Evaluation of SCC test methods for Inconel 600 in low temperature aqueous solutions
Conference
·
OSTI ID:6168778
In late 1981, widespread leakage was encountered in Alloy 600 steam-generator tubing at the Three Mile Island Unit 1 nuclear power plant. The phenomenon was identified as low-temperature intergranular stress-corrosion cracking (SCC) initiated from the inner surfaces of the tubes exposed to the primary coolant. A testing program was initiated to examine the material and environmental factors relevant to these failures, which were found to be associated with sensitization of the material and contamination of the coolant by air and sodium thiosulfate. The test solutions contained 1.3% boric acid with various additions of sulfur compounds and lithium hydroxide. Constant extension rate testing was used as the primary tool to examine environmental effects such as the inhibition of cracking by lithium hydroxide. Important effects of crack-initiation frequency on the specimen potential (and therefore crack velocity) are demonstrated.
- Research Organization:
- Brookhaven National Lab., Upton, NY (USA)
- DOE Contract Number:
- AC02-76CH00016
- OSTI ID:
- 6168778
- Report Number(s):
- BNL-NUREG-32798; CONF-8204128-1; ON: DE83011358
- Country of Publication:
- United States
- Language:
- English
Similar Records
Evaluation of SCC test methods for Inconel 600 in low-temperature aqueous solutions
Electrochemical and metallurgical aspects of stress corrosion cracking of sensitized Alloy 600 in simulated primary water containing sulfur contamination
Investigation of the sulfur and lithium to sulfur ratio threshold in stress corrosion cracking of sensitized alloy 600 in borated thiosulfate solution
Journal Article
·
Sat Dec 31 23:00:00 EST 1983
· Am. Soc. Test. Mater., Spec. Tech. Publ.; (United States)
·
OSTI ID:6180421
Electrochemical and metallurgical aspects of stress corrosion cracking of sensitized Alloy 600 in simulated primary water containing sulfur contamination
Conference
·
Mon Dec 31 23:00:00 EST 1984
·
OSTI ID:6467767
Investigation of the sulfur and lithium to sulfur ratio threshold in stress corrosion cracking of sensitized alloy 600 in borated thiosulfate solution
Technical Report
·
Sun Jul 01 00:00:00 EDT 1984
·
OSTI ID:6932038
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200* -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
36 MATERIALS SCIENCE
360105 -- Metals & Alloys-- Corrosion & Erosion
ALLOYS
BOILERS
CHEMICAL REACTIONS
CHEMISTRY
CHROMIUM ALLOYS
CORROSION
CORROSION INHIBITORS
CRACKS
ENRICHED URANIUM REACTORS
INCONEL 600
INCONEL ALLOYS
INTERGRANULAR CORROSION
IRON ALLOYS
LEAKS
MATERIALS
NICKEL ALLOYS
NICKEL BASE ALLOYS
NIOBIUM ALLOYS
POWER REACTORS
PWR TYPE REACTORS
REACTOR MATERIALS
REACTORS
STEAM GENERATORS
STRESS CORROSION
THERMAL REACTORS
THREE MILE ISLAND-1 REACTOR
TUBES
VAPOR GENERATORS
WATER CHEMISTRY
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200* -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
36 MATERIALS SCIENCE
360105 -- Metals & Alloys-- Corrosion & Erosion
ALLOYS
BOILERS
CHEMICAL REACTIONS
CHEMISTRY
CHROMIUM ALLOYS
CORROSION
CORROSION INHIBITORS
CRACKS
ENRICHED URANIUM REACTORS
INCONEL 600
INCONEL ALLOYS
INTERGRANULAR CORROSION
IRON ALLOYS
LEAKS
MATERIALS
NICKEL ALLOYS
NICKEL BASE ALLOYS
NIOBIUM ALLOYS
POWER REACTORS
PWR TYPE REACTORS
REACTOR MATERIALS
REACTORS
STEAM GENERATORS
STRESS CORROSION
THERMAL REACTORS
THREE MILE ISLAND-1 REACTOR
TUBES
VAPOR GENERATORS
WATER CHEMISTRY
WATER COOLED REACTORS
WATER MODERATED REACTORS