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Oxidation of Zircaloy-4 in steam-nitrogen mixtures at 600-1200°C

Steinbrueck, Martin; Grosse, Mirco; Oliveira da Silva, Fabio

Abstract (englisch):
Zirconium alloys are worldwide used as cladding materials for the fuel in light water reactors due to their low neutron absorption and excellent mechanical and corrosion properties at operational conditions. However, at higher temperatures relevant for nuclear accident scenarios the oxidation of zirconium becomes severe, thus impairing the barrier effect of the claddings against the release of fission products and being the main source for the release of hydrogen and chemical heat. A significantly accelerating effect of nitrogen on the oxidation kinetics has been observed in the framework of studies on air ingress scenarios during severe accidents in reactors and spent fuel pools. Only very limited information of the influence of nitrogen (used for inertization of BWR containments and as pressurizing gas for emergency core cooling systems) on the oxidation of zirconium in steam-nitrogen mixtures were available.
Isothermal oxidation tests with 1-cm cladding segments of Zircaloy-4 (Zr-1.5%Sn) in steam-nitrogen mixtures have been performed at 600, 800, 1000, and 1200°C. All experiments were conducted in a NETZSCH STA-409 thermal balan ... mehr


Zugehörige Institution(en) am KIT Institut für Angewandte Materialien - Angewandte Werkstoffphysik (IAM-AWP)
Publikationstyp Proceedingsbeitrag
Jahr 2016
Sprache Englisch
Identifikator ISBN: 978-0-89448-725-5
KITopen ID: 1000062554
HGF-Programm 32.02.11; LK 01
Erschienen in Proceedings of the 2016 International Congress on Advances in Nuclear Power Plants, ICAPP2016, American Nuclear Society
Verlag American Nuclear Society, LaGrange Park, Illinois
Seiten 535-545
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