KSME International Journal

, Volume 18, Issue 1, pp 74–81 | Cite as

The analysis of flow-induced vibration and design improvement in KSNP steam generators of UCN #5, 6



The KSNP Steam Generators (Youngkwang Unit 3 and 4, Ulchin Unit 3 and 4) have a problem of U-tube fretting wear due to Flow Induced Vibration (FIV). In particular, the wear is localized and concentrated in a small area of upper part of U-bend in the Central Cavity region. The region has some conditions susceptible to the FIV, which are high flow velocity, high void fraction, and long unsupported span. Even though the FIV could be occurred by many mechanisms, the main mechanism would be fluid-elastic instability, or turbulent excitation. To remedy the problem, Eggcrate Flow Distribution Plate (EFDP) was installed in the Central Cavity region of Ulchin Unit 5 and 6 steam generators, so that it reduces the flow velocity in the region to a certain level. However, the cause of the FIV and the effectiveness of the EFDP was not thoroughly studied and checked. In this study, therefore the Stability Ratio (SR), which is the ratio of the actual velocity to the critical velocity, was compared between the value before the installation of EFDP and that after. Also the possibility of fluid-elastic instability of KSNP steam generator and the effectiveness of EFDP were checked based on the ATHOS3 code calculation and the Pettigrew’ s experimental results. The calculated results were plotted in a fluid-elastic instability criteria-diagram (Pettigrew, 1998, Fig. 9). The plotted result showed that KSNP steam generator with EFDP had the margin of Fluid-Elastic Instability by almost 25%.

Key Words

Flow-Induced Vibration (FIV) Eggcrate Flow Distribution Plates (EFDP) Stability Ratio (SR) Fluid-Elastic Instability Critical Velocity Korea Standard Nuclear Plant (KSNP) 



Outside diameter of tube


Equivalent diameter of flow boundary

fn , f

Natural frequency


Instability constant


Hydraulic (Added) mass


Effective mass of tube


Pitch of tube array


Critical velocity


Effective velocity


Gap velocity


Cross flow velocity spanwise variation


Spanwise coordinate measured along tube axis


Pith to diameter ratio



Void fraction


Logarithmic decrement ( = 2πζ)


Spanwise variation in normalized modal displacement


Secondary fluid density (average)


Secondary fluid density spanwise variation


Two phase density


Damping ratio


Film dynamic damping


Structural damping


Two-phase damping


Viscous damping


Kinetic viscosity of liquid


Kinetic viscosity of gas



Doo San Heavy Industry Company


Eggcrate Flow Distribution Plate


Korea Standard Nuclear Plant


Stability Ratio


Westinghouse Electric Company Limited Liability Company


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  1. Au-Yang, M. K., 1987, “ Development of Stabilizers for Steam Generator Tube Repair,”Journal Nuclear Engineering and Design, Vol. 122, pp. 189–197.Google Scholar
  2. Axiasa, F., 1986, “ Flow Induced Vibration of Steam Generator Tubes,”Electric Power Research Institute Report EPRI NP-4559. Google Scholar
  3. Chen, S. S., 1985, “Flow Induced Vibration of Circular Cylindrical Structures,”Report No. ANL-85-51. Google Scholar
  4. Jo, J. C, 1992, “A Study on the Thermal-hydraulic and Flow-induced Tube Vibration Analysis of Nuclear Steam Generators,”KINS/ AR-198.Google Scholar
  5. Kim, S. N., 1998, “Fluid-elastic Vibration in a Rod Bundle,”Ministry of Science and Technology (MOST), Korea.Google Scholar
  6. Kim, S. N., 2000, “ Critical Velocity of Fluidelastic Vibration in a Nuclear Fuel Bundle,”KSME International Journal, Vol. 14, pp, 816–822.Google Scholar
  7. Kim, S. N. and Sin, C., 2001, “ The Experimental of Flow Induced Vibration in PWR RCCAs,”KSME International Journal, Vol. 15, pp. 291–299Google Scholar
  8. Lee, C. H., 1991, “Vibration & Structural Analysis of The Tubes & Tube supports,”ABBCENP. Google Scholar
  9. Lee, L. S., 1971, “Vibration of U-Bend Segments of Heat Exchanger Tubes,”Atomic Energy of Canada Limited, AECL 3735.Google Scholar
  10. Pettigrew, M. J., Taylor, C. E., Fisher, N. J., Yetisir, M. and Smith. B. A. W., 1998, “Flowinduced Vibration: Recent Findings and Open Questions,”Journal Nuclear Engineering and Design, Vol. 185, pp. 249–276.CrossRefGoogle Scholar
  11. “Palo Verde Steam Generator Tube Degradation and YGN3 and 4 Steam Generator Design,”ABB-CENP Report Google Scholar
  12. Pettigrew, M. J. and Taylor, C. E., 1994, “ Two-Phase Flow-Induced Vibration”,ASEM, Vol. 116, pp. 233–252.Google Scholar
  13. Pettigrew, M. J., Taylor, C. E. and Kim, B. S., 1989, “ Vibration of Tube Bundles in Two-Phase Cross Flow: Parti Hydrodynamic Mass and Damping,”Journal of Pressure Vessel Technology Trans. of the ASME, Vol. 3, pp. 466–477.Google Scholar
  14. Singhal, A. K. and Keeton, L. W., 1990, “ATHOS3 Mod-01 : A Computer Program for Thermal-Hydraulic Analysis of Steam Generators,”EPRI NP-4604-CCML, Vol. 1–3.Google Scholar
  15. Singhal, A. K. and Srikantiah, G., 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.Google Scholar
  16. Slack, D. G., 2001, “Vibration and Structural Analysis of the Tubes and Tube supports (UE-2I1SS-302),” WECLLC.Google Scholar

Copyright information

© The Korean Society of Mechanical Engineers (KSME) 2004

Authors and Affiliations

  1. 1.Department of Nuclear EngineeringKyunghee UniversityKyung- gi- doKorea

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