Corrosion fatigue of alloys 600 and 690 in simulated LWR environments
- Argonne National Lab., IL (United States)
Crack growth data were obtained on fracture-mechanics specimens of Alloys 600 and 690 to investigate environmentally assisted cracking (EAC) in simulated boiling water reactor and pressurized water reactor environments at 289 and 320 C. Preliminary information was obtained on the effect of temperature, load ratio, stress intensity (K), and the dissolved-oxygen and -hydrogen concentrations of the water on EAC. Specimens of Type 316NG and sensitized Type 304 stainless steel (SS) were included in several of the experiments to assess the behavior of these materials and Alloy 600 under the same water chemistry and loading conditions. The experimental data are compared with predictions from an Argonne National Laboratory (ANL) model for crack growth rates (CGRs) of SSs in water and the ASME Code Section 11 correlation for CGRs in air at the K{sub max} and load-ratio values in the various tests. The data for all of the materials were bounded by ANL model predictions and the ASME Section 11 ``air line.``
- Research Organization:
- Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Argonne National Lab., IL (United States)
- Sponsoring Organization:
- Nuclear Regulatory Commission, Washington, DC (United States)
- OSTI ID:
- 219363
- Report Number(s):
- NUREG/CR--6383; ANL--95/37; ON: TI96008966
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
36 MATERIALS SCIENCE
BWR TYPE REACTORS
CHROMIUM-NICKEL-MOLYBDENUM STEELS
CORROSION FATIGUE
CRACK PROPAGATION
DATA
INCONEL 600
INCONEL 690
MICROSTRUCTURE
PWR TYPE REACTORS
REACTOR MATERIALS
STAINLESS STEEL-304
STRESS INTENSITY FACTORS
TEMPERATURE DEPENDENCE
TEMPERATURE RANGE 0400-1000 K
WATER CHEMISTRY