NDLI: Rod Bundle Heat Transfer: Steady-State Steam Cooling Experiments

Content Provider	The American Society of Mechanical Engineers (ASME) Digital Collection
Author	Spring, J. P. Mclaughlin, D. M.
Copyright Year	2006
Abstract	Through the joint efforts of the Pennsylvania State University and the United States Nuclear Regulatory Commission, an experimental rod bundle heat transfer (RBHT) facility was designed and built. The rod bundle consists of a 7×7 square pitch array with spacer grids and geometry similar to that found in a modern pressurized water reactor. From this facility, a series of steady-state steam cooling experiments were performed. The bundle inlet Reynolds number was varied from 1 400 to 30 000 over a pressure range from 1.36 to 4 bars (20 to 60 psia). The bundle inlet steam temperature was controlled to be at saturation for the specified pressure and the fluid exit temperature exceeded 550 °C in the highest power tests. One important quantity of interest is the local convective heat transfer coefficient defined in terms of the local bulk mean temperature of the flow, local wall temperature, and heat flux. Steam temperatures were measured at the center of selected subchannels along the length of the bundle by traversing miniaturized thermocouples. Using an analogy between momentum and energy transport, a method was developed for relating the local subchannel centerline temperature measurement to the local bulk mean temperature. Wall temperatures were measured using internal thermocouples strategically placed along the length of each rod and the local wall heat flux was obtained from an inverse conduction program. The local heat transfer coefficient was calculated from the data at each rod thermocouple location. The local heat transfer coefficients calculated for locations where the flow was fully developed were compared against several published correlations. The Weisman and El-Genk correlations were found to agree best with the RBHT steam cooling data, especially over the range of turbulent Reynolds numbers. The effect of spacer grids on the heat transfer enhancement was also determined from instrumentation placed downstream of the spacer grid locations. The local heat transfer was found to be greatest at locations immediately downstream of the grid, and as the flow moved further downstream from the grid it became more developed, thus causing the heat transfer to diminish. The amount of heat transfer enhancement was found to depend not only on the spacer grid design, but also on the local Reynolds number. It was seen that decreasing Reynolds number leads to greater heat transfer enhancement.
Sponsorship	Nuclear Engineering Division
Starting Page	741
Ending Page	747
Page Count	7
File Format	PDF
ISBN	0791842452
DOI	10.1115/ICONE14-89734
e-ISBN	0791837831
Volume Number	Volume 4: Computational Fluid Dynamics, Neutronics Methods and Coupled Codes; Student Paper Competition
Conference Proceedings	14th International Conference on Nuclear Engineering
Language	English
Publisher Date	2006-07-17
Publisher Place	Miami, Florida, USA
Access Restriction	Subscribed
Subject Keyword	Temperature Turbulence Cooling Reynolds number Thermocouples Instrumentation Momentum Steady state Flow (dynamics) Pressure Heat flux Convection Fuel rods Wall temperature Design Geometry Heat conduction Temperature measurement Fluids Steam Heat transfer coefficients Pressurized water reactors Heat transfer
Content Type	Text
Resource Type	Article

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