Abstract
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.
Details
Original language | English |
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Title of host publication | IOP Conference Series: Materials Science and Engineering |
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Subtitle of host publication | 12th Hungarian Conference on Materials Science (HMSC12) 13-15 October 2019, Balatonkenese, Hungary |
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Pages | 1-9 |
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Number of pages | 9 |
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Volume | 903 |
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DOIs | |
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Publication status | Published - 4 Dec 2020 |
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Name | IOP Conference Series: Materials Science and Engineering |
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Publisher | IOP Publishing Ltd |
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Volume | 903 |
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ISSN (Print) | 1757-899X |
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