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

微动磨损 高温下流引起振动时管束的损坏

• 批准号: 580454-2022
• 负责人: Hassan, MarwanM
• 金额: $ 5.16万
• 依托单位: University of Guelph
• 依托单位国家: 加拿大
• 项目类别: Alliance Grants
• 财政年份: 2022
• 资助国家: 加拿大
• 起止时间: 2022-01-01 至 2023-12-31
• 项目状态: 已结题
• 来源: https://www.nserc-crsng.gc.ca/ase-oro/Details-Detailles_eng.asp?id=753144
• 关键词: Fretting; Wear; Damage; tubular; bundles

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.

流激振动是核反应堆管道微动磨损、疲劳和裂纹的主要原因之一。 在过去的40年里，人们做了大量的工作来预测压水堆蒸汽发生器支撑件的微动磨损。 这些努力是针对预测流致振动响应，管/支撑冲击力和磨损系数。 使用这些量（响应、法向接触和磨损系数），可以预测磨损量，从而预测部件的使用寿命。 典型的反应器材料在操作条件下的微动磨损的特点。 虽然SMR设计基于现有的部署技术，但其他设计基于先进的概念，建议尺寸范围高达300 MWe，温度范围超过850°C。此外，这些组成部分将在完全不同的环境下运作。 此外，SMR的拟议材料主要考虑其保压特性和腐蚀性。 虽然获得了一些拟议SMR材料（如800 HT、Hastelloyx等）的强度和腐蚀性能的合理知识，但这些材料在高温和不同环境条件下的微动磨损特性尚不清楚。 本项目旨在了解微动磨损的退化机制，并提供更高的信心在高温下的性能指导。 该项目解决了NSERC-CNSC小型模块化反应堆研究资助计划中“高温对反应堆组件的影响”下的研究挑战和知识差距。