Locally averaged heat transfer measurements in a rod bundle downstream of support grids with and without flow-enhancing features are investigated for Reynolds numbers of 28,000 and 42,000. Support grids with disk blockage flow-enhancing features and support grids with split-vane pair flow enhancing features are examined. Grid pressure loss coefficients and feature loss coefficients are determined based on pressure drop measurements for each support grid design. Results indicate the greatest heat transfer enhancement downstream of the support grid designs with disk blockages. In addition, the local heat transfer measurements downstream of the split-vane pair grid designs indicate a region of decreased heat transfer below that of the hydrodynamically fully developed value. This decreased region of heat transfer is more pronounced for the lower Reynolds number case. A correlation for the local Nusselt numbers downstream of the standard support grid designs is developed based on the blockage of the support grid. In addition, a correlation for the local Nusselt numbers downstream of support grids with flow-enhancing features is developed based on the blockage ratio of the grid straps and the normalized feature loss coefficients of the support grid designs. The correlations demonstrate the tradeoff between initial heat transfer enhancement downstream of the support grid and the pressure drop created by the support grid.

1.
Rehme
,
K.
,
1973
, “
Pressure Drop Correlations for Fuel Element Spacers
,”
Nucl. Technol.
,
17
, pp.
15
23
.
2.
Herer, C., 1991, “3D Flow Measurements in Nuclear Fuel Rod Bundles Using Laser Doppler Velocimetry,” ASME Fluid Measurement and Instrumentation Forum, 108, ASME International, New York, pp. 95–99.
3.
Yang
,
S. K.
, and
Chung
,
M. K.
,
1998
, “
Turbulent Flow Through Spacer Grids in Rod Bundles
,”
ASME J. Fluids Eng.
,
120
, pp.
786
791
.
4.
Karoutas, Z., Gu, C. Y., and Scholin, B., 1995, “3-D Flow Analyses for Design of Nuclear Fuel Spacer,” Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7, 1, pp. 3153–3174.
5.
In
,
K. W.
,
2001
, “
Numerical Study of Coolant Mixing Caused by the Flow Deflectors in a Nuclear Fuel Bundle
,”
Nucl. Technol.
,
134
, pp.
187
195
.
6.
McClusky
,
H. L.
,
Holloway
,
M. V.
,
Beasley
,
D. E.
, and
Conner
,
M. E.
,
2002
, “
Development of Swirling Flow in a Rod Bundle Subchannel
,”
ASME J. Fluids Eng.
,
124
, pp.
747
755
.
7.
McClusky, H. L., Holloway, M. V., Conover, T. A., Beasley, D. E., Conner, M. E., and Smith, L. D., III, 2003, “Mapping of the Lateral Flow Field in Typical Subchannels of a Support Grid with Vanes,” ASME J. Fluids Eng., 125(6), pp. 987–996.
8.
Saffman, P. G., 1992, Vortex Dynamics, Cambridge University Press.
9.
de Crecy
,
F.
,
1994
, “
The Effect of Grid Assembly Mixing Vanes on Critical Heat Flux Values and Azimuthal Location in Fuel Assemblies
,”
Nucl. Eng. Des.
,
149
, pp.
233
241
.
10.
Blum, H. A., and Oliver, L. R., 1966, “Heat Transfer in a Decaying Vortex System,” Proceedings of the 1966 ASME HTD Winter Annual Meeting, 62, pp. 1–8.
11.
Hay
,
N.
, and
West
,
P. D.
,
1975
, “
Heat Transfer in Free Swirling Flow in a Pipe
,”
ASME J. Heat Transfer
,
97, Series C
, No.
3
, pp.
411
416
.
12.
Chang
,
F.
, and
Dhir
,
V. K.
,
1995
, “
Mechanisms of Heat Transfer Enhancement and Slow Decay of Swirl in Tubes Using Tangential Injection
,”
Int. J. Heat Fluid Flow
,
16
, pp.
78
87
.
13.
Figliola, R. S., and Beasley, D. E., 1998, Theory and Design for Mechanical Measurements, 3rd Edition, John Wiley & Sons.
14.
Armfield, M. V., 2001, “Effects of Support Grid Design on Local, Single-Phase Turbulent Heat Transfer in Rod Bundles,” Master’s Thesis, Clemson University, Clemson, SC, USA.
15.
Armfield, M. V., Langford, H. M., Beasley, D. E., and Conner, M. E., 2000, “Average Heat Transfer Coefficient Measurements in a Fuel Bundle: Method Development,” Proceedings of the ASME Heat Transfer Division, HTD-366-2, ASME International, New York, pp. 163–170.
16.
Rehme
,
K.
,
1976
, “
Pressure Drop of Spacer Grids in Smooth and Roughened Rod Bundles
,”
Nucl. Technol.
,
33
, pp.
314
317
.
17.
Marek, J., and Rehme, K., 1979, “Heat Transfer in Smooth and Roughened Rod Bundles Near Spacer Grids,” Presented at the ASME Winter Annual Meeting Dec 2–7, 1979, pp. 163–170.
18.
Yao
,
S. C.
,
Hochreiter
,
L. E.
, and
Leech
,
W. J.
,
1982
, “
Heat-Transfer Augmentation in Rod Bundles Near Grid Spacers
,”
ASME J. Heat Transfer
,
104
, pp.
76
81
.
19.
Kidd, G. J., Hoffman, H. W., and Stelzman, W. J., 1968, “The Temperature Structure and Heat Transfer Characteristics of an Electrically Heated Model of a Seven-Rod Cluster Fuel Element,” ASME Paper 68-WA/HT-33.
20.
Kreith
,
F.
, and
Sonju
,
O. K.
,
1965
, “
The Decay of a Turbulent Swirl in a Pipe
,”
J. Fluid Mech.
,
22
, pp.
257
271
.
21.
Armfield, M. V., Langford, H. M., Beasley, D. E., and Conner, M. E., 2001, “Single-Phase Turbulent Rod Bundle Heat Transfer,” Proceedings of 2001 ASME International Mechanical Engineering and Exposition, IMECE2001/HTD-24116, pp. 1–8.
22.
Guellouz
,
M. S.
, and
Tavoularis
,
S.
,
1992
, “
Heat Transfer in Rod Bundle Subchannels with Varying Rod-Wall Proximity
,”
Nucl. Eng. Des.
,
132
, pp.
351
366
.
23.
Dingee, D. A., and Chastain, J. W., 1956, “Heat Transfer from Parallel Rods in Axial Flow,” Reactor Heat Transfer Conference of 1956, TID-7529 (Pt. 1), Book 2, pp. 462–501.
24.
Comte-Bellot
,
G.
, and
Corrsin
,
S.
,
1966
, “
The Use of a Contraction to Improve the Isotropy of Grid-Generated Turbulence
,”
J. Fluid Mech.
,
25
, pp.
657
682
.
25.
Mills
,
A. F.
,
1962
, “
Experimental Investigation of Turbulent Heat Transfer in the Entrance Region of a Circular Conduit
,”
J. Mech. Eng. Sci.
4
, No. 1 pp.
63
77
.
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