High-temperature solution annealing as an IASCC mitigation technique

Conference · Tue Jan 01 00:00:00 EST 1991

OSTI ID:5736866

Jacobs, A J ^[1]; Shepherd, C M ^[2]; Bell, G E.C. ^[3]; Wozadlo, G P ^[4]

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1. General Electric Co., San Jose, CA (USA)
2. UKAEA Harwell Lab. (UK)
3. Oak Ridge National Lab., TN (USA)
4. General Electric Co., Pleasanton, CA (USA). Vallecitos Nuclear Center

Pre-irradiation solution annealing treatments for limited times in the temperature range from 2200{degrees}F (1204{degrees}C) to 2400{degrees}F (1316{degrees}C) were shown to be effective in eliminating irradiation assisted stress corrosion cracking (IASCC) in Type 304 stainless steel (SS), which had been irradiated to fluences between 2.58 and 3.08 {times} 10{sup 21} n/cm{sup 2} (E > 1 MeV). Varying resistance to IASCC as a function of heat treatment parameters was demonstrated in constant extension rate tensile (CERT) tests performed in boiling water reactor (BWR) simulated water. Measures of IASCC susceptibility used in the CERT tests (% IGSCC, % elongation, and maximum stress) could not be correlated with data obtained from HNO{sub 3}/Cr{sup +6} corrosion tests, or from Auger and analytical electron microscopy (AEM) analyses of grain boundary composition. Corrosion test data, however, could be correlated with the Auger and AEM results. The absence of a correlation between IASCC susceptibility and the grain boundary contents of Si, P, and S suggests that other impurities that are known to segregate and undergo nuclear transmutations, such as N and B, may play an important role in the IASCC mechanism. Tensile test data showed that the high temperature solution annealed (HTSA) material had undergone less radiation strengthening than mill annealed (MA) material. Elongations of approximately 10% were measured in the HTSA material compared to 2% in the MA material. 13 refs., 24 figs., 10 tabs.

• Research Organization: Oak Ridge National Lab., TN (USA)
• Sponsoring Organization: USDOE; USDOE, Washington, DC (USA)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 5736866
• Report Number(s): CONF-910808-1; ON: DE91012220
• Resource Relation: Conference: 5. international symposium on environmental degradation on materials in nuclear power systems - water reactors, Monterey, CA (USA), 25-29 Aug 1991
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
STAINLESS STEEL-304
ANNEALING
STRESS CORROSION
AUGER ELECTRON SPECTROSCOPY
BORON
CHROMIUM IONS
CRACK PROPAGATION
ELECTRON MICROSCOPY
ELONGATION
EXPERIMENTAL DATA
GRAIN BOUNDARIES
HEAT TREATMENTS
IRRADIATION
IRRADIATION PROCEDURES
MITIGATION
NITRIC ACID
NITROGEN
RADIATION EFFECTS
SEGREGATION
TENSILE PROPERTIES
TRANSMUTATION
VERY HIGH TEMPERATURE
ALLOYS
AUSTENITIC STEELS
CHARGED PARTICLES
CHEMICAL REACTIONS
CHROMIUM ALLOYS
CHROMIUM-NICKEL STEELS
CORROSION
CORROSION RESISTANT ALLOYS
CRYSTAL STRUCTURE
DATA
ELECTRON SPECTROSCOPY
ELEMENTS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HIGH ALLOY STEELS
HYDROGEN COMPOUNDS
INFORMATION
INORGANIC ACIDS
IONS
IRON ALLOYS
IRON BASE ALLOYS
MATERIALS
MECHANICAL PROPERTIES
MICROSCOPY
MICROSTRUCTURE
NICKEL ALLOYS
NONMETALS
NUMERICAL DATA
SEMIMETALS
SPECTROSCOPY
STAINLESS STEELS
STEEL-CR19NI10
STEELS
360103* - Metals & Alloys- Mechanical Properties
360106 - Metals & Alloys- Radiation Effects
360101 - Metals & Alloys- Preparation & Fabrication