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Modelling the Corrosion of Zirconium Alloys in Nuclear Reactors Cooled by High Temperature Water

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Modelling Aqueous Corrosion

Part of the book series: NATO ASI Series ((NSSE,volume 266))

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Abstract

This paper summarises the general approach to the modelling of the corrosion of Zircaloy fuel cladding in PWRs. It discusses the problems arising from the present approach and presents evidence for why the current models break down. An alternative model based on recent experimental evidence on the mechanism of the irradiation and LiOH effects on Zircaloy corrosion is proposed, and it is shown that it can give oxidation curves of the observed form. Not all the evidence necessary for a complete description of this approach is yet available.

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Authors and Affiliations

  1. Centre for Nuclear Engineering, University of Toronto, 184 College Street, Toronto, Ontario, M5S 1A4, Canada

    B. Cox

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  1. B. Cox
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Editors and Affiliations

  1. Royal Naval Engineering College, Manadon, Plymouth, UK

    Kenneth R. Trethewey

  2. Royal Military College of Canada, Kingston, Ontario, Canada

    Pierre R. Roberge

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© 1994 Springer Science+Business Media Dordrecht

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Cox, B. (1994). Modelling the Corrosion of Zirconium Alloys in Nuclear Reactors Cooled by High Temperature Water. In: Trethewey, K.R., Roberge, P.R. (eds) Modelling Aqueous Corrosion. NATO ASI Series, vol 266. Springer, Dordrecht. https://doi.org/10.1007/978-94-011-1176-8_9

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  • DOI: https://doi.org/10.1007/978-94-011-1176-8_9

  • Publisher Name: Springer, Dordrecht

  • Print ISBN: 978-94-010-4513-1

  • Online ISBN: 978-94-011-1176-8

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