Fretting Wear Damage of tubular bundles in the presence of flow-induced vibrations at elevated temperatures

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Flow-induced vibrations is one of the main causes of fretting wear, fatigue, and cracking in tubes in nuclear reactors. There have been considerable work efforts in past 4 decades that aimed at predicting fretting wear at the supports in the pressurized water reactor steam generators. These efforts were directed at predicting the flow-induced vibrations response, tube/support impact forces and wear coefficients. Using these quantities (response, normal contact, and wear coefficients) one can predict the wear volume and consequently the age of the component. The fretting wear of typical reactors materials under operating conditions is well characterized. While SMR designs are based on existing deployed technologies, others are based on advanced concepts, suggesting a range of sizes up to 300 MWe and a range of temperatures of more than 850°C. In addition, the environment under which these components will operate are entirely different. Moreover, the proposed materials for the SMR are mainly considered for their pressure retaining characteristics and corrosion. While a reasonable knowledge of strength and corrosion performance is gained for some of the proposed SMR materials such as 800HT, Hastelloyx, etc. the fretting wear characteristics at these materials at high temperatures and different environmental conditions are unknown. This project aims at gaining an understanding of the fretting wear degradation mechanisms and providing guidance on the performance at high temperature with higher confidence. This project addresses the research challenges and knowledge gaps under "Consequences of high temperature on reactor components" in the NSERC-CNSC Small Modular Reactors Research Grant Initiative.