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    A Comparison of Fuel Pin Deformations with Pressurized Tube Creep Tests


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    Fuel pins are a key component in the fast breeder reactor (FBR), and the economical performance of the reactor depends upon the behavior and life of fuel pin assemblies. One of the limiting phenomena in fuel pin and assembly performance is deformation. A thermomechanical performance code “SIFAIL” is being used at the Hanford Engineering Development Laboratory (HEDL) to predict fuel cladding deformation. SIFAIL utilizes empirical descriptions of cladding strain as a function of temperature, stress, and neutron fluence to predict fuel pin profiles. The equations are derived primarily from in-reactor tests on tubes pressurized with an inert gas and irradiated under conditions of constant temperature and stress. These equations were modified in SIFAIL to account for the variations in temperature and stress with time which occur in fuel pins. It is the purpose of these investigations to develop analytical methods which can be used to design fuel pins with optimum long-life performance.

    Equations were developed to describe the in-reactor creep behavior for two different classes of AISI 316 stainless steel tubing, a developmental class for the fast test reactor (FTR) and the FTR first-core tubing. The behavior of these two tubing classes was sufficiently different that two distinctly separate equations were required to describe their in-reactor creep behaviors.

    The results of the analyses showed several features. First of all, the SIFAIL performance code predicted the experimental fuel cladding strains to within a mean value of +0.03 percent using fission gas pressure as the only loading mechanism. This indicates that mechanical interaction between the fuel pellets and the stainless steel cladding in addition to other loading mechanisms generated less that 0.1 percent mean strain. Secondly, there appeared to be little correlation between creep and swelling. The variability in swelling among fuel pins was not reflected in the variability of in-reactor creep. Thirdly, the use of an in-reactor creep model which relied on a steady-state swelling rate to predict the creep rate (swelling-enhanced creep model) underpredicted the fuel pin cladding strains by approximately 50 percent. Finally, the deformations in fuel pin cladding reflected similar differences between the developmental class and the first-core class as found in the pressurized tube creep results for these two classes of tubing.


    fuel pins, deformation, strain, temperature, stress, neutron fluence, equation, in-reactor creep, fission gas pressure, profilometry, thermal creep, measurements

    Author Information:

    Taylor, JP
    Engineer and senior engineers, Hanford Engineering Development Laboratory, Richland, Washington

    Gilbert, ER
    Engineer and senior engineers, Hanford Engineering Development Laboratory, Richland, Washington

    Lovell, AJ
    Engineer and senior engineers, Hanford Engineering Development Laboratory, Richland, Washington

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP38192S