CEA experience on alloy X-750 and alternate materials (alloy 718 and A-286) after 27,000 hours testing of actual pins and tensile specimens
Description
The stress corrosion cracking behavior of high-strength alloys X-750, 718, A-286 and Uranus 50 used in reactor pressure vessel internal applications was evaluated using bent beam, U-bend, and constant load tensile tests. This testing indicated that A-286 material has the best resistance to SCC in 350 degree C water. A-286 material took 10 times longer to crack than alloy X-750. Constant load tensile tests produced no failures of alloys X-750 or 718 after 35,000 hours of exposure when specimens were loaded to less than 0.8 and 1.0 yield strength, respectively. Constant load tensile tests (0.8 to 1.2 yield strength) of alloy X-750 in 350 degree C primary water plus hydrogen indicated improved resistance to SCC when solution annealed at 1090 degree C (2000 degree F) and aged at 704 degree C (1,300 degree F). The SCC resistance of alloy X-750 was improved by intermediate cold drawing, water quenching, and zirconium additions. Constant extension rate tests were performed on alloy X-750 in a pure water and simulated primary water environment with hydrogen at 300 degree C to 360 degree C. The materials tested annealed at 1,090 degree C for 1 hour and single aged at 700 degree C for 20 hours produced no SCC. Alloy X-750 annealed at 885 degree C produced severe intergranular stress corrosion cracking (IGSCC). Alloys X-750 and 718 subjected to various thermal treatments were fatigue tested in air at 300 degree C and pure water at 320 degree C. The fatigue strength of both alloys improved when solution annealed at lower temperatures. Alloy X-750 solution annealed at 1,090 degree C and aged at 704 degree C had the best resistance to SCC. Design changes made to mitigate IGSCC include a parabolic root radius, low torque, and polishing
Additional details
Publishing Information
- Imprint Title
- Proceedings: 1986 workshop on advanced high-strength materials
- Imprint Pagination
- 578 p.
- Journal Page Range
- p. 16.1-16.40.
- Report number
- EPRI-NP--6363
Conference
- Title
- Workshop on advanced high-strength materials.
- Dates
- 12-13 Mar 1986.
- Place
- Clearwater Beach, FL (USA).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 21036773
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S36: MATERIALS SCIENCE; S36: MATERIALS SCIENCE;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- ALLOY-NI53CR19FE19NB5MO3; ALLOY-NI73CR15FE7TI3; ANNEALING; CORROSION RESISTANCE; CRACKS; FASTENERS; FATIGUE; HEAT TREATMENTS; HIGH TEMPERATURE; INTERGRANULAR CORROSION; MATERIALS TESTING; MITIGATION; PRESSURE VESSELS; PRIMARY COOLANT CIRCUITS; PWR TYPE REACTORS; REACTOR MATERIALS; STAINLESS STEELS; STRESS CORROSION; TEST FACILITIES; VERY HIGH TEMPERATURE; YIELD STRENGTH; ZIRCONIUM ADDITIONS
- Descriptors DEC
- ALLOYS; ALUMINIUM ADDITIONS; CARBON ADDITIONS; CHEMICAL REACTIONS; CHROMIUM ALLOYS; CONTAINERS; COOLING SYSTEMS; CORROSION; CORROSION RESISTANT ALLOYS; ENRICHED URANIUM REACTORS; HEAT RESISTING ALLOYS; HIGH ALLOY STEELS; INCONEL ALLOYS; IRON ALLOYS; IRON BASE ALLOYS; MATERIALS; MECHANICAL PROPERTIES; MOLYBDENUM ALLOYS; NICKEL ALLOYS; NICKEL BASE ALLOYS; NIOBIUM ADDITIONS; NIOBIUM ALLOYS; POWER REACTORS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; STEELS; TESTING; THERMAL REACTORS; TITANIUM ADDITIONS; TITANIUM ALLOYS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-8603261--.