Solubility of UO2 in molten Zircaloy-4 from 2000 to 2300 degrees C

This paper reports solubility measurements for unirradiated UO2 fuel at 2000, 2100, 2200, and 2300 degrees C in initially oxygen-free Zircaloy-4 (Zr-4) and at 2100, 2200, and 2300 degrees C in Zircaloy-4 with an initial 25 at.% oxygen content (Zr-4/25% oxygen); the latter alloy was used to represent the oxygen-saturated Zr-4 component of steam-oxidized cladding. The work was performed to determine the maximum extent of fuel dissolution that could occur at various temperatures during the high-temperature transient of a hypothetical reactor accident involving impaired cooling. These values could then be used to estimate the fractional release of the volatile fission product inventory, which, to a first approximation, would be proportional to the volume fraction of dissolved fuel