STP754: Effects of Temperature and Pressure on the In-Reactor Creepdown of Zircaloy Fuel Cladding

    Hobson, DO
    Metallurgist, Metals and Ceramics Division, nuclear engineer, Engineering and Technology Division, and physicist, Metals and Ceramics Division, Oak Ridge National Laboratory, Oak Ridge, Tenn.

    Thoms, KR
    Metallurgist, Metals and Ceramics Division, nuclear engineer, Engineering and Technology Division, and physicist, Metals and Ceramics Division, Oak Ridge National Laboratory, Oak Ridge, Tenn.

    Dodd, CV
    Metallurgist, Metals and Ceramics Division, nuclear engineer, Engineering and Technology Division, and physicist, Metals and Ceramics Division, Oak Ridge National Laboratory, Oak Ridge, Tenn.

    van der Kaa, T
    Engineer, Stichting Energieonderzoek Centrum Nederland, Petten, North Holland,

    Pages: 20    Published: Jan 1982


    Abstract

    Descriptions and results for seven of the eight in-reactor creepdown tests of Zircaloy fuel cladding, which were part of a joint program between the U.S. Nuclear Regulatory Commission and Energieonderzoek Centrum Nederland, are presented. These tests were conducted to study the behavior of Zircaloy fuel cladding under conditions that approximate those found in an operating pressurized-water power reactor.

    Radial surface displacement values as functions of time, average diametral-circumferential strain as a function of time, and isochronal deformation surfaces are presented. Tests were conducted at 343°C with external pressures from 13.3 to 18.7 MPa. Three of the specimens were subjected to stress reversal during their test runs, wherein they were pressurized internally from 3.5 to 6.9 MPa. Fast flux [E > 1.0 MeV (0.2 pJ)] ranged from 3.8 × 1017 to 5.0 × 1017 neutrons (n)/(m2·2). Maximum fluence was approximately 7 × 1025 n/m2.

    The most important conclusion to be drawn from this study involves the deformation of the cladding during testing. Contrary to similar tests conducted out-of-reactor, the in-reactor specimens did not deform uniformly, that is, by diametral contraction and smooth ovalization. Rather, the deformation surfaces were nonuniform with hills and valleys being formed at irregular intervals. This implies that conventional concepts of creep rate and simplified modeling procedures will not work for predicting cladding behavior. Sufficient data have been generated in this program to supply modelers with detailed descriptions of the cladding surface shapes from which new interpretations can be derived to predict cladding behavior.

    Keywords:

    zirconium, nuclear industry, creep properties, irradiation, nuclear fuel cladding, deformation


    Paper ID: STP37054S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP37054S


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