Zircaloy cladding, providing the first containment of UO2 fuel in Pressurised Water Reactors, can be exposed to air during accidental situations. This might occur during reactor operation (in case of a core meltdown accident with subsequent reactor pressure vessel breaching), under shutdown conditions with the upper head of the vessel removed, in spent fuel storage pools after accidental loss of cooling or during degraded transport situations. The fuel assemblies inadequately cooled, heat up and as a result, corrosion of Zircaloy claddings takes place. This paper is devoted to the kinetic analysis of Zy4 corroded at 850°C in 20% oxygen – 80% nitrogen partial pressure atmosphere to support the comprehension of the degradation mechanisms involved during the post-transition stage.
- Nuclear Engineering Division
Study of Zircaloy-4 Cladding Air Degradation at High Temperature
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Lasserre, M, Coindreau, O, Pijolat, M, Peres, V, Mermoux, M, & Mardon, J. "Study of Zircaloy-4 Cladding Air Degradation at High Temperature." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 6: Beyond Design Basis Events; Student Paper Competition. Chengdu, China. July 29–August 2, 2013. V006T16A041. ASME. https://doi.org/10.1115/ICONE21-16440
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