Slow strain rate tensile testing of fusion spectrally-tailored type 316 material irradiated between 60 and 400C in oxygenated, high-purity water environments
Creators
- 1. Oak Ridge National Lab., TN (United States)
- 2. Japan Atomic Energy Research Inst., Ibaraki (Japan)
Description
The objective of this work is to determine the susceptibility of a fusion-relevant alloy to irradiation-assisted stress corrosion cracking (ISACC) by performing slow strain rate tensile (SSRT) testing of fusion spectrally-tailored materials in aqueous environments. Specimens of a type 316 stainless steel (SS) that had been irradiated in the fusion spectrally-tailored MFE-6J/7J experiments in the Oak Ridge Research (ORR) to 7.4 dpa at temperatures of 400, 330, 200, and 60C were SSRT tested in a recirculating autoclave system in oxygenated, high-purity water. Complete intergranular (IG) failure was observed only for the specimen irradiated at 400C. Only 10% IG failure was observed for the 330C irradiated specimen and no IG failure was observed for the specimens irradiated at 200 and 60C
Additional details
Publishing Information
- Imprint Title
- Fusion Reactor Materials semiannual progress report for period ending September 30, 1991
- Imprint Pagination
- 330 p.
- Journal Page Range
- p. 253-262.
- Report number
- DOE/ER--0313/11
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 24031636
- Subject category
- S70: PLASMA PHYSICS AND FUSION TECHNOLOGY; S36: MATERIALS SCIENCE; S36: MATERIALS SCIENCE;
- Resource subtype / Literary indicator
- Numerical Data, Progress Report
- Descriptors DEI
- AQUEOUS SOLUTIONS; CRACK PROPAGATION; EXPERIMENTAL DATA; FAILURES; INTERGRANULAR CORROSION; MORPHOLOGY; ORR REACTOR; PHYSICAL RADIATION EFFECTS; PROGRESS REPORT; STEEL-CR17NI12MO3; STRAIN RATE; STRESS CORROSION; TEMPERATURE RANGE 0273-0400 K; TEMPERATURE RANGE 0400-1000 K; THERMONUCLEAR REACTOR MATERIAL; WATER
- Descriptors DEC
- ALLOYS; AUSTENITIC STEELS; CARBON ADDITIONS; CHEMICAL REACTIONS; CHROMIUM ALLOYS; CHROMIUM-NICKEL STEELS; CHROMIUM-NICKEL-MOLYBDENUM STE; CORROSION; CORROSION RESISTANT ALLOYS; DATA; DISPERSIONS; DOCUMENT TYPES; ENRICHED URANIUM REACTORS; HEAT RESISTING ALLOYS; HIGH ALLOY STEELS; HOMOGENEOUS MIXTURES; HYDROGEN COMPOUNDS; INFORMATION; IRON ALLOYS; IRON BASE ALLOYS; MATERIALS; MIXTURES; MOLYBDENUM ALLOYS; NICKEL ALLOYS; NUMERICAL DATA; OXYGEN COMPOUNDS; RADIATION EFFECTS; REACTORS; SOLUTIONS; STAINLESS STEELS; STEELS; TANK TYPE REACTORS; TEMPERATURE RANGE; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- ORNL/M--1945.