Post irradiation examinations and the validity of computational analysis for high burn-up UO2 and MOX spent fuels

Post irradiation examinations (PIE) to obtain basic data for evaluation of neutron source and burn-up credit were carried out on a high burn-up UO2 spent fuel pin (average burn-up around 60 GWd/tU) and two MOX spent fuel segments (average burn-up around 45 GWd/tU). The validity of the ORIGEN-2 and VIM-BURN code was discussed by use of PIE data. From electron probe microanalysis (EPMA) results, the local peak burn-up of Pu agglomerate near pellet surface of MOX fuel reached 267 GWd/t. In the results of ORIGEN-2 calculation for nuclides composition, U and Pu agreed within 13% except for Pu-238. In the results for fission products, Nd-143, Nd-145 and Eu-153 which have large neutron absorption cross sections agreed within 10%. From the results of VIM-BURN calculation for radial profile of U, Pu and important fission products, U and Pu radial profile agreed within uncertainties coming from measurement procedure. The radial burn-up profile in high burn-up UO2 fuel pellet agreed within 15% and in MOX fuel pellet agreed within uncertainties coming from measurement procedure. (author)