Research on the properties of high-burnup and high plutonium content mixed-oxide fuels

Abstract

This paper presents the major results and outcome from a study on the properties of uranium and plutonium mixed oxide (MOX) fuel, aiming to expand the technical bases concerning MOX fuels and to review the previous reports on MOX fuel properties issued by the former Japanese nuclear safety commission. An irradiation test was conducted on MOX fuels fabricated through different processes, up to a local burnup of about 74 GWd/t, which is beyond the current Japanese licensing limit of 62 GWd/t in pellet burnup. Another irradiation test was conducted on MOX fuels with an initial Pu content of 14.3%, which also exceeds the Japanese licensing limit of 13%. After the irradiation tests, MOX fuel properties were studied using different measuring methods in the post-irradiation examination. The results showed no obvious deviation from the MOX fuel properties in the previous reports, while various new useful results were obtained. Elemental analysis via secondary ion mass spectrometry on the MOX fuels revealed that Pu rich agglomerates retained more fission gas bubbles than the matrix. Fuel centerline temperature measurements during irradiation and thermal diffusivity measurements using the laser flash method revealed that the thermal conductivity of MOX fuel was lower than that of UO2 fuel prior to irradiation, whereas it became comparable with that of the UO2 fuel by irradiation. Volume change measurements using both a non-destructive method and destructive methods indicated that the swelling rates of the MOX fuel with Pu rich agglomerates were slightly higher than the UO2 swelling rate in a reference.