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.
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