Corrosion fatigue of alloys 600 and 690 in simulated LWR environments

Corrosion fatigue of alloys 600 and 690 in simulated LWR environments [electronic resource]

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 temperatu...

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Bibliographic Details
Online Access: Online Access
Corporate Authors: Argonne National Laboratory (Researcher), United States. Department of Energy. Office of Scientific and Technical Information (Researcher)
Format: Government Document Electronic eBook
Language:English
Published: Oak Ridge, Tenn. : Oak Ridge, Tenn. : United States. Dept. of Energy. Office of Scientific and Technical Information ; distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy, 1996.
Subjects:
Bwr Type Reactors.
Reactor Materials.
Pwr Type Reactors.
Inconel 600.
Crack Propagation.
Corrosion Fatigue.
Inconel 690.
Stainless Steel-304.
Chromium-nickel-molybdenum Steels.
Temperature Range 0400-1000 K.
Temperature Dependence.
Stress Intensity Factors.
Water Chemistry.
Data.
Microstructure.
Materials Science ;21 Nuclear Power Reactors And Associated Plants.

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Call Number: E 1.99:ANL--95/37
E 1.99:ANL--95/37 Available

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