Feedback on neutron capture cross-section of 234U from analysis of measured inventories of UO2 fuel irradiated in TMI unit 1

Abstract

ABSTRACT For the 12 fuel samples taken from the two 15 × 15 PWR fuel assemblies irradiated in Three Mile Island (TMI) Unit 1, nuclide inventory calculations were performed using the continuous-energy Monte Carlo burn-up code MVP-BURN with JENDL-4.0. The calculated results of 234U were compared with the measured data to evaluate the accuracy of the neutron capture cross-section of 234U in JENDL-4.0. The average deviation of the calculated inventory/the measured inventory (C/E) from 1.0 (C/E－1s) was 2.3% for the eight fuel samples (average calculated burnup of 26.1 GWd/t) taken from the one-cycle-irradiation fuel assembly, and it was 10.5% for the four fuel samples (average calculated burnup of 51.5 GWd/t) from the two-cycle-irradiation fuel assembly. The positive trend in the C/E－1s with JENDL-4.0 for 234U was consistent with the C/E－1s for many of the fuel samples in the previous studies. The comparison of the C/E－1s in the present study with those obtained by SCALE 5.1 with ENDF/B-V and SCALE 6.1.2 with ENDF/B-VII.0 indicated that the larger neutron capture cross-sections of 234U in ENDF/B-V and ENDF/B-VII.0 mainly brought the C/E－1s closer to zero than those with JENDL-4.0.