Environmentally assisted cracking in Light Water Reactors. Volume 16: Semiannual report, October 1992--March 1993
This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.
- Research Organization:
- US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Div. of Engineering; Argonne National Lab. (ANL), Argonne, IL (United States)
- Sponsoring Organization:
- Nuclear Regulatory Commission, Washington, DC (United States)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 10187623
- Report Number(s):
- NUREG/CR-4667-Vol.16; ANL-93/27-Vol.16; ON: TI94000758; TRN: 93:023024
- Resource Relation:
- Other Information: PBD: Sep 1993
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
36 MATERIALS SCIENCE
PIPES
FATIGUE
CRACKS
PRESSURE VESSELS
REACTOR COMPONENTS
BWR TYPE REACTORS
PWR TYPE REACTORS
MICROSTRUCTURE
AGING
RADIATION EFFECTS
CRACK PROPAGATION
FRACTURE MECHANICS
STRESS CORROSION
STEEL-ASTM-A106
STEEL-ASTM-A533-B
PROGRESS REPORT
210100
210200
360103
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
BOILING WATER COOLED
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
MECHANICAL PROPERTIES