STP780

    Heat Exchanger Tube Fretting Wear: Correlation of Tube Motion and Wear

    Published: Jan 1982


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    Abstract

    A typical steam generator in a nuclear power station consists of about 4000 tubes which form the boundary separating the light water in the secondary circuit from the primary circuit water. The integrity and the life expectancy of these tubes are therefore of prime concern to the designers.

    One of the several tube failure mechanisms, which include corrosion and fatigue, is fretting wear due to flow-induced tube vibrations. The tube-tube support clearance needed for design considerations allows periodic separation and impacting between the tube and the tube support hole and hence ready removal of loose wear particles by the fluid flow. It has been found that the combined rubbing and impacting motion, together with the periodic separation of contacting surfaces, all contribute to an accelerated wear process.

    A tube-fretting test apparatus has been developed to study the effect on tube wear of various parameters, such as tube-tube support interaction, materials, temperature, and clearance. Tests have been conducted in water at room temperature and at steam-generator operating temperature (265°C). Recently the authors have looked into the more fundamental aspect of tube fretting wear mechanisms. Some of the worn surfaces were further studied by a surface analyzer interfaced with a computer and by a scanning electron microscope. The results from these analyses were correlated to other tube fretting parameters, namely, tube motion, clearance, and impact forces at the support. We have found that the effects of these parameters on wear are interrelated. The understanding of these interrelated effects is of great importance in predicting long-term tube wear. This paper will describe our fretting test facilities and present the results from recent studies on wear correlations.

    Keywords:

    fretting, wear, heat exchanger tube, flow-induced vibration, impact force, surface analysis


    Author Information:

    Ko, PL
    Research engineer, technologist, and technician, Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories, Chalk River, Ontario

    Tromp, JH
    Research engineer, technologist, and technician, Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories, Chalk River, Ontario

    Weckwerth, MK
    Research engineer, technologist, and technician, Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories, Chalk River, Ontario


    Paper ID: STP29398S

    Committee/Subcommittee: G02.30

    DOI: 10.1520/STP29398S


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