STP754

    Effect of Zirconium Oxide on the Stress-Corrosion Susceptibility of Irradiated Zircaloy Cladding

    Published: Jan 1982


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    Abstract

    Cladding specimens were obtained from two fuel rods irradiated in the Big Rock Point Reactor to a burnup of ∼8 gigawatt days per ton. Both claddings had a uniform, thick (∼4 µm) zirconium oxide layer on the inner surface. The significant difference between the two rods was the degree of fission-gas release (0.2 versus 14.3 percent). The high-gas-release rod contained a significantly greater amount of fission-product deposits, and, in some areas, breaks in the inner-surface oxide were observed. The cladding speciments, with the fuel removed, were subjected to stress-rupture tests in an iodine environment using internal gas pressurization to evaluate their stress corrosion cracking (SCC) susceptibility. All tests were conducted at an initial iodine concentration of 0.6 mg/cm2 and a temperature of 325°C. Specimens from the high-gas-release rod exhibited signifciantly increased susceptibility to iodine SCC, with a threshold stress level of ∼200 MPa as compared with ∼280 MPa in the specimens from the low-gas-release rod. The results suggest that the inner-surface oxide provides a barrier to iodine penetration. Hence, the thick unbreached oxide on the low-gas-release rod provided greater resistance to iodine SCC than the breached oxide on the high-gas-release rod.

    Keywords:

    zirconium, nuclear industry, iodine, stress-corrosion cracking, irradiation, surface oxide


    Author Information:

    Mattas, RF
    Metallurgists and senior metallurgist, Argonne National Laboratory, Argonne, Ill.

    Yaggee, FL
    Metallurgists and senior metallurgist, Argonne National Laboratory, Argonne, Ill.

    Neimark, LA
    Metallurgists and senior metallurgist, Argonne National Laboratory, Argonne, Ill.


    Paper ID: STP37053S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP37053S


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