Journal of Mechanical Science and Technology

The Korean Society of Mechanical Engineers (대한기계학회)
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An Experimental Study of Pressure Drop Correlations for Wire-Wrapped Fuel Assemblies

  • Chun, Moon-Hyun (Department of Nuclear Engineering, Korea Advanced Institute of Science and Technology) ;
  • Seo, Kyong-Won (Department of Nuclear Engineering, Korea Advanced Institute of Science and Technology) ;
  • Park, Seok-Ki (Korea Atomic Energy Research Institute) ;
  • Nam, Ho-Yun (Korea Atomic Energy Research Institute)
  • Published : 2001.03.01
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Abstract

The main objective of the present study is to perform an experimental evaluation of five existing correlations for the subchannel pressure drop analysis of a wire-wrapped fuel assembly. For this purpose, a series of water experiments have been performed using a helical wire-wrapped 19-pin fuel assembly for various test parameters. Four different test sections with different pitch to rod diameter ratios (P/D) and wire lead length to rod diameter ratios (H/D) have been fabricated. A series of pressure drop measurements were made to obtain friction factors for these four test sections. The new data along with existing data are used to evaluate existing correlations. Both the original and the simplified Cheng and Todreas correlations give the best agreement with experimental data for all flow regions.

Keywords

References

  1. Cheng, S. K. and Todreas, N. E., 1986, 'Hydrodynamic Models and Correlations for Bare and Wire-Wrapped Hexagonal Rod Budles-Bundle Friction Factors, Subchannel Friction Factors and Mixing Parameters,' Nucl. Eng. Des., 92, 227 https://doi.org/10.1016/0029-5493(86)90249-9
  2. Engel, F. C., Markley, R. A. and Bishop, A. A., 1979,'Laminar, Transition, and Tubulent Parallel Flow Pressure Drop Across Wire-Wrap-Spaced Rod Bundles,' Nucl. Sci. Eng., 69, pp. 290
  3. Kim, W. S. and Kim, Y. G., 1998, 'MATRA-LMR Code for Thermal-Hydraulics Subchannel Analysis of LMR.' NTHAS98: First Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety, Pusan, Korea, October, 21-24, 227
  4. Novendstern, E. H., 1972,'Turbulent Flow Pressure Drop Model for Fuel Rod Assemblies Utilizing A Helical Wire-Wrap Spacer System,' Nucl. Eng. Des., 22, 19 https://doi.org/10.1016/0029-5493(72)90059-3
  5. Rehme, K., 1972, 'Pressure Drop Correlations for Fuel Element Spacers,' Nucl. Tech., 17, 15
  6. Wheeler, C. L., 1976, 'COBRA-IV-1 : An Interim Version of COBRA for Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores,' BNWL-1662
  7. Yoo, Y. J. and Hwang, D. H., 1997, 'Development of Subchannel Analysis Code MATRA ${\alpha}$ -version,' Proceedings of Korea Nuclear Society Autumn Meeting, Taegu, Korea, October 24-25