Abstract: The methodology of fatigue crack growth analysis in evaluating structural integrity of nuclear components has been well established over the years, even to the point where a recommended practice ...

Abstract: Fatigue crack-growth behavior was investigated for types 304 and 316 stainless steel exposed to a pressurized water reactor environment. The effects of test frequency, stress ratio, specimen ...

Abstract: Reference fatigue crack growth rate curves have been contained in Appendix A of Section XI since its inception. The curves have been designed to be applicable to carbon and low alloy pressure ...

Abstract: The overspeed capability of the large steel flywheels used on light water reactor primary coolant pumps has been evaluated through a combined analytical and experimental effort. Limiting speeds ...

Abstract: A series of experiments were conducted to characterize the effects of high temperature service (550–650°F or 288–343°C) on the mechanical properties of cast stainless steels. The material ...