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Corrosion Behavior of Zirconium Alloy Nuclear Fuel Cladding

Published online by Cambridge University Press:  21 February 2011

Anna C. Fraker  and
Jonice S. Harris
Anna C. Fraker
Affiliation:
Metallurgy Division, National Institute of Standards and Technology, Gaithersburg, MD 20899
Jonice S. Harris
Affiliation:
Metallurgy Division, National Institute of Standards and Technology, Gaithersburg, MD 20899
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Abstract

Zircaloy−2 and −4 are used as nuclear fuel cladding. Both alloys are more than ninety-eight percent 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 22°C and 95°C. 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.

Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 176: Symposium U – Scientific Basis for Nuclear Waste Management XIII , 1989 , 549
Copyright
Copyright © Materials Research Society 1990

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References

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