Abstract
Austenitic stainless steel Alloy 709 is a fairly recent alloy developed for high temperature structural applications in power plants. It exhibits excellent creep strength, corrosion resistance and it is being considered alongside other steels as a potential material of choice in structural components for Very High Temperature Gas Cooled Reactors, Sodium Cooled Fast Reactors and Fluoride Salt Cooled High Temperature Reactors. It is essential for the nuclear engineers to have access to detailed databases and predictive computational models for all important properties of Alloy 709, including creep-fatigue crack growth behavior, in order to compare its pros and cons versus other materials. This will allow for informed decisions to be made concerning selection of appropriate materials for nuclear reactor applications. The proposed project aims to address the research needs outlined in Department of Energy programmatic documents. Specifically, the project objectives are: To measure and report creep-fatigue crack growth rates in Alloy 709 under various creep-fatigue loading in accelerated test conditions and to use this knowledge to predict the extended service life of structural components. To measure and report the creep-fatigue crack growth rates in Alloy 709 in as-received and aged conditions, and to identify extrapolation rules between these conditions. To develop and provide a flaw evaluation procedure for creep-fatigue service conditions. The flaw evaluation technique will be validated using predictive computational simulations and experimental testing for validation on several typical crack growth specimens.