About this Abstract |
| Meeting |
2026 TMS Annual Meeting & Exhibition
|
| Symposium
|
Interrelated Extremes in Materials Degradation for Fission and Fusion Environments
|
| Presentation Title |
Assessment of Brittle Fracture of 304L Austenitic Stainless Steel – Material of Light Water Reactor Core Internals |
| Author(s) |
Maxim Gussev, T. Lach, S. Kang, M. Ickes, C. Cmar, G. Was, F. Garner, B. Ensign, X. Chen |
| On-Site Speaker (Planned) |
Maxim Gussev |
| Abstract Scope |
A specific phenomenon—brittle fracture during tensile deformation at ambient temperature—was observed during post-irradiation evaluation of 304L steel samples harvested from baffle plate and baffle former assemblies of a decommissioned light water reactor. This phenomenon occurred at relatively low damage doses, below 12–20 dpa, a range previously considered safe. Fracture during deformation developed without visible necking and was dominated by intergranular cracking. This observation suggests the presence of a novel, under-explored form of irradiation embrittlement, potentially introducing substantial additional risks during reactor refueling or maintenance operations at ambient temperatures.
The present study discusses the results of tensile tests performed within a temperature range from room temperature to 320 °C, accompanied by fractographic analyses and microstructural assessments using transmission electron microscopy and electron backscatter diffraction. Preliminary measurements of hydrogen and helium content are also presented. Potential mechanisms and driving forces behind this phenomenon are considered. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Other, Nuclear Materials, Mechanical Properties |