Segregation in a neutron-irradiated type 316 stainless steel
- Oak Ridge National Lab., TN (United States)
- Westinghouse Electric Corp., Pittsburgh, PA (United States). Science and Technology Center
Radiation-induced segregation (RIS) and associated irradiation- assisted stress corrosion cracking (IASCC) of austenitic alloys may be a major factor in limiting component lifetimes in water-cooled nuclear reactors. There are some similarities between radiation- induced sensitization/IASCC and thermally-induced sensitization/intergranular stress corrosion cracking. Both processes are associated with chromium depletion at grain boundaries. Segregation to boundaries in a neutron irradiated type 316 stainless steel has been investigated with both energy-dispersive X-ray spectrometry (EDXS) and parallel detection electron energy loss spectrometry (PEELS). This report discusses work on segregation in a neutron-irradiated SS-316 sample.
- Research Organization:
- Oak Ridge National Lab., TN (United States)
- Sponsoring Organization:
- DOE; USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC05-84OR21400; AC05-76OR00033
- OSTI ID:
- 5246524
- Report Number(s):
- CONF-920819-10; ON: DE92011039
- Country of Publication:
- United States
- Language:
- English
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36 MATERIALS SCIENCE
360105 -- Metals & Alloys-- Corrosion & Erosion
360106 -- Metals & Alloys-- Radiation Effects
ALLOYS
AUSTENITIC STEELS
CHEMICAL REACTIONS
CHROMIUM
CHROMIUM ALLOYS
CHROMIUM-NICKEL STEELS
CHROMIUM-NICKEL-MOLYBDENUM STEELS
CORROSION
CORROSION RESISTANT ALLOYS
CRYSTAL STRUCTURE
ELEMENTS
GRAIN BOUNDARIES
HEAT RESIS
HIGH ALLOY STEELS
IRON ALLOYS
IRON BASE ALLOYS
METALS
MICROSTRUCTURE
MOLYBDENUM
MOLYBDENUM ALLOYS
NICKEL ALLOYS
PHYSICAL RADIATION EFFECTS
RADIATION EFFECTS
REACTOR COMPONENTS
REACTORS
SEGREGATION
STAINLESS STEEL-316
STAINLESS STEELS
STEEL-CR17NI12MO3
STEELS
STRESS CORROSION
TRANSITION ELEMENTS
WATER COOLED REACTORS