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Study of Zircaloy-4 Cladding Air Degradation at High Temperature

[+] Author Affiliations
Marina Lasserre, Olivia Coindreau

IRSN, Saint Paul lez Durance, France

Michèle Pijolat, Véronique Peres

ENSM-SE, Saint-Etienne, France

Michel Mermoux

LEPMI-Phelma Campus, Saint Martin d’Hères, France

Jean-Paul Mardon

AREVA-NP, Lyon, France

Paper No. ICONE21-16440, pp. V006T16A041; 9 pages
  • 2013 21st International Conference on Nuclear Engineering
  • Volume 6: Beyond Design Basis Events; Student Paper Competition
  • Chengdu, China, July 29–August 2, 2013
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5583-6
  • Copyright © 2013 by ASME


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.

Copyright © 2013 by ASME



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