STP1245

    Corrosion Behavior of Zircaloy-4 Cladding with Varying Tin Content in High-Temperature Pressurized Water Reactors

    Published: Jan 1994


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    Abstract

    Fuel rods clad with Zircaloy-4 with varying tin contents (1.33 to 1.58% Sn) and annealing parameters (1.0 to 4.1 × 1017 h with Q/R = 40 000°K) were irradiated in demonstration fuel assemblies in a high-temperature pressurized water reactor (PWR) to burnups in excess of 35 giga watt days per metric ton of uranium (GWd/MTU). The same cladding variants were subjected to long-term static water autoclave tests at 633°K of duration greater than 1100 days. Production fuel rods fabricated with low-tin (1.33 Sn) and high-tin (1.55 Sn) Zircaloy-4 cladding were also irradiated in regular fuel assemblies in two high-temperature PWRs to burnups up to 48 GWd/MTU. Poolside cladding oxide thickness measurements were conducted on 167 high-tin rods and 67 low-tin rods during refueling outages. The measured, circumferentially averaged, peak cladding oxide thickness values ranged from 3 to 113 μm. At high burnups, the oxide thickness on low-tin cladding was 30 to 40% lower than that on high-tin cladding. The long-term autoclave results also showed the beneficial effect of lower tin level on the corrosion rate, although to a lower degree than in PWRs. The 633 K water autoclave test appears to rank the corrosion resistance of the investigated Zircaloy-4 variants in the same order as in PWRs. Hydrogen analysis results indicate that tin level does not influence the hydrogen uptake of autoclave-tested samples. The observed in-PWR influence of tin level on the Zircaloy-4 cladding corrosion rate was incorporated in the ESCORE clad corrosion model by adjusting the pre-exponential term in the post-transition corrosion rate equation.

    A corrosion rate acceleration at high burnups may be related to either hydride precipitation at the metal oxide interface or to degradation of the oxide thermal conductivity. The observed effect of tin on the uniform corrosion resistance of Zircaloy-4 in high temperature water is consistent with the corrosion mechanism that assumes the migration of 02- anion vacancies through the oxygen deficient zirconia to be the rate-controlling process. It is postulated that lower tin level in Zircaloy-4 decreases the vacancy concentration in the oxide and, thereby, increases the corrosion resistance. It may be possible to further improve the corrosion resistance of low-tin Zircaloy-4 by optimizing the nitrogen and tin concentrations in Zircaloy.

    Keywords:

    corrosion, autoclave corrosion, tin content, corrosion modeling, hydrogen uptake, annealing effect, anion vacancy migration, zirconium, zirconium alloys, nuclear materials, nuclear applications, radiation effects


    Author Information:

    Garde, AM
    Consulting engineer, supervisor, consulting engineer, and consulting engineer, ABB Combustion Engineering, Nuclear Fuel, Windsor, CT

    Pati, SR
    Consulting engineer, supervisor, consulting engineer, and consulting engineer, ABB Combustion Engineering, Nuclear Fuel, Windsor, CT

    Krammen, MA
    Consulting engineer, supervisor, consulting engineer, and consulting engineer, ABB Combustion Engineering, Nuclear Fuel, Windsor, CT

    Smith, GP
    Consulting engineer, supervisor, consulting engineer, and consulting engineer, ABB Combustion Engineering, Nuclear Fuel, Windsor, CT

    Endter, RK
    Principal engineer, ABB Combustion Engineering, Nuclear Services, Windsor, CT


    Paper ID: STP15219S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP15219S


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