Irradiated UO_2 Fuel Dissolution by molten Zircaloy: Some Results from the SCA Corium Interaction Thermochemistry Project

摘要

After preliminary testing, a test series was carried out in the hot cells using cladded segments of 3mm height of both natural UO_2 and irradiated UO_2 (54 GWd/tU burn-up) heated to 2000 deg C for durations of upto 200s in a direct electrical heating furnace under inert (Ar) atmosphere. The attack by the molten zircaloy cladding appeared macroscopically to be faster for the irradiated samples. The non-irradiated samples underwent a slower break-up whereas the irradiated samples broke into smaller irregular pieces due to the pre-existing cracking. Microprobe analysis showed that the melt was composed of two main phases: a metallic Zr-rich alpha-(Zr,U)O phase and a U-rich (U,Zr)O_2 ceramic phase. However the irradiated fuel showed a very high porosity as fission gas bubbles were formed and released. This causes a larger, melt/fuel interface and hence in an increased interaction rate compared with non-irradiated fuel. This implies similar dissolution rates may be expected in the irradiated fuel at over 100 deg C lower.