Growth of IGSC cracks in Type 304 stainless steel at 100 degrees C in an aqueous environment
Intergranular stress corrosion (IGSC) cracking has been observed in the primary coolant system of the Savannah River Site Reactors. There have been several cases during the over one hundred reactor-years of plant operating experience when IGSC cracks have grown through-wall and minor leaks have occurred. Approximately 7% of the heat affected zones of pipe-to-pipe butt welds show indications of IGSC cracking during ultrasonic testing (UT). Other piping and component areas, sensitized by flame washing or hot forming, have also developed IGSC cracks. The entire system was fabricated in the 1950's from Type 304 stainless steel. All joining was by the metal inert gas welding process. IGSC crack growth rates have been measured on compact tension specimens under controlled environmental conditions that encompass the observed conditions in the SRS reactor primary coolant systems. Growth rates were measured extending from less than 10{sup {minus}9} to approximately 10{sup {minus}5} millimeter per second. These growth rates bound the growth rates that have been inferred from a statistical analysis of UT indications. The UT data were collected since 1984 from weld heat affected zones in pipe-to-pipe butt welds in the SRS reactor primary coolant piping. Chloride and sulfate anions, dissolved oxygen, and peroxide have been identified as the water impurities that influence IGSC cracking in the SRS reactor primary coolant systems. A quantitative relationship has been established for susceptibility to IGSC cracking in terms of concentrations of these impurities and temperature. The heavy water reactor moderator and coolant is acidified with nitric acid to a pH of 4.7 to minimize corrosion of the aluminum cladding on the fuel elements.
- Research Organization:
- Westinghouse Savannah River Co., Aiken, SC (United States)
- Sponsoring Organization:
- USDOE; USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC09-89SR18035
- OSTI ID:
- 5667220
- Report Number(s):
- WSRC-MS-90-296; CONF-910808-8; ON: DE92009446
- Resource Relation:
- Conference: 5. international symposium on environmental degradation of materials in nuclear power systems - water reactors, Monterey, CA (United States), 25-29 Aug 1991
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
36 MATERIALS SCIENCE
PIPES
STRESS CORROSION
PRIMARY COOLANT CIRCUITS
SAVANNAH RIVER PLANT
STAINLESS STEEL-304
CHLORIDES
CRACK PROPAGATION
HEAT AFFECTED ZONE
INTERGRANULAR CORROSION
PRODUCTION REACTORS
SULFATES
ALLOYS
AUSTENITIC STEELS
CHEMICAL REACTIONS
CHLORINE COMPOUNDS
CHROMIUM ALLOYS
CHROMIUM-NICKEL STEELS
COOLING SYSTEMS
CORROSION
CORROSION RESISTANT ALLOYS
HALIDES
HALOGEN COMPOUNDS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HIGH ALLOY STEELS
IRON ALLOYS
IRON BASE ALLOYS
MATERIALS
NATIONAL ORGANIZATIONS
NICKEL ALLOYS
OXYGEN COMPOUNDS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
STAINLESS STEELS
STEEL-CR19NI10
STEELS
SULFUR COMPOUNDS
US AEC
US DOE
US ERDA
US ORGANIZATIONS
ZONES
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors
360105 - Metals & Alloys- Corrosion & Erosion