STP1245

    Effects of Pressurized Water Reactor (PWR) Coolant Chemistry on Zircaloy Corrosion Behavior

    Published: Jan 1994


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    Abstract

    A number of the test programs currently in progress at the Halden Project are aimed at evaluating Zircaloy cladding performance in terms of corrosion behavior, particularly at high burnup.

    A pressurized water reactor (PWR) facility has been used to determine the effects of high (4 to 4.5 ppm) lithium on the corrosion behavior of Zircaloy-4 cladding material exposed to nucleate boiling (1% void) and to one-phase cooling conditions. The test employed four fuel rod segments, base-irradiated to an average burnup of 28.5 MWd/kg UO2 in a commercial power plant, and with average initial oxide layers of either 10, 20, or 40 μm, respectively. Four sets of oxide layer thickness measurements were performed on the four rods during the course of the investigation. Pre-test eddy current measurements were performed on the as-received segments and two interim sets of measurements were made after 80 and 245 full power days of irradiation. Final oxide measurements were made after 425 full power days, when maximum rod burnup was 45 MWd/kg UO2.

    Average oxide layer increases of 30 and 55 μm were observed for the rods exposed to one-phase cooling and nucleate boiling conditions, respectively. Although maximum oxide thickness exceeded 95 μm, no evidence of spalling was apparent. The measured oxide layer thicknesses compared favorably with predicted values based on model calculations derived from power plant data, with little evidence of enhanced corrosion rates in the presence of increased lithium concentrations.

    Keywords:

    zirconium alloys, lithium hydroxide, surface boiling, in-pile tests, model prediction, zirconium, nuclear materials, nuclear applications, radiation effects


    Author Information:

    Karlsen, T
    Research scientist and program manager, OECD Halden Reactor Project, Halden,

    Vitanza, C
    Research scientist and program manager, OECD Halden Reactor Project, Halden,


    Paper ID: STP15220S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP15220S


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