STP633

    Nodular Corrosion of Zircaloy-2 and Some Other Zirconium Alloys in Steam Generating Heavy Water Reactors and Related Environments

    Published: Jan 1977


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    Abstract

    In steam generating heavy water reactors (SGHWR), nodular or “patch-type” corrosion is the dominant form of attack on Zircaloy, as it is in other boiling or oxygenated reactor environments. The most severe oxidation occurs close to the stainless steel spacer grids, although nodules are seen elsewhere on cladding and on non-fueled components, a maximum thickness of 160 μm at 20 MWd/kg uranium local burn-up having been recorded to date. The extensive data from post-irradiation examination of SGHWR fuel elements are reviewed and discussed from the points of view of determination of rate laws for oxide growth and the establishment of corrosion mechanisms.

    A rate law for oxide growth on Zircaloy-2 at grid positions is determined empirically by seeking correlations between oxidation and irradiation exposure. It is concluded that a linear relationship between oxide thickness and local fast fluence or burn-up represents the data reasonably well, the dependence upon channel flow and exit quality being less significant. Zr-2.5Nb alloy, in the recrystallized condition, was more resistant to nodular oxidation than was Zircaloy-2.

    The mechanism of nodular corrosion cannot be expressed in quantitative terms yet, but a qualitative approach is valuable in suggesting wider ranging correlations and possible ameliorative measures. The suggested mechanism requires: (a) an enhancement of the general corrosion rate, controlled in-reactor by coolant chemistry and radiolysis but dependent also upon galvanic coupling or flow effects, or both; and (b) factors specifically responsible for nodule initiation and growth, for example, local cathodic depolarization at intermetallic inclusions and stress-induced recrystallization of the oxide.

    The effect of metallurgical factors in the metal substrate and of surface treatment of the alloy also are discussed and current experimental work, intended to check some of the postulated features of the oxidation mechanism, is describe.

    Keywords:

    zirconium alloys, zirconium, nuclear fuel claddings, boiling water, reactors, corrosion, corrosion mechanisms, neutron flux, zirconium oxides, initiation, recrystallization (metallurgy), stresses, cathodic depolarization, intermetallics, galvanic, corrosion, radiation chemistry, linear regression, water, steam


    Author Information:

    Trowse, FW
    Section leader, Chemical Technology Group, Reactor Fuel Element Laboratories, United Kingdom Atomic Energy Authority, Springfields, Salwick,

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

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


    Paper ID: STP35574S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP35574S


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