Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Low-frequency modes in the fluid-structure interaction of a U-tube model for the steam generator in a PWR

  • Zhang, Hao (Electric Vehicle Research and Development Center, Tianjin Chinese-Academy-of-Sciences Institute of Advanced Technology, Chinese Academy of Sciences) ;
  • Chang, Se-Myong (School of Mechanical Convergence System Engineering, Kunsan National University) ;
  • Kang, Soong-Hyun (Department of Mechanical Engineering, Kunsan National University)
  • Received : 2018.10.11
  • Accepted : 2019.02.05
  • Published : 2019.05.25
Download PDF
⟨ Previous Next ⟩

Abstract

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

Keywords

References

  1. US-NRC Training Center, 0603, Reactor Concepts Manual: Pressurized Water Reactor Systems, vol. 4, 2014, pp. 1-28. also found in a web: http://www.nrc.gov/reading-rm/basic-ref/teachers/04.pdf.
  2. S.J. Green, G. Hertsroni, PWR steam generators, Int. J. Multiph. Flow 21 (Suppl) (1995) 1-97.
  3. Yonhap New Agency, Korea, Nuclear Reactor Shuts Down Due to Cracks in Steam Generator, 2014. Oct. 17.
  4. C.H. Park, Heat transfer tube in steam generator of nuclear plant, Nuclear Industry 5 (2013) 51-53.
  5. S.Y. Lee, S.M. Chang, G. Jang, Research trends of wear damage in steam generator tubes, in: KSME2002R1002, Trans. Of the Korean Society for Mechanical Engineers Spring Meeting, Pp. 5-8, Jeonju, Korea, May 18, 2012.
  6. K.W. Ryu, C.Y. Park, H. Lee, Effects of support structure changes on flow-induced vibration characteristics of steam generator tubes, Nuclear Engineering and Technology 42 (2010) 97-108. https://doi.org/10.5516/NET.2010.42.1.097
  7. G.H. Gim, S.M. Chang, S.Y. Lee, G. Jang, Fluid-structure interaction in a U-tube with surface roughness and pressure drop, Nuclear Engineering and Technology 46 (2014) 633-640. https://doi.org/10.5516/NET.02.2014.001
  8. F.M. White, Fluid Mechanics, sixth ed., McGraw-Hill, New York, 2010.
  9. F.R. Mentor, Two-equation eddy-viscosity turbulence models for engineering application, AIAA J. 32 (1994) 1598-1605. https://doi.org/10.2514/3.12149
  10. F.M. Tech, Priciples of Vortex Flowmeters, FM Tech Inc., Ulsan, Korea, 2008.
  11. H. Zhang, S.M. Chang, Modal analysis of a U-tube model for the steam generator with experiment and computation, in: Trans. Of the Korean Nuclear Society Autumn Meeting, Gyeongju, Korea, Oct. 27-28, 2016.

Cited by

  1. Temperature Coupling Analysis Between Nuclear Steam Generators and Heat Exchanger Inside Pressurized Water Reactors vol.195, pp.1, 2019, https://doi.org/10.1080/00295639.2020.1787013
  2. Modal analysis of central impression cylinder based on fluid-solid coupling method vol.40, pp.2, 2019, https://doi.org/10.1177/1461348420918413