Long-term in-reactor corrosion and hydriding of Zircaloy-2 tubing

An extensive destructive examination of a selected fuel rod from each of three fuel bundles produced significantly greater end-of-life rod average oxide film thicknesses when compared with corresponding values calculated from time-temperature history of each component, employing a set of empirical equations generated from out-of-pile (autoclave) testing of Zircaloy coupons. The pretransition region and the time to transition are in adequate agreement with the empirical equations generated from the ex-reactor data but there is a significant acceleration of the posttransition corrosion rate due to the irradiation exposure. Posttransition corrosion rate data were used to develop an empirical expression of the form where R' is the in-reactor posttransition corrosion rate for Zircaloy-2 tubing, A and n are experimentally determined constants, phi is the posttransition fast flux exposure, and R is the posttransition temperature-dependent corrosion rate generated from exreactor (autoclave) testing. The Multipurpose Extended Life Blanket Assembly cladding in the mid-sections of the rods (α microstructure) was found to have corroded at a rate that is approximately 25 percent greater than the corresponding welded end cap regions (β-quenched structure). It has been observed that the total hydrogen content of the cladding was less than that calculated from the ex-reactor expressions, using the measured oxide film thickness. It has also been found that the hydrogen pickup per unit weight gain is two to three times greater in the late posttransition period than that observed in the early posttransition region, possibly casting doubt on the validity of the application of the ex-reactor hydrogen pickup expression (that is, a constant ΔH/ΔO ratio for the entire posttransition region) to in-reactor exposures