Browse > Article

Fretting-Wear Characteristics of Steam Generator Tubes by Foreign Object  

Jo Jong Chull (Korea Institute of Nuclear Safety)
Jhung Myung Jo (Korea Institute of Nuclear Safety)
Kim Woong Sik (Korea Institute of Nuclear Safety)
Choi Young Hwan (Korea Institute of Nuclear Safety)
Kim Hho Jung (Korea Institute of Nuclear Safety)
Kim Tae Hyung (Korea Power Engineering Company, Inc.)
Publication Information
Nuclear Engineering and Technology / v.35, no.5, 2003 , pp. 442-453 More about this Journal
Abstract
This study investigates the safety assessment of the potential for fretting-wear damages on steam generator (SG) U-tubes caused by foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions are obtained from three-dimensional SG flow calculation using the ATHOS3 code. Modal analyses are performed for the finite element modelings of U-tubes to get the natural frequency, corresponding mode shape and participation factor. The wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted. Also, discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube.
Keywords
fretting-wear; steam generator U-tubes; modal analyses; mode shape; participation factor; foreign object; vibration;
Citations & Related Records
Times Cited By KSCI : 2  (Citation Analysis)
연도 인용수 순위
1 Archard, J. F., and Hirst, T., 1956, 'The Wear of Metals under Unlubricated Conditions,' Proceedings of the Royal Society of London, Vol.A(236), p.397
2 Au-Yang, M. K., 2001, Flow-induced Vibration of Power and Process Plant Components, ASME Press, New York
3 Lindeburg, M. R., 1994, Mechanical Engineering Reference Manual, 9th ed., Professional Publications, Inc.
4 Singhal, A. K., and Srikantiah, G. S., 1991, 'A Review of Thermal Hydraulic Analysis Methodology for PWR Steam Generators and ATHOS3 Code Applications,' Progress in Nuclear Energy, Vol.25, No.1, pp.7-70   DOI   ScienceOn
5 Jo, J. C. et al., 1992, A Study on the Thermal-hydraulic and Flow-induced Tube Vibration Analysis of Nuclear Steam Generators, KINS/AR-198, Korea Institute of Nuclear Safety, Daejeon, Korea
6 Pettigrew, M. J., and Taylor, C. E., 1994, 'Two-Phase Flow-Induced Vibration: An Overview,' ASME Journal of Pressure Vessel Technology, Vol.116, pp.233-253   ScienceOn
7 Keeton, L. W., and Singhal, A. K., 1986, ATHOS3 : A Computer Program for Thermal Hydraulic Analysis of Steam Generators, NP-4604-CCM, Electric Power Research Institute, Palo Alto, CA.
8 ANSYS, 2001, ANSYS Structural Analysis Guide, ANSYS, Inc., Houston
9 Connors, H. J., 1981, 'Flow-Induced Vibration and Wear of Steam Generator Tubes,' Nuclear Technology, Vol.55, pp.311-331