Abstract
Because of its resistance to high temperature (>1000°C) attack by steam and good mechanical properties above 400°C, iron-chromium-aluminum alloys (FeCrAl) are considered attractive candidates to replace current zirconium (Zr) based alloys as cladding for light water reactor (LWR) fuels. FeCrAl alloys were sporadically investigated for nuclear applications since the 1950s. Most of the knowledge on these alloys in the literature is for dry environments and temperatures above 1000°C and little information was available for lower temperatures or wet environments. As cladding for LWR, the FeCrAl alloys are expected to survive the nuclear fuel cycle, that it, from manufacturing to used fuel disposal. The presentation will cover the corrosion behavior aspects of FeCrAl from manufacturing using powder metallurgy and welding, to residence in the reactor core for up to ten years (including loss of coolant scenarios), to used fuel residence in cooling pools, to fuel retrievability and reprocessing followed by long term dry cask storage and eventual geologic repository disposal.
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