of 238 Pu on UO 2 and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of 238 Pu and 241 Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.