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Fluidelastic Instability Characteristics of Helical Steam Generator Tubes

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  );
  • 초록

    This study investigates the fluidelastic instability characteristics of helical steam generator type tubes used in operating nuclear power plants. To obtain a natural frequency, corresponding mode shape, and participation factor, modal analyses using various conditions are performed for helical type tubes. Investigated are the effects of the number of turns, the number of supports, and the status of the inner fluid on the modal and fluidelastic instability characteristics of the tubes, which are expressed in terms of the natural frequency, the corresponding mode shape, and the stability ratio.


  • 주제어

    fluidelastic instability .   steam generator helical tubes .   modal analyses .   mode shape .   participation factor .   stability ratio.  

  • 참고문헌 (12)

    1. Jo, J.C., Jhung, M.J., Kim, W.S., Choi, Y.H., Kim, H.J., and Kim, T.H., 2003, 'Fretting-wear characteristics of steam generator tubes by foreign object,' Journal of the Korean Nuclear Society, Vol.35, No.5, pp.442-453     
    2. Jo, J.C., Jhung, M.J., Kim, W.S., Kim, H.J., and Kim, T.H., 2003, 'Vibration characteristics of steam generator U-tubes with defect,' Transactions of the Korean Society of Noise and Vibration Engineering, Vol.13, No.5, pp.400-408     
    3. ANSYS, 2003, ANSYS Structural Analysis Guide, ANSYS, Inc., Houston 
    4. Connors, H.J., 1981, 'Flow-Induced Vibration and Wear of Steam Generator Tubes,' Nuclear Technology, Vol.55, pp.311-331 
    5. Au-Yang, M.K., 2001, Flow-Induced Vibration of Power and Process Plant Components, ASME Press, New York 
    6. ASME, 1998, Flow-Induced Vibration of Tubes and Tube Banks, ASME Boiler and Pressure Vessel Code, Section III, Appendix N-1300, The American Society of Mechanical Engineers, New York 
    7. Pettigrew, M.J., and Gorman, D.J., 1981, 'Vibration of heat exchanger tube bundles in liquid and two-phase cross-flow,' Flow-Induced Vibration Design Guidelines, The American Society of Mechanical Engineers, New York, pp.89-110 
    8. Paidoussis, M.P., 1983, 'A review of flow-induced vibrations in reactors and reactor components,' Nuclear Science Engineering, Vol.74, pp.31-60 
    9. Yetisir, M, and Pettigrew, M.J., 2001, 'A simple approach to estimate fretting-wear damage in heat exchanger tubes ; verification and validation,' PVP-Vol.420.1, pp.7-16 
    10. Eisinger, F.L., Rao, M.S.M., and Steininger, D.A., 1989, 'Numerical Simulation of Fluidelastic Instability of Multispan Tubes Partially Exposed to Cross Flow,' Proceedings of the 10th International Conference on Structural Mechanics in Reactor Technology, Vol.T, pp.45-57 
    11. Eisinger, F.L., and Juliano, T.M., 1975, Flow-Induced Vibration Analysis of Recuperator Tube Bank, 156-MA-75-48, Foster Wheeler Energy Corp., New Jersey 
    12. Cioncolini, A, Cammi, L., Castelli, G., Lombardi, C., Luzzi, L., Ricotti, M.E., 2003, 'Thermal Hydraulic Analysis of IRIS Reactor Coiled Tube Steam Generator,' Nuclear Mathematical and Computational Science : A Century in Review, American Nuclear Society 

 저자의 다른 논문

  • Jo, Jong-Chull (17)

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    2. 1991 "PWR 가압기에서 오동작 보조살수 과도시 용기벽의 열적 과도응답" Journal of the Korean Nuclear Society = 원자력학회지 23 (2): 183~199    
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    7. 2003 "결함을 가진 증기발생기 U-튜브의 진동특성" 한국소음진동공학회논문집 = Transactions of the Korean society for noise and vibration engineering 13 (5): 400~408    
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    10. 2006 "IMPACT ANALYSIS OF A WATER STORAGE TANK" Nuclear engineering and technology : an international journal of the Korean Nuclear Society 38 (7): 681~688    
  • Jhung, Myung Jo (29)

  • Kim Woong Sik (1)

  • 최영환 (44)

  • 김효정 (27)

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