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    Corrosion and Hydriding Performance of Zircaloy Tubing After Extended Exposure in the Shippingport Pressurized Water Reactor

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    Three original blanket fuel bundles from the Zircaloy-2 clad pressurized water reactor (PWR) at the Shippingport Atomic Power Station were discharged after continuous exposure for the entire first core, consisting of one original seed and three seed refuelings, and for the first seed life of the second core. Detailed visual examinations of these components after ∼4100 calendar days of operation (41 000 EFPH) revealed only the anticipated uniform gray-tan (post-transition) corrosion products with no evidence of a gross Zircaloy corrosion acceleration; all corrosion films were found to be tightly adherent to the underlying cladding.

    An extensive destructive evaluation of two fuel rods from each of the three fuel bundles produced greater oxide film thicknesses (7 to 12 μm or 0.28 to 0.47 mils) than the values predicted from the time-temperature history of the components (3 to 5 μm or 0.12 to 0.20 mils); an average corrosion acceleration factor of 2.4 can be calculated from the data. In slightly different terms, the observed corrosion acceleration was equivalent to raising the average surface temperature by ∼40°F. The accelerated corrosion attack was accompanied by a considerably lower total hydrogen pickup than that calculated from the observed film thicknesses. This combination Of an irradiation-induced corrosion acceleration and a less-than-predicted hydrogen absorption is suggestive of operation in an oxygenated environment although the Shippingport PWR coolant is maintained at a hydrogen concentration in the range 10 to 60 c3 H2/kg H2O.


    Zircaloys, corrosion, hydrogen pickup, irradiation, pressurized water reactors, tubes, hydride redistribution, zirconium, zirconium alloys

    Author Information:

    Hillner, E
    Fellow engineer, Westinghouse Electric Corp., Bettis Atomic Power Lab., West Mifflin, Pa.

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP32131S