Deformation analysis of SiC-SiC channel box for BWR applications

G. Singh, J. Gorton, D. Schappel, N. R. Brown, Y. Katoh, B. D. Wirth, K. A. Terrani

    Research output: Contribution to journalArticlepeer-review

    18 Scopus citations


    Silicon carbide fiber-reinforced silicon carbide matrix (SiC-SiC) composites are being considered as components in light water reactor cores to improve accident tolerance, including channel boxes and fuel cladding. In the nuclear reactor environment, core components like a channel box will be exposed to neutron and other radiation damage and temperature gradients. To ensure reliable and safe operation of a SiC-SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperatures on the deformation behavior of the channel box over the course of one year. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions that have been calculated using the neutronics and thermal-hydraulics codes Serpent and CTF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5 mm. The channel box bowing behavior is time dependent and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.

    Original languageEnglish (US)
    Pages (from-to)71-85
    Number of pages15
    JournalJournal of Nuclear Materials
    StatePublished - Jan 2019

    All Science Journal Classification (ASJC) codes

    • Nuclear and High Energy Physics
    • General Materials Science
    • Nuclear Energy and Engineering


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