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KSME International Journal

, Volume 15, Issue 3, pp 403–409 | Cite as

An experimental study of pressure drop correlations for wire-wrapped fuel assemblies

  • Moon-Hyun Chun
  • Kyong-Won Seo
  • Seok-Ki Choi
  • Ho-Yun Nam
Article

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. For 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.

Key Words

Pressure Drop Correlations Wire-Wrapped Rod Bundle Subchannel Analysis 

Nomenclature

A

Axial average flow area (mm2)

D

Rod diameter (mm)

Dw

Wire spacer diameter (mm)

De

Hydraulic equivalent diameter (mm)

f

Friction factor

H

Wire lead length (mm)

L

Axial length (mm)

P

Pressure (Pa)

P

Rod pitch (mm)

Pw

Wetted perimeter (mm)

Re

Reynolds number

V

Flow velocity (m/s)

X

Flow split parameter

Greek

ϱ

Density (m3/kg)

μ

Dynamic viscosity (Ns/m2)

ϕ

Intermittency factor

Subscript

i

Subchannel type index

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References

  1. Cheng, S. K. and Todreas, N. E., 1986, “Hydrodynamic Models and Correlations for Bare and Wire-Wrapped Hexagonal Rod Bundles-Bundle Friction Factors, Subchannel Friction Factors and Mixing Parameters,”Nucl. Eng. Des., 92, 227.CrossRefGoogle Scholar
  2. Engel, F. C., Markley, R. A. and Bishop, A. A., 1979, “Laminar, Transition, and Turbulent Parallel Flow Pressure Drop Across Wire-Wrap-Spaced Rod Bundles,”Nucl. Sci. Eng., 69, 290.Google Scholar
  3. Kim, W. S. and Kim, Y. G., 1998, “MATRA-LMR Code for Thermal-Hydraulic Subchannel Analysis of LMR,”NTHAS98: First Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety, Pusan, Korea, October 21–24, 227.Google Scholar
  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.CrossRefGoogle Scholar
  5. Rehme, K., 1972, “Pressure Drop Correlations for Fuel Element Spacers,”Nucl. Tech., 17, 15.Google Scholar
  6. Wheeler, C. L., 1976, “COBRA-IV-I: An Interim Version of COBRA for Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores,”BNWL-1662.Google Scholar
  7. Yoo, Y. J. and Hwang, D. H., 1997, “Development of Subchannel Analysis Code MATRA α-version,”Proceedings of Korea Nuclear Society Autumn Meeting, Taegu, Korea, October 24–25.Google Scholar

Copyright information

© The Korean Society of Mechanical Engineers (KSME) 2001

Authors and Affiliations

  • Moon-Hyun Chun
    • 1
  • Kyong-Won Seo
    • 1
  • Seok-Ki Choi
    • 2
  • Ho-Yun Nam
    • 2
  1. 1.Department of Nuclear EngineeringKorea Advanced Institute of Science and TechnologyTaejonKorea
  2. 2.Korea Atomic Energy Research InstituteKorea

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