Development of new cladding types for nuclear fuel: IOP Conference Series: Materials Science and Engineering

Zoltán Hózer, T. Novotny, E. Perez-Feró, M. Horváth, P. Szabó, L. Illés, Marc Schyns, Rémi Delville, D. Kim, M. Sevecek

    Research outputpeer-review


    Three different cladding types were tested for nuclear fuel in traditional light water reactors and generation IV gas-cooled fast reactors. Cr coated Zr cladding was tested in steam atmosphere up to 1200 °C to demonstrate moderate oxidation and hydrogen production in accident conditions. 15-15Ti stainless steel alloy and SiCf/SiC cladding tube samples were treated in helium atmosphere with different impurities for several hours at 1000 °C. Additional mechanical testing and microstructure examinations were carried out with as-received samples and with specimens after high temperature treatments. The experiments results indicated the applicability of the tested materials for reactor conditions in the investigated range of parameters.
    Original languageEnglish
    Title of host publicationIOP Conference Series: Materials Science and Engineering
    Subtitle of host publication12th Hungarian Conference on Materials Science (HMSC12) 13-15 October 2019, Balatonkenese, Hungary
    Number of pages9
    StatePublished - 4 Dec 2020

    Publication series

    NameIOP Conference Series: Materials Science and Engineering
    PublisherIOP Publishing Ltd
    ISSN (Print)1757-899X

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