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U.S. Department of Energy
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Corrosion fatigue characterization of reactor pressure vessel steels. [PWR; BWR]

Technical Report ·
OSTI ID:6821903

During routine operation, light water reactor (LWR) pressure vessels are subjected to a variety of transients that result in time-varying stresses. Consequently, fatigue and environmentally-assisted fatigue are mechanisms of growth relevant to flaws in these pressure vessels. To provide a better understanding of the resistance of nuclear pressure vessel steels to these flaw growth processes, fracture mechanics data were generated on the rates of fatigue crack growth for SA508-2 and SA533B-1 steels in both room temperature air and 288/sup 0/C water. Areas investigated were: the relationship of crack growth rate to prior loading history; the effects of loading frequency and R ratio (K/sub min//K/sub max/) on crack growth rate as a function of the stress intensity factor range (..delta..K); transient aspects of the fatigue crack growth behavior; the effect of material chemistry (sulphur content) on fatigue crack; and growth rate; water chemistry effects (high-purity water versus simulated pressurized water reactotr (PWR) primary coolant).

Research Organization:
Babcock and Wilcox Co., Alliance, OH (USA). Research and Development Div.
OSTI ID:
6821903
Report Number(s):
EPRI-NP-2775; ON: DE83901109
Country of Publication:
United States
Language:
English