Corrosion behavior of irradiated Zircaloy
- GE Nuclear Energy, Pleasanton, CA (United States). Vallecitos Nuclear Center
There is ample evidence in the literature of the effects of reactor irradiation on the microstructure and corrosion behavior of zirconium alloys. Specifically, it has been shown that boiling water reactor (BWR) irradiation generally induces nodular corrosion and causes marked changes in precipitate structure and composition. The purpose of this study is to determine the effects of irradiation-induced microstructural changes on post-irradiation corrosion behavior and to gain insight into the operating in-reactor corrosion mechanisms. Zircaloy-2 and Zircaloy-4 were irradiated in BWRs at a temperature near 561 K. Neutron fluences were at various values between 2 and 14 {times} 10{sup 25} n/m{sup 2} (E > 1 MeV). Post-irradiation corrosion tests were conducted at 589, 673, and 793 K using standard techniques. Transmission electron microscopy (STEM) was conducted on unirradiated, as-irradiated, and corrosion-tested materials. STEM studies of the Zircaloy-4 specimens indicate that the observed corrosion behaviors can be correlated to irradiation-induced increases in the matrix solute concentration. The effects of redistribution and re-precipitation of solute during the corrosion tests are also examined. These results are related to the basic mechanisms of both nodular and uniform corrosion in a BWR.
- OSTI ID:
- 55663
- Report Number(s):
- CONF-930611-; ISBN 0-8031-2011-7; TRN: 95:012908
- Resource Relation:
- Conference: 10. international symposium on zirconium in the nuclear industry, Baltimore, MD (United States), 21-24 Jun 1993; Other Information: PBD: 1994; Related Information: Is Part Of Zirconium in the nuclear industry: Tenth international symposium. ASTM STP 1245; Garde, A.M.; Bradley, E.R. [eds.]; PB: 818 p.
- Country of Publication:
- United States
- Language:
- English
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