Phenomenological study of in-reactor corrosion of Zircaloy-4 in pressurized water reactors
- Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)
Uniform in-reactor corrosion has been examined on pressurized water reactor (PWR) Zircaloy-4 fuel cladding that were irradiated up to four cycles in the KORI Unit 1. The oxide layer that was formed on the Zircaloy-4 fuel claddings changes from a uniform black oxide into the nonuniform white granular oxides when the oxide layer reaches about 5 to 10 {mu}m. With irradiation exposure, enough of the white granular oxides have nucleated above the black oxide layer to entirely cover the fuel cladding, resulting in a gray or white cladding surface. The change in the oxide layer growth pattern from the uniform black oxide layer into the nonuniform granular oxides resulted in the formation of macro-pores or cracks in the oxide layer, causing the corrosion rate to change into a linear rate from the cubic rate, especially at the beginning of fuel rod life. Consequently, the nucleation of white granular oxides is a phenomenological indicator of the transition of the in-reactor corrosion rate. To explain the effect of hydride precipitates on the in-reactor corrosion, the corrosion behavior for a defective fuel rod with a through-hole and in intact fuel rod adjacent to it (both were irradiated for two cycles in the KORI Unit 1) has been investigated. Even through both of them were found to have almost identical operating power histories, their corrosion behavior was quite different: the maximum oxide layer thickness for the intact fuel rod was less than 10 {mu}m, while the defective fuel rod had an abnormally thick oxide of 50 {mu}m, especially at an elevation where a large extent of hydrides (corresponding to 1,295 to 1,520 ppm hydrogen) were precipitated at the cladding outer surface.
- OSTI ID:
- 55681
- Report Number(s):
- CONF-930611--; ISBN 0-8031-2011-7
- Country of Publication:
- United States
- Language:
- English
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