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.
Similar content being viewed by others
Abbreviations
- A :
-
Axial average flow area (mm2)
- D :
-
Rod diameter (mm)
- D w :
-
Wire spacer diameter (mm)
- D e :
-
Hydraulic equivalent diameter (mm)
- f :
-
Friction factor
- H :
-
Wire lead length (mm)
- L :
-
Axial length (mm)
- P :
-
Pressure (Pa)
- P :
-
Rod pitch (mm)
- P w :
-
Wetted perimeter (mm)
- Re:
-
Reynolds number
- V :
-
Flow velocity (m/s)
- X :
-
Flow split parameter
- ϱ:
-
Density (m3/kg)
- μ:
-
Dynamic viscosity (Ns/m2)
- ϕ:
-
Intermittency factor
- i :
-
Subchannel type index
References
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.
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.
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.
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.
Rehme, K., 1972, “Pressure Drop Correlations for Fuel Element Spacers,”Nucl. Tech., 17, 15.
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.
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.
Author information
Authors and Affiliations
Corresponding author
Additional information
First Author
Rights and permissions
About this article
Cite this article
Chun, MH., Seo, KW., Choi, SK. et al. An experimental study of pressure drop correlations for wire-wrapped fuel assemblies. KSME International Journal 15, 403–409 (2001). https://doi.org/10.1007/BF03185224
Received:
Revised:
Issue Date:
DOI: https://doi.org/10.1007/BF03185224