You are being redirected because this document is part of your ASTM Compass® subscription.
    This document is part of your ASTM Compass® subscription.

    If you are an ASTM Compass Subscriber and this document is part of your subscription, you can access it for free at ASTM Compass

    Prediction of Creep Anisotropy in Zircaloy Cladding

    Published: 01 January 1994

      Format Pages Price  
    PDF (240K) 14 $25   ADD TO CART
    Complete Source PDF (21M) 796 $159   ADD TO CART

    Cite this document

    X Add email address send
      .RIS For RefWorks, EndNote, ProCite, Reference Manager, Zoteo, and many others.   .DOCX For Microsoft Word


    Due to the hexagonal crystal structure of zirconium and the radial orientation of the basal poles in Zircaloy cladding, the deformation of light water reactor Zircaloy fuel cladding is anisotropic. Plastic deformation of this cladding can be defined by the R and P factors that are the circumferential/radial and axial/radial contractile strain ratios under uniaxial deformation along the axial and circumferential direction, respectively. The in-reactor deformation performance of the cladding can be modeled with good accuracy if the R and P values of the irradiation-induced creep are known.

    In a boiling water reactor (BWR) fuel assembly, most fuel rods have a hoop stress to axial stress ratio of about 2:1. The assembly also contains several fuel rods (tie rods) that connect the upper and lower tie plates. The tie rods have an additional stress component in the axial direction. BWR assemblies can, furthermore, contain water rods that are free of stress, and therefore, provide a measure of stress-free irradiation growth. Post-irradiation deformation measurements of these three types of rods are used to derive the R and P factors for BWR cladding during irradiation.

    Laboratory tensile and creep tests at several temperatures have been performed on five types of cladding to determine if a short-term test could be used to obtain R values representative of the in-reactor values. For standard Zircaloy-2 cladding that was examined in-reactor, a good correlation with the in-reactor R value was obtained from both tensile and creep tests performed at 382°C. For another type of Zircaloy-2 cladding (late beta-quenched (LBQ) cladding), inreactor deformation performance correlates better with the tensile test results than the creep test results. The other three types of cladding (Zircaloy-4) with differing textures and processing histories have exhibited significant differences in the R values in the laboratory tests.


    zirconium alloys, contractile strain ratio, boiling water reactor, creep (materials), anisotropy, in-reactor deformation, zirconium, nuclear materials, nuclear applications, radiation effects

    Author Information:

    Perkins, RA
    Senior engineers, Siemens Power Corporation, Richland, WA

    Shann, S-H
    Senior engineers, Siemens Power Corporation, Richland, WA

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP15204S