Most of the current operating nuclear power plants in the United States were designed using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, for fatigue design curves. These design curves were developed in the late 1960s and early 1970s. They were often referred to as “air curves” because they were based on tests conducted in laboratory air environments at ambient temperatures. In recent years, laboratory fatigue test data showed that the light-water reactor environment could have significant impact on the fatigue life of carbon and low-alloy steels, austenitic stainless steel, and nickel-chromium-iron (Ni-Cr-Fe) alloys. United States Nuclear Regulatory Commission, Regulatory Guide 1.207 provides a guideline for evaluating fatigue analyses incorporating the life reduction of metal components due to the effects of the light-water reactor environment for new reactors. It recommend following the method developed in NUREG/CR-6909 [3] when designing reactor coolant pressure boundary components. The industry has invested a lot of effort in developing methods and rules for applying environmental fatigue evaluations for ASME Class 1 components and piping. However, the industry experience in applying the environmental fatigue evaluation for reactor core support structures and internal structures has been very limited. During the recent aging management programs, reactor internal component environmental fatigue evaluations for several pressurized water reactors were evaluated. The analyses calculated the cumulative fatigue usage using the recorded plant-specific transient cycles and the projected cycles for 60 years of plant life. The study concludes that the actual fatigue usages of the components are substantially lower than the specified original design conditions. Even assuming the most severe light-water reactor coolant environmental effects, fatigue will not be a concern for 60 years of plant life. The experiences with environmental fatigue evaluation for reactor internals are still very limited. This study shall provide the industry with beneficial information to develop the approaches and rules addressing the environmental effect on the fatigue life of reactor internals.
EFFECT OF LIGHT-WATER REACTOR FUEL ROD FAILURE ON THE AREA AVAILABLE FOR EMERGENCY COOLANT FLOW FOLLOWING A LOSS-OF-COOLANT ACCIDENT.
