STP681

    Further Evidence of Zircaloy Corrosion in Fuel Elements Irradiated in a Steam Generating Heavy Water Reactor

    Published: Jan 1979


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    Abstract

    An earlier paper (Trowse et al, ASTM STP 633) described the nodular corrosion of Zircaloy and other zirconium alloys in a steam generating heavy water reactor (SGHWR) and related environments. The observations have now been extended to more highly irradiated elements up to the design burn-up of 20 MWD/kgU mean. Maximum observed local oxide thicknesses were less than the earlier predictions of 200 μm for SGHWR cladding at the design burn-up and are acceptable from fuel performance considerations.

    Most of the data presented relates to Zircaloy-2 in the stress-relieved condition, but results from elements with recrystallized cladding show that oxidation resistance can be improved by this treatment. In particular, the oxide nodules are much finer in the recrystallized alloy, leading to a generally smaller coverage of nodular oxide.

    The maximum depth of oxidation is correlated to fast neutron dose and is apparently little affected by coolant mass flow-rate or steam quality.

    Previous work had shown evidence for oxidation enhancement of Zircaloy at support grid positions by galvanic coupling to stainless steel and also by localized coolant flow disturbance. In order to determine the relative importance of these two factors, an experimental element fitted with alternate stainless steel and Zircaloy grids of similar geometry was irradiated in SGHWR. The results confirmed that most of the oxidation enhancement is due to the galvanic coupling but that significant flow effects at grids were also present in the higher flow, higher steam quality region of the element down-stream of the peak flux position.

    Keywords:

    zirconium, zirconium alloys, nuclear fuel claddings, boiling water reactors, irradiation, neutron flux, zirconium oxide, nodules, recrystallization (metallurgy), intermetallics, galvanic corrosion


    Author Information:

    Sumerling, R
    Senior assessor, Water Reactor Fuel Examination, section leader, Mechanical Properties and Corrosion, and section leader, Water Reactor Fuel Examination, United Kingdom Atomic Energy Authority, Windscale Nuclear Laboratories, Sellafield, Seascale, Cumbria,

    Garlick, A
    Senior assessor, Water Reactor Fuel Examination, section leader, Mechanical Properties and Corrosion, and section leader, Water Reactor Fuel Examination, United Kingdom Atomic Energy Authority, Windscale Nuclear Laboratories, Sellafield, Seascale, Cumbria,

    Stuttard, A
    Senior assessor, Water Reactor Fuel Examination, section leader, Mechanical Properties and Corrosion, and section leader, Water Reactor Fuel Examination, United Kingdom Atomic Energy Authority, Windscale Nuclear Laboratories, Sellafield, Seascale, Cumbria,

    Hartog, JM
    Research manager and section leader, Chemical Technology Group, United Kingdom Atomic Energy Authority, Springfields Nuclear Laboratories, Salwick, Preston,

    Trowse, FW
    Research manager and section leader, Chemical Technology Group, United Kingdom Atomic Energy Authority, Springfields Nuclear Laboratories, Salwick, Preston,

    Sims, P
    Physicist, Technology Branch, United Kingdom Atomic Energy Authority, Atomic Energy Establishment, Winfrith, Wool, Dorset,


    Paper ID: STP36675S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP36675S


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