Irradiation performance of 9--12 Cr ferritic/martensitic stainless steels and their potential for in-core application in LWRs
Ferritic-martensitic stainless steels exhibit radiation stability and stress corrosion resistance that make them attractive replacement materials for austenitic stainless steels for in-core applications. Recent radiation studies have demonstrated that 9% Cr ferritic/martensitic stainless steel had less than a 30C shift in ductile-to-brittle transition temperature (DBTT) following irradiation at 365C to a dose of 14 dpa. These steels also exhibit very low swelling rates, a result of the microstructural stability of these alloys during radiation. The 9 to 12% Cr alloys to also exhibit excellent corrosion and stress corrosion resistance in out-of-core applications. Demonstration of the applicability of ferritic/martensitic stainless steels for in-core LWR application will require verification of the irradiation assisted stress corrosion cracking behavior, measurement of DBTT following irradiation at 288C, and corrosion rates measurements for in-core water chemistry.
- Research Organization:
- Pacific Northwest Lab., Richland, WA (United States)
- Sponsoring Organization:
- USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC06-76RL01830
- OSTI ID:
- 10189971
- Report Number(s):
- PNL-SA-22538; CONF-930825-10; ON: DE94001675; TRN: 93:025557
- Resource Relation:
- Conference: 6. international symposium on environmental degradation of materials in nuclear power systems: water reactors,San Diego, CA (United States),1-5 Aug 1993; Other Information: PBD: Aug 1993
- Country of Publication:
- United States
- Language:
- English
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