Corrosion behavior of zirconium alloy nuclear fuel cladding

The purpose of this study was to provide data for use in evaluating corrosion behavior of the zirconium alloys, Zircaloy-2 and Zircaloy-4, and for use in determining whether long term credit can be claimed for the cladding in preventing radionuclide release to the environment. The US Nuclear Regulatory Commission (NRC) requires that nuclear waste containment shall be substantially complete for a period of 300 to 1,000 years and that thereafter, no more than one part in 105 of the inventory of radionuclides present at 1,000 years after closure may be released annually from the engineered barrier system. It is not known whether it would be necessary to take credit for the cladding to meet the release requirement. Both alloys are more than 98% zirconium and are corrosion resistant to various media. Electrochemical measurements using polarization techniques have been made on these alloys in aqueous media with a pH of 8.5 and varying ionic concentration (1X and 10X) at temperatures of 22C and 95C. Results showed that under the test conditions of the study these alloys passivated and had negligible corrosion rates, but there were some variations in passivation due to surface preparation and some crevice corrosion was observed. Data are presented and discussed in terms of passivity, breakdown potential and susceptibility to localized corrosion