About this Abstract |
| Meeting |
2026 TMS Annual Meeting & Exhibition
|
| Symposium
|
Materials Corrosion Behavior in Advanced Nuclear Reactor Environments III
|
| Presentation Title |
Effects of Oxygen Contaminated Liquid Sodium on the Degradation of Austenitic and Ferritic/Martensitic Stainless Steels |
| Author(s) |
Tolin O. Skov Black, Dev Chidambaram |
| On-Site Speaker (Planned) |
Tolin O. Skov Black |
| Abstract Scope |
GEN IV nuclear reactor designs promise to produce clean energy, while being passively safe and producing minimal waste. Sodium fast reactors (SFRs), one such design, utilize liquid sodium as the coolant; however, due to its reactivity with moisture and oxygen, leaks are considered significantly dangerous. In previous generation SFRs, austenitic steels were used as both the cladding and structural material but had limited core lifetime due to poor neutron irradiation resistance. Therefore, recent GEN IV designs call for the use of irradiation resistant ferritic/martensitic stainless steels as the cladding. In this work, austenitic and ferritic/martensitic alloys were exposed to varying levels of oxygen polluted static liquid sodium at 550 °C to allow a comparison of the microstructural surface corrosion properties. Described will be the development of the static corrosion testing vessel, mechanism of corrosion, analysis of microstructural surface properties, and estimation of long-term corrosion rates. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, Environmental Effects, Iron and Steel |