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    The Effect of Neutron Irradiation on the Mechanical Properties of Zirconium Alloy Fuel Cladding in Uniaxial and Biaxial Tests

    Published: Jan 1970

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    Short-time axial tension, transverse ring tension, and biaxial closed end burst tests were conducted on sections of fuel cladding from 17 different batches and heat treatments of Zircaloy 2, Zircaloy 4, and Zr-2.5Cb alloy. Specimens were irradiated at 2 to 3×1020 n/cm2, E > × MeV, at temperatures of 125 to 250 C and tested at temperatures of 20 and 300 C.

    The level of residual cold work was found to have the greatest effect on the mechanical properties of the Zircaloys prior to irradiation, and, although irradiation altered the general level of each property, the effect of cold work was preserved after irradiation for all properties except the uniform elongation. Very low values of uniform elongation (< 1 percent) were obtained in the burst test after irradiation at all levels of cold work. In the instances where the texture and grain size influenced the properties, the effect was also preserved after irradiation. For the Zircaloys, the increment in the strength properties due to irradiation was constant at this fluence, independent of cold work, texture, and grain size. The decrements in the uniform and total elongations due to irradiation decreased with increasing cold work. The irradiation hardening of the Zr-2.5Cb alloy was greater than that experienced by the Zircaloys.

    Beta heat-treated Zircaloy had higher strength and lower ductility than alpha annealed material in the tension test. Grain size was found to have a much more significant effect on the strength and ductility in the burst and ring tests than in the axial tension test.


    neutron irradiation, fast neutrons, radiation effects, nuclear fuel cladding, tubing, zirconium alloys, Zircaloys, mechanical properties, ductility, elongation, tensile strength, yield strength, fractures (materials), cold working, heat treatment, texture, zirconium hydrides, axial stress, strains, tension tests

    Author Information:

    Hardy, DG
    Metallurgist, Chalk River Nuclear Laboratories, Atomic Energy of Canada Limited, Chalk River, Ontario

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP26623S