Autoclave study of zirconium alloys with and without hydride rim

J. Wei, P. Frankel, M. Blat, A. Ambard, R. J. Comstock, L. Hallstadius, S. Lyon, R. A. Cottis, M. Preuss

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    Abstract

    Autoclave corrosion experiments were conducted on a number of zirconium alloys in different heat treatment conditions. The alloys tested in the present work were Zircaloy-4, ZIRLOH (ZIRLO is a registered trademark of Westinghouse Electric Company LLC in the USA and may be registered in other countries throughout the world. All rights reserved. Unauthorised use is strictly prohibited.) and two variants of ZIRLO with significantly lower Sn levels, referred to here as A-0·6Sn and A-0·0Sn. Typical corrosion kinetics with a change from pre- to post-initial transition was observed with ZIRLO and Zircaloy-4 displaying the shortest time to the initial transition after 120-140 days of autoclave exposure, followed by A-0·6Sn materials after 140-260 days. A-0·0Sn materials showed no sign of transition even after 360 days although one sample tested to 540 days had gone through transition. Material in the stress relieved condition generally experienced initial transition earlier than the same alloy in the recrystallised condition. Pretransition samples had a universally black oxide layer, which eventually developed grey patches when transition occurred. Practically, all non-hydrogen charged alloys showed a strong trend towards cubic oxide growth rates. Cathodic hydrogen charging was conducted to simulate end of life condition of cladding tubes, forming a hydride rich rim region at the outer surface of the cladding tubes. Hydrogen charged materials generally experienced accelerated corrosion of different degrees with the exception of recrystallised A-0·0Sn and partially recrystallised A-0·6Sn showing no sign of acceleration. It therefore seems that increasing tin levels has a negative impact on autoclave corrosion behaviour for materials with and without a hydride rich rim. In developing advanced alloys for use in cladding, this effect has been balanced against the benefits that Sn is known to provide in-reactor, including robustness in corrosion behaviour and reduced irradiation growth. It was noted that most materials with a hydride rich rim exhibit parabolic corrosion kinetics with decreased initial weight gain but increased overall weight gain. © 2012 Institute of Materials.
    Original languageEnglish
    Pages (from-to)516-528
    Number of pages12
    JournalCorrosion Engineering Science and Technology
    Volume47
    Issue number7
    DOIs
    Publication statusPublished - Nov 2012

    Keywords

    • Autoclave testing
    • Nuclear cladding
    • Zircaloy-4
    • Zirconium alloy corrosion
    • Zirconium alloys
    • Zirconium hydride
    • ZIRLO
    • ZrO 2

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