RPS2

    Mechanistic Modeling of Zircaloy Deformation and Fracture in Fuel Element Analysis

    Published: Jan 2010

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    Abstract

    A review is given of the comprehensive model developed in the 1960s at the Bettis Atomic Power Laboratory to explain the creep of Zircaloy during neutron irradiation and applied to fuel to element analysis and design. The in-pile softening observed at low stresses was hypothesized to be due to a combination of the growth-directed Roberts-Cottrell yielding creep originally proposed for α-uranium and the formation of point defect loops preferentially on certain planes in response to the applied stress, with the second process being of relatively greater importance. The in-pile hardening observed at high stresses (or strain-rates) was proposed to be due to the cutting by dislocations of radiation-produced obstacles. In this stress (strain-rates) region, in-pile behavior was proposed to be identical to post-irradiation behavior. At intermediate stresses (strain-rates) a mechanism of radiation-enhanced climb around obstacles was suggested as being rate-controlling. As the stress is decreased, the climb process becomes easier, and the rate was then predicted to be controlled by glide at a flow-stress characteristic of unirradiated, annealed material, where radiation-enhanced diffusion enabled climbing around the normal strain-hardening obstacles. At still lower stresses, this glide process became negligibly slow compared with the growth-connected creep mechanism that was presumed to operate independently. The overall scheme was shown to be in good agreement with all the in-pile data then available and implemented into the computer analysis of fuel element behavior. The original paper was published by ASTM International in STP 939, Zirconium in the Nuclear Industry: Seventh International Symposium, 1987, pp. 5–22.

    Keywords:

    zirconium, Zircaloy-2, neutron irradiation, irradiation growth, irradiation creep, stress, temperature, damage rate, dose, strain, strain rate, texture, mechanical properties, pressurized water reactor, deformation, modelling, nuclear fuel cladding, fracture


    Author Information:

    Nichols, F. A.
    Senior Scientist and Section Manager, Argonne National Laboratory, Argonne, IL


    Paper ID: MNL12107R

    Committee/Subcommittee: B10.02

    DOI: 10.1520/MNL12107R


    CrossRef ASTM International is a member of CrossRef.

    ISBN10:
    ISBN13: 978-0-8031-7018-6