SEDL / STP / STP754-EB / STP37066S

Long-Term In-Reactor Corrosion and Hydriding of Zircaloy-2 Tubing

Hillner, E
Fellow engineer, Bettis Atomic Power Laboratory, West Mifflin, Pa.

Pages: 29    Published: Jan 1982

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Abstract

Three original Zircaloy-2 clad blanket fuel bundles from the pressurized-water reactor (PWR) at the Shippingport Atomic Power Station were discharged after continuous exposure during Cores 1 and 2. Detailed visual examination of these components after ∼6300 calendar days of operation (51 140 effective full power hours) revealed only the anticipated uniform light gray (posttransition) corrosion products with no evidence of unexpected corrosion deterioration, fuel rod warpage, or other damage. All corrosion films were found to be tightly adherent to the underlying cladding.

An extensive destructive examination of a selected fuel rod from each of three fuel bundles produced, as expected, significantly greater end-of-life rod average oxide film thicknesses when compared with corresponding values calculated from the time-temperature history of each component, employing a set of empirical equations generated from the out-of-pile (autoclave) testing of Zircaloy coupons. It is shown that the pretransition region and the time to transition are in adequate agreement with the empirical equations generated from the ex-reactor data but that there is a significant acceleration of the posttransition corrosion rate due to the irradiation exposure. Post-transition corrosion rate data were used to develop an empirical expression of the form R′=A(φ)n·R where R′ is the in-reactor posttransition corrosion rate for Zircaloy-2 tubing, A and n are experimentally determined constants, φ 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 struture).

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/Δ0 ratio for the entire posttransition region) to in-reactor exposures.