About this Abstract |
Meeting |
2024 TMS Annual Meeting & Exhibition
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Symposium
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Materials Corrosion Behavior in Advanced Nuclear Reactor Environments
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Presentation Title |
Investigation of Oxidation Behavior on Mo-doped FeCrAl Alloys in Low-temperature (400°C) and High-temperature (1200°C) Steam Environments |
Author(s) |
Haozheng Qu, Hamdy Abouelella, Indranil Roy , Andrew Hoffman, Raul Rebak, Rajnikant Umretiya |
On-Site Speaker (Planned) |
Haozheng Qu |
Abstract Scope |
In this talk, the effect of 0, 1, 3 wt% Mo addition on the oxidation behavior of FeCrAl alloys in both Light Water Reactor (LWR) and Loss of Coolant Action (LOCA) conditions are investigated. FeCrAl alloy systems are emerging as promising candidates for next-generation accident-tolerant fuel (ATF) cladding material to replace traditional Zr-based alloys. However, the elemental effects on the oxidation resistance of FeCrAl alloys remain unclear. Two model Fe-21Cr-5.5Al-xMo (x = 0, 1, 3 wt%) alloys are exposed to 400°C steam (LWR) for 100 hours and 1200°C steam (LOCA) for 2 hours. The corrosion products are systematically characterized and the corrosion mechanism will be discussed in the context of various Mo contents. The results will advance the current understanding of elemental modification on FeCrAl alloy oxidation behavior and provide principal guidance for future FeCrAl ATF cladding alloy design. |
Proceedings Inclusion? |
Planned: |
Keywords |
Nuclear Materials, Environmental Effects, |