• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.2026-2033
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• 2019

Fluidelastic instability of a tube array is a key factor of the security of a nuclear power plant. An unsteady model of the fluidelastic instability of a tube array subjected to two-phase flow was developed to analyze the fluidelastic instability of tube bundles in two-phase flow. Based on this model, a computational program was written to calculate the eigenvalue and the critical velocity of the fluidelastic instability. The unsteady model and the program were verified by comparing with the experimental results reported previously. The influences of void fraction and the tube's material properties on the critical velocity were investigated. Numerical results showed that, with increasing the void fraction of the two-phase flow, the tube array becomes more stable. The results indicate that the critical velocities of the tube array made of stainless are much higher than those of the other two tube arrays within void fraction ranging from 20% to 80%.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.349-358
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• 2007

A single-beam gamma densitometer is utilized to measure the average void fraction in a small diameter stainless steel pipe under critical flow conditions. A typical design of a single-beam gamma densitometer is composed of a sealed gammaray source, a collimator, a scintillation detector, and a data acquisition system that includes an amplifier and a single channel analyzer. It is operated in the count mode and can be calibrated with a test pipe and various types of phantoms made of polyethylene. A good average void fraction is obtained for a small diameter pipe with various flow regimes of the core, annular, stratified, and bubbly flows. Several factors influencing the performance of the gamma densitometer are examined, including the distance between the source and the detector, the measuring time, and the ambient temperature. The void fraction is measured during an adiabatic downward two-phase critical flow in a vertical pipe. The test pipe has an inner diameter of 10.9 mm and a thickness of 3.2 mm. The average void fraction was reasonably measured for a two-phase critical flow in the presence of nitrogen gas.

• Nuclear Engineering and Technology
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• v.29 no.1
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• pp.35-44
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• 1997

Critical too-Phase flow rates of subcooled water through Short Pipes (L 140039n) with small diameters (D$\leq$7.15 min) have been experimentally investigated for wide ranges of subcooling (0~199$^{\circ}C$) and pressure (0.5~2.0 MPa). To examine the effects of various parameters (i.e., the location of flashing inception, the degree of subcooling, the stagnation temperature and pressure, and the pipe size) on the critical two-phase flow rates of subcooled water through short pipes with small diameters, a total of 135 runs were made for various combinations of test parameters using four different L/D test sections. Experimental results that show effect of various parameters on subcooled critical two phase flow rates are presented in the form of graphs such as the dimensionless mass flux ( $G^{*}$) versus the dimensionless subcooling ( $T_{sub}$$^{*}$) curve. An empirical correlation expressed in terms of a dimensionless subcooling is also obtained for subcooled two-phase flow rates through present test sections. Comparisons between the mass fluxes calculated by present correlation and a total of 755 selected experimental data points of 9 different investigators show that the agreement is fairly good except for very low subcooling data obtained from small L/D (less than 10) orifices.s.s.s.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.273-278
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• 1997

Critical two-phase flow rates of subcooled water through very short tube (L=20 mm) with small diameters (D=1.0 mm) has been measured for wide ranges of subcooling(0~186$^{\circ}C$) and pressure (0.5~2.0 MPa). Experimental results show that subcooled critical two-phase flow rates can be expressed in terms of two scaling parameters for geometries and initial conditions. They are discharge coefficient of cold water, ( $C_{d}$ )$_{ref}$, and dimensionless subcooling, $\Delta$ $T^{*}$$_{sub}$, respectively. A new empirical correlation expressed in terms of ( $C_{d}$ )$_{ref}$ and $\Delta$ $T^{*}$$_{sub}$ is obtained for subcooled two-phase flow rates through very short length tube. Comparisons between the mass fluxes calculated by Present correlation and a number of experimental data show that the agreement is very good.ood.ood.ood.

• Journal of the Korean Society of Propulsion Engineers
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• v.7 no.1
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• pp.57-65
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• 2003

As a compressed water is rapidly expanded through a nozzle, two-phase flow of vapor and liquid is formed in the nozzle due to the flash evaporation. In the present study, critical flow of two-phase fluids is analysized using an Isentropic-Homogeneous-Equilibrium model and a Leung model. Calculation results show that the choke of the two-phase flow can be two different types of continuous and discontinuous chokings. For the stagnation pressure below 10 Mpa it is found that the continuous choking, which is similar to the choking phenomenon of single-phase gas flow, is possible only when the degree of subcooling is less than 10K.

• International Journal of Air-Conditioning and Refrigeration
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• v.16 no.2
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• pp.51-63
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• 2008

A study of literatures on flow boiling heat transfer and two-phase flows inside microchannels is summarized. The potential applications, fabrication method and efforts to determine certain dimensional threshold for microchannels classifications are discussed. For the last two decades, numerous two-phase flow and heat transfer models for microchannels have been developed; many of them were derived from empirical models originally applied for conventional channels. Those models are discussed here along with a brief review on recent development of theoretical and phenomenological-based models for microchannels. This study is devoted to provide a review of important issues on flow boiling heat transfer and two-phase flows inside microchannels, including two-phase flow patterns, boiling heat transfer mechanism and correlations developments, pressure drop and prediction methods, and critical heat flux.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.135-138
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• 2015

A new correlation predicting the idealized critical mass-flow rates of water for subcooled and saturated liquid water including two-phase water flow was developed for a wide range of upstream stagnation pressures (e.g., 0.5-20.0 MPa). A choking correction factor dependent on the upstream stagnation pressure and subcooled temperature was introduced into a new correlation, and its values were suggested to satisfy the idealized nozzle data within 10% error ranges. The suggested correlation will be instructive and helpful for related studies and/or engineering works.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.12
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• pp.939-946
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• 2009

Wettability is a critical parameter in micro-scale two-phase system. Several previous results indicate that wettability has influential affect on two-phase flow pattern in a microchannel. However, previous studies conducted using circular microtube, which was made by conventional fabrication techniques. Although most applications for micro thermal hydraulic system has used a rectangular microchannel, data for the rectangular microchannel is totally lack. In this study, a hydrophilic rectangular microchannel was fabricated using a photosensitive glass. And a hydrophobic rectangular microchannel was prepared using silanization of glass surfaces with OTS (octa-dethyl-trichloro-siliane). Experiments of two-phase flow in the hydrophilic and the hydrophobic rectangular microchannels were conducted using water and nitrogen gas. Visualization of twophase flow pattern was carried out using a high-speed camera and a long distance microscope. Visualization results show that the wettability was important for two-phase flow pattern in rectangular microchannel. In addition, two-phase frictional pressure drop was highly related with flow patterns. Finally, Two-phase frictional pressure drop was analyzed with flow patterns.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.148.1-148
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• 2002

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.701-709
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• 1995

The realistic discharge coefficient for the critical How model of RELAP5/AOD3/KAERI are determined for the subcooled and too-phase critical flow by assessments of nine MARVIKEN Critical flew Test(CFT). The selected test runs include a high initial subcooling and large nozzle aspect rat-io(L/D). The code assessment results show that RELAP5/MOD3/KAERI over-predicts the subcooled critical flow and under-predicts the two-phase critical flow. Using these result, the realistic discharge coefficients of critical flow models are quantified by an iterative method. The realistic discharge coefficients are determined to be 0.89 for the subcooled critical How and 1.07 for the two-phase critical flow, and the associated standard deviations are 0.0349 and 0.1189, respectively. The results obtained from this study can be applied to calculate the realistic system response of Large Break Loss of Coolant Accident and to evaluate the realistic Emergency Core Cooling System performance.