STP1295

    Zircaloy-2 Lined Zirconium Barrier Fuel Cladding

    Published: Jan 1996


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    Abstract

    The introduction of Zr-lined “barrier” fuel clad tubing by GE in the early 1980s to counter the pellet-clad interaction (PCI) failure mechanism in fuel for boiling water reactors provided a major improvement in fuel reliability and operational flexibility. While the frequency of fuel failures has been substantially reduced in the past decade, an increased tendency has been observed for failed fuel rods to exhibit post-failure degradation in the form of longer cracks that allow release of radioactive off-gas and contamination of the reactor coolant circuit with tramp fuel material. One factor involved in this degradation is hydriding of the cladding at a location remote from the initial perforation of the fuel rod. This local hydriding can lead to secondary crack initiation in the cladding when stressed by the fuel expansion accompanying a power increase.

    A modification of Zr-lined “barrier” fuel clad tubing has been developed to retard post-failure local hydriding while retaining the proven PCI resistance of the high-purity sponge Zr barrier. By adding a thin inner layer of corrosion-resistant Zircaloy-2 bonded to the inner surface of the Zr-barrier tube, the resistance to internal corrosion and hydrogen generation in a perforated fuel cladding tube is made equivalent to that of an all-Zircaloy-2 tube. Tests show that the PCI mitigating capability of the Zr barrier is not compromised by this inner Zircaloy-2 liner. Materials considerations and manufacturing technology used to integrate this optional inner liner with other Zr barrier tubing properties and performance requirements are discussed with a summary of testing experience.

    Keywords:

    nuclear fuel cladding, zirconium barrier, inner Zircaloy liner, pellet clad interaction, post-failure degradation, corrosion resistance, hydriding, manufacturing processes, inreactor fuel performance, zirconium alloys, nuclear applications, radiation effects, crack propagation


    Author Information:

    Williams, CD
    General Electric Nuclear Energy, Wilmington, N.C.

    Marlowe, MO
    General Electric Nuclear Energy, Wilmington, N.C.

    Adamson, RB
    General Electric Nuclear Energy, Wilmington, N.C.

    Wisner, SB
    General Electric Nuclear Energy, Pleasanton, CA

    Rand, RA
    General Electric Nuclear Energy, Wilmington, N.C.

    Armijo, JS
    General Electric Nuclear Energy, Wilmington, N.C.


    Paper ID: STP16196S

    Committee/Subcommittee: B10.01

    DOI: 10.1520/STP16196S


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