Flow-induced vibration and related technologies in nuclear components

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Mechanical damage due to excessive vibration may limit the life of nuclear station components and cause costly station shutdowns. It is important to understand flow-induced vibration and damage mechanisms to prevent problems at the design stage and assist station owners in predicting the life of nuclear components. This paper outlines state-of-the-art technologies in the related areas of thermal-hydraulic analysis, flow-induced vibration and mechanical damage prediction. In this paper, nuclear component means steam generators, heat exchangers, condensers, piping systems and reactor internals. Vibration problems often occur locally in areas where excessive flow velocities exist, such as inlet regions and around sealing strips in heat exchangers. Detailed three-dimensional thermal-hydraulic analyses are required to obtain flow velocity distribution in components. Recent development work to predict flow velocity distribution around sealing strips and in the U-bend tube region of steam generators is described. Vibration excitation mechanisms must be formulated to predict the vibration response of nuclear components. Turbulence-induced excitation is important in axial flow. In cross flow, fluidelastic instability is usually the cause of excessive vibration. While reasonable amount of information is available in liquid and gas flow, very little work has been done in two-phase flow. The results of recent experimental work on vibration of tube bundles in two-phase cross-flow is presented. Damping, fluidelastic instability and turbulence-induced excitation are discussed. Design guidelines are suggested. Damage mechanisms must be understood to predict component life. Studies on fretting-wear damage of steam generator materials under simulated operating environments are discussed. Techniques to model the dynamic interaction between vibrating components and their supports are outlined.