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 |
A Study on Stressed Ferritic/Martensitic Steel Weldments Exposed to High-temperature Liquid Sodium |
Author(s) |
Dustin Mangus, Ian Arndt, Logan Smith, Caitlin Huotilainen, Guillaume Mignot, Samuel Briggs |
On-Site Speaker (Planned) |
Dustin Mangus |
Abstract Scope |
The combination of applied stress and liquid metal corrosion can potentially accelerate degradation of weldments in the high-temperature regions of sodium-cooled fast reactor systems. Oregon State University has developed the Glovebox for Experimental Liquid Sodium facility, enabling in situ materials testing in simulated sodium-cooled reactor environments. In this study, weldments of commercial ferritic/martensitic steels, including Grade 92 and HT9, have been tested using static flexure (triple-point bend testing) and load-controlled tensile tests during exposure to low oxygen liquid sodium. A parametric study of applied stress will be used to assess the susceptibility of cracking in the heat-affected zone of butt-type welded joints. Selected welding techniques employed include conventional gas tungsten arc welding and cold metal transfer to assess how weld heat input affects weld cracking behavior. This investigation on liquid sodium environmental effects on weldments will provide strategic information to support commercialization and licensing next-generation, sodium-cooled reactor systems. |
Proceedings Inclusion? |
Planned: |
Keywords |
Nuclear Materials, High-Temperature Materials, Environmental Effects |