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


    Fracture of Zircaloy Cladding by Interactions with Uranium-Dioxide Pellets in Water Reactor Fuel Rods

    Published: 0

      Format Pages Price  
    PDF (264K) 14 $25   ADD TO CART
    Complete Source PDF (15M) 675 $62   ADD TO CART


    The paper summarizes the main features of a detailed theoretical study of Zircaloy cladding fracture that is caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn up. It is shown that unless stress corrosion crack growth in irradiated Zircaloy is able to proceed at very low stress intensifications, fracture is unlikely when there are uniform friction effects at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (that is, “uniform” conditions). Otherwise, the observed fuel rod failures must be due to departures from uniform conditions, and this investigation shows that a very high interfacial friction coefficient, and particularly fuel-cladding bonding, are means of providing sufficient stress intensification at a cladding crack tip to explain the occurrence of cladding fractures.

    The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking.


    zirconium, zirconium alloys, stress corrosion, iodine, power ramps, irradiation, interfacial friction, fuel-cladding bonding

    Author Information:

    Smith, E
    Consultant to Failure Analysis Associates, University of Manchester/UMIST, Manchester,

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP35590S