Critical heat flux in a two-phase thermosyphon: Relationship between maximum falling liquid rate and instability on interface of falling liquid

1997 ◽  
Vol 26 (6) ◽  
pp. 372-384 ◽  
Author(s):  
Masanori Monde ◽  
Yuhichi Mitsutake ◽  
Takeshi Hataya
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Two Phase

scholarly journals Determination of Pool Boiling Critical Heat Flux Enhancement in Nanofluids

2007 ◽  
Author(s):  
Bao H. Truong
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pool Boiling ◽  
Nucleate Boiling ◽  
Porous Coating ◽  
Transfer Coefficients ◽  
Two Phase ◽  
Sem Images ◽  
Heat Transfer Coefficients ◽  
Nucleate Boiling Heat Transfer

Nanofluids are engineered colloids composed of nano-size particles dispersed in common fluids such as water or refrigerants. Using an electrically controlled wire heater, pool boiling Critical Heat Flux (CHF) of Alumina and Silica water-based nanofluids of concentration less than or equal to 0.1 percent by volume were measured. Silica nanofluids showed a CHF enhancement up to 68% and there seems to be a monotonic relationship between the nanoparticle concentration and the magnitude of enhancement. Alumina nanofluids had a CHF enhancement up to 56% but the peak occurred at the intermediate concentration. The boiling curves in nanofluid were found to shift to the left of that of water and correspond to higher nucleate boiling heat transfer coefficients in the two-phase flow regime. Scanning Electron Microscopy (SEM) images show a porous coating layer of nanoparticles on wires subjected to nanofluid CHF tests. These coating layers change the morphology of the heater’s surface, and are responsible for the CHF enhancement. The thickness of the coating was estimated using SEM and was found ranging from 3.0 to 6.0 micrometers for Alumina, and 3.0 to 15.0 micrometers for Silica.

Experimental Investigation of the CHF Condition During Flow Boiling of Water in Microtubes

2007 ◽  
Author(s):  
Anand P. Roday ◽  
Michael K. Jensen
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Mass Flux ◽  
Flow Boiling ◽  
Heat Transfer Process ◽  
Working Fluid ◽  
Two Phase ◽  
Flow Boiling Heat Transfer ◽  
C 50 ◽  
Different Diameters

The critical heat flux (CHF) condition sets an upper limit on the flow-boiling heat transfer process. With the growing demand for the use of two-phase flow in micro and nano-sized devices, there is a strong need to understand the CHF phenomenon in channels of such small dimensions. This study experimentally investigates the critical heat flux condition during flow boiling in a single stainless steel microtube of two different diameters—0.427mm, and 0.286 mm. Degassed water is the working fluid. The effects of various parameters—diameter, mass flux (350–1500 kg/m2s), inlet subcooling (2°C–50°C), and length-to-diameter ratio (75–200) on the CHF condition are studied for the exit condition being nearly atmospheric pressure. The CHF increases with an increase in mass flux. The effect of the inlet subcooling on the CHF condition is more complex. With a decreasing inlet subcooling, the CHF decreases until saturated liquid is reached; thereafter, the CHF increases with quality.

Experimental Study of Two Phase Flow Behavior Past BWR Spacer Grids

2002 ◽  
Author(s):  
Ruwan K. Ratnayake ◽  
L. E. Hochreiter ◽  
K. N. Ivanov ◽  
J. M. Cimbala
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Test Section ◽  
Visual Information ◽  
Two Phase Flow ◽  
Flow Behavior ◽  
Phase Flow ◽  
Spacer Grid ◽  
Two Phase ◽  
Spacer Grids

Performance of best estimate codes used in the nuclear industry can be significantly improved by reducing the empiricism embedded in their constitutive models. Spacer grids have been found to have an important impact on the maximum allowable Critical Heat Flux within the fuel assembly of a nuclear reactor core. Therefore, incorporation of suitable spacer grids models can improve the critical heat flux prediction capability of best estimate codes. Realistic modeling of entrainment behavior of spacer grids requires understanding the different mechanisms that are involved. Since visual information pertaining to the entrainment behavior of spacer grids cannot possibly be obtained from operating nuclear reactors, experiments have to be designed and conducted for this specific purpose. Most of the spacer grid experiments available in literature have been designed in view of obtaining quantitative data for the purpose of developing or modifying empirical formulations for heat transfer, critical heat flux or pressure drop. Very few experiments have been designed to provide fundamental information which can be used to understand spacer grid effects and phenomena involved in two phase flow. Air-water experiments were conducted to obtain visual information on the two-phase flow behavior both upstream and downstream of Boiling Water Reactor (BWR) spacer grids. The test section was designed and constructed using prototypic dimensions such as the channel cross-section, rod diameter and other spacer grid configurations of a typical BWR fuel assembly. The test section models the flow behavior in two adjacent sub channels in the BWR core. A portion of a prototypic BWR spacer grid accounting for two adjacent channels was used with industrial mild steel rods for the purpose of representing the channel internals. Symmetry was preserved in this practice, so that the channel walls could effectively be considered as the channel boundaries. Thin films were established on the rod surfaces by injecting water through a set of perforations at the bottom ends of the rods, ensuring that the flow upstream of the bottom-most spacer grid is predominantly annular. The flow conditions were regulated such that they represent typical BWR operating conditions. Photographs taken during experiments show that the film entrainment increases significantly at the spacer grids, since the points of contact between the rods and the grids result in a peeling off of large portions of the liquid film from the rod surfaces. Decreasing the water flow resulted in eventual drying out, beginning at positions immediately upstream of the spacer grids.

Two-Phase Pipe Quenching Correlations for Liquid Nitrogen and Liquid Hydrogen

2019 ◽  
Vol 141 (4) ◽  
Author(s):  
S. R. Darr ◽  
J. W. Hartwig ◽  
J. Dong ◽  
H. Wang ◽  
A. K. Majumdar ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Liquid Nitrogen ◽  
Critical Heat Flux ◽  
Flow Boiling ◽  
Nucleate Boiling ◽  
Liquid Hydrogen ◽  
Two Phase ◽  
Cryogenic Flow ◽  
Two Phase Heat Transfer

Recently, two-phase cryogenic flow boiling data in liquid nitrogen (LN2) and liquid hydrogen (LH2) were compared to the most popular two-phase correlations, as well as correlations used in two of the most widely used commercially available thermal/fluid design codes in Hartwig et al. (2016, “Assessment of Existing Two Phase Heat Transfer Coefficient and Critical Heat Flux on Cryogenic Flow Boiling Quenching Experiments,” Int. J. Heat Mass Transfer, 93, pp. 441–463). Results uncovered that the correlations performed poorly, with predictions significantly higher than the data. Disparity is primarily due to the fact that most two-phase correlations are based on room temperature fluids, and for the heating configuration, not the quenching configuration. The penalty for such poor predictive tools is higher margin, safety factor, and cost. Before control algorithms for cryogenic transfer systems can be implemented, it is first required to develop a set of low-error, fundamental two-phase heat transfer correlations that match available cryogenic data. This paper presents the background for developing a new set of quenching/chilldown correlations for cryogenic pipe flow on thin, shorter lines, including the results of an exhaustive literature review of 61 sources. New correlations are presented which are based on the consolidated database of 79,915 quenching points for a 1.27 cm diameter line, covering a wide range of inlet subcooling, mass flux, pressure, equilibrium quality, flow direction, and even gravity level. Functional forms are presented for LN2 and LH2 chilldown correlations, including film, transition, and nucleate boiling, critical heat flux, and the Leidenfrost point.

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