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    In-Reactor Creep of Zr-2.5Nb Fuel Cladding

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    Commercially produced Zr-2.5Nb fuel cladding was biaxially creep tested in and out-reactor to generate data for fuel modeling studies. A creep equation was developed describing the steady-state hoop-creep rate at temperatures between 300 and 500°C. The equation assumes that two mechanisms of creep operate at low and high stresses and that the rates of these are additive. The results show little effect of a fast neutron flux of 5 × 1017 neutrons (n)/m2/s on creep rate at 400°C and above but an enhancement of about two in the creep rate at 320°C. The biaxial creep of Zr-2.5Nb fuel cladding is about ten times more rapid than that of pressure-tube materials of the same composition. Texture and second-phase distribution are considered to be the causes of the differences in behavior.

    Some measurements also have been made on irradiation growth of fuel cladding in the longitudinal direction. These are discussed.


    zirconium, zirconium alloys, nuclear fuel claddings, creep properties, irradiation, internal pressure, texture, growth

    Author Information:

    Kohn, E
    Engineer, Westinghouse Canada LimitedAtomic Energy of Canada Limited, Whiteshell Nuclear Research Establishment (WNRE), Hamilton Ontario,

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP35583S