Prediction of critical heat flux (CHF) for vertical round tubes with uniform heat flux in medium pressure regime

2004 ◽  
Vol 21 (1) ◽  
pp. 75-80 ◽  
Author(s):  
W. Jaewoo Shim ◽  
Joo-Yong Park
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Uniform Heat Flux ◽  
Medium Pressure ◽  
Uniform Heat ◽  
Pressure Regime

Critical Heat Flux in Tubes With Cosine Axial Heat Flux

2005 ◽  
Author(s):  
W. Jaewoo Shim ◽  
Joo-Yong Park ◽  
Ji-Su Lee ◽  
Dong Kook Kim
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Mass Flux ◽  
Flux Distribution ◽  
Average Error ◽  
Uniform Heat Flux ◽  
Heat Flux Distribution ◽  
Uniform Heat ◽  
System Pressure ◽  
Data Points

In this study a method to predict CHF (Critical Heat Flux) in vertical round tubes with cosine heat flux distribution was examined. For this purpose a uniform correlation, based on local condition hypothesis, was developed from 9,366 CHF data points of uniform heat flux heaters. The CHF data points used were collected from 13 different sources had the following parameter ranges: 1.01 ≤ P (pressure) ≤ 206.79 bar, 9.92 ≤ G (mass flux) ≤ 18,619.39 kg/m2s, 0.00102 ≤ D (diameter) ≤ 0.04468 m, 0.0254 ≤ L (length) ≤ 4.966 m, 0.11 ≤ qc (CHF) ≤ 21.42 MW/m2, and −0.87 ≤ X (exit qualities) ≤ 1.58. The result of this work showed that the uniform CHF correlation could be used to predict CHF accurately in a non-uniform heat flux heater for wide flow conditions. Furthermore, the location, where CHF occurs in non-uniform heat flux distribution, can also be determined accurately with the local variables: the system pressure (P), tube diameter (D), mass flux of water (G), and true mass flux of vapor (GXt). The new correlation predicted CHF with cosine heat flux, 297 data points from 5 different published sources, within the root mean square error of 12.42% and average error of 1.06% using the heat balance method.

CFD prediction of critical heat flux in vertical heated tubes with uniform and non-uniform heat flux

2018 ◽  
Vol 326 ◽  
pp. 403-412 ◽  
Author(s):  
Quan Li ◽  
M. Avramova ◽  
Yongjun Jiao ◽  
Ping Chen ◽  
Junchong Yu ◽  
...  
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Uniform Heat Flux ◽  
Uniform Heat

A New Facility for Measurements of Three-Dimensional, Local Subcooled Flow Boiling Heat Flux and Related Critical Heat Flux

2000 ◽  
Author(s):  
Ronald D. Boyd ◽  
Penrose Cofie ◽  
Qing-Yuan Li ◽  
Ali Ekhlassi
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Test Section ◽  
Flow Boiling ◽  
Three Dimensional ◽  
Uniform Heat Flux ◽  
Subcooled Flow Boiling ◽  
Uniform Heat ◽  
Subcooled Flow ◽  
New Facility

Abstract In the development of plasma-facing components (PFC) for fusion reactors and high heat flux heat sinks (or components) for electronic applications, the components are usually subjected to a peripherally non-uniform heat flux. Even if the applied heat flux is uniform in the axial direction [which is unlikely], both intuition and recent investigations have clearly shown that both the local heat flux and the eventual critical heat flux (CHF) in this three-dimensional case will differ significantly from similar quantities found in the voluminous body of data for uniformly heated tubes and flow channels. Although this latter case has been used in the past as an estimate for the former case, more study has become necessary to examine the three-dimensional temperature and heat flux distributions and related CHF. Work thus far has shown that the non-uniform peripheral heat flux condition enhances CHF in some cases. In order to avoid the excess costs associated with using electron- or ion-beams to produce the non-uniform heat flux, a new facility was developed which will allow three-dimensional conjugate heat transfer measurements and two-dimensional local subcooled flow boiling heat flux and related critical heat flux measurements. The configurations under study consist of: (1) a non-uniformly heated cylindrical-like test section with a circular coolant channel bored through the center, and (2) a monoblock which is a square cross-section parallelepiped with a circular drilled flow channel through the center line along its length. The theoretical or idealization of the cylindrical-like test section would be a circular cylinder with half (−90 degrees to +90 degrees) of its outside boundary subjected to a uniform heat flux and the remaining half insulated. For the monoblock, a uniform heat flux is applied to one of the outside surfaces and the remaining surfaces are insulated. The outside diameter of the cylindrical-like test section is 30.0 mm and its length is 200.0 mm. The monoblock square has lengths 30.0 mm. The inside diameter of the flow channel for both types of test sections is 10.0 mm. Water is the coolant. The inlet water temperature can be set at any level in the range from 26.0 °C to 130.0 °C and the exit pressure can be set at any level in the range from 0.4 MPa to 4.0 MPa. Thermocouples are placed at forty-eight locations inside the solid cylindrical-like or monoblock test section. For each of four axial stations, three thermocouples are embedded at four circumferential locations (0, 45, 135, and 180 degrees, where 0 degrees corresponds to that portion of the axis of symmetry close to the heated surface) in the wall of the test section. Finally, the mass velocity can be set at any level in the range from 0.6 to 10.0 Mg/m2s.

A Nondimensional Analysis of Boiling Dry a Vertical Channel with a Uniform Heat Flux

1981 ◽  
Vol 103 (4) ◽  
pp. 667-672 ◽  
Author(s):  
K. H. Sun ◽  
R. B. Duffey ◽  
C. Lin
Keyword(s):  
Heat Flux ◽  
Nuclear Power ◽  
Closed Form Solution ◽  
Vertical Channel ◽  
Form Solution ◽  
Uniform Heat Flux ◽  
Two Phase ◽  
Drift Flux ◽  
Rod Bundle ◽  
Uniform Heat

A thermal-hydraulic model has been developed for describing the phenomenon of hydrodynamically-controlled dryout, or the boil-off phenomenon, in a vertical channel with a spatially-averaged or uniform heat flux. The use of the drift flux correlation for the void fraction profile, along with mass and energy balances for the system, leads to a dimensionless closed-form solution for the predictions of two-phase mixture levels and collapsed liquid levels. The physical significance of the governing dimensionless parameters are discussed. Comparisons with data from single-tube experiments, a 3 × 3 rod bundle experiment, and the Three Mile Island nuclear power plant show good agreement.

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