• Fire Science and Engineering
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• v.21 no.4
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• pp.25-31
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• 2007

The combustion characteristics and combustion heat of sandwich panel cores were analysed using variable external irradiation level. The characteristics such as ignition time, critical heat flux, ignition temperature and surface temperature profile were measured. Fuel samples were exposed to incident heat fluxes from 15 to $50\;kW/m^2$. For the measurement of various combustion characteristics, the size of specimen was $100\;mm\;{\times}\;100\;mm\;{\times}\;50\;mm$ and the samples were 3 different kinds. As results, Type B showed the best characteristics in measurement of combustion heat and ignition temperature and Type C showd the best characteristics in critical heat flux and surface temperature profile than that of the other two. In conclusion, we knew that Type C had the best performance in fire safety from all data of this study.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2002.07a
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• pp.74-77
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• 2002

The channel of the superconducting Flux Flow Transistor has been fabricated with plasma etching method using ICP. The ICP conditions were 700 W of ICP power, 150 W of rf chuck power, 5 mTorr of the pressure in chamber and 1:1 of Ar : Cl$_2$, respectively. The channel etched by plasma gas showed superconducting characteristics of over 77 K and superior surface morphology. The critical current of SFFT was altered by varying the external applied current. As the external applied current increased from 0 to 12 mA, the critical current decreased from 28 to 22 mA. Then the obtained r$\sub$m/ values were smaller than 0.1Ω at a bias current of 40 mA. The current gain was about 0.5. Output resistance was below 0.2 Ω.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.348-359
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• 1997

This paper reviews cooling aspects of the diverter system in Tokamak fusion devices with primary emphasis on the critical heat flux (CHF) issues for oater-cooled designs. General characteristics of four (4) coolant options for diverter cooling gases, oater, liquid metal, and organic liquid - are discussed first, focusing on the comparison of advantages and disadvantages of those options. Then results of recent studies on the high-heat-flux CHF of water at subcooled high-velocity conditions are reviewed to provide a general idea on the feasibility of the water-cooled diverter concept for future Tokamak fusion reactors. Water is assessed to be the most viable and practical coolant option for diverters of future experimental Tokamaks.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.349-355
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• 1996

In this study, the mechanistic critical heat flux (CHF) model and correlation for water are derived based on flow excursion (or Ledinegg instability) criterion and the simplified two-phase homogeneous model. The relationship between CHF for the water and the principal parameters such as mass flux heat of vaporization, heated length-to-diameter ratio, vapor-liquid density ratio and inlet subcooling is derived on the developed correlation. The developed CHF correlation predicts very well at the applicable ranges, 1 < P < 40 bar, 1, 300 < G 27, 00 kg/$m^2$s and inlet quality is less than -0.1. The overall mean ratio of predicted to experimental CHF value is 0.988 with standard deviation of 0.046.

• Transactions on Electrical and Electronic Materials
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• v.6 no.6
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• pp.284-288
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• 2005

We have systematically studied the superconducting properties and flux pinning enhancement of $(Y_{0.33}Sm_{0.33}Nd_{0.33})\;Ba_2Cu_3O_y$ [(YSN)-123] composite oxides by melt growth process in air. A sample prepared by this method showed well-textured microstructure, and $(Y_{0.33}Sm_{0.33}Nd_{0.33})\;BaCuO_5$ [(YSN)211] nonsuperconducting particles were uniformly dispersed in large (YSN) 123 superconducting matrix. The sample showed a sharp superconducting transition at 91 K. The magnetization measurements of the (YSN)-123 sample exhibited the enhanced flux pinning, compared with $YBa_2Cu_3O_y$ (Y-123) sample without Sm and Nd. Critical current densities of (YSN)-123 sample was $2.5{\times}10^4 A/cm^2$ at 2 T and 77 K.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.546-551
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• 1997

This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variations in pressure, tube diameter and length, mass flux and inlet subcooling.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.335-346
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• 2013

The Critical Heat Flux (CHF) of water with dispersed alumina nanoparticles was measured for the geometry and flow conditions relevant to the In-Vessel Retention (IVR) situation which can occur during core melting sequences in certain advanced Light Water Reactors (LWRs). CHF measurements were conducted in a flow boiling loop featuring a test section designed to be thermal-hydraulically similar to the vessel/insulation gap in the Westinghouse AP1000 plant. The effects of orientation angle, pressure, mass flux, fluid type, boiling time, surface material, and surface state were investigated. Results for water-based nanofluids with alumina nanoparticles (0.001% by volume) on stainless steel surface indicate an average 70% CHF enhancement with a range of 17% to 108% depending on the specific flow conditions expected for IVR. Experiments also indicate that only about thirty minutes of boiling time (which drives nanoparticle deposition) are needed to obtain substantial CHF enhancement with nanofluids.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2839-2846
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• 2021

In this study, experiments of downward-facing flow boiling were conducted to investigate material effects on CHF. Experiments were conducted using aluminum, copper, and carbon steel. It was found that different materials had different CHFs. Aluminum has the biggest CHF while copper has the lowest CHF for each mass flux. After experiment, surface wettability increased and surface became rougher, which was probably due to the oxidation process during nucleate boiling. The CHF difference is likely to be related to the surface wettability, roughness and thermal effusivity, which influences the bubble behavior and in turn affects CHF. Further studies are needed to determine which factor is dominant.

• Design & Manufacturing
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• v.13 no.1
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• pp.42-47
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• 2019

Nucleate boiling heat transfer is one of the most important phenomenon in the various industries. Especially, critical heat flux (CHF) refers to the upper limit of the pool boiling heat transfer region. Therefore, many researchers have found that CHF can be significantly increased by adding very small amounts of nanoparticles. In this study, the CHF and heat transfer coefficient were tested under the pool boiling state using copper and multi wall carbon nanotube nanoparticles. The results showed that two different types of nanoparticles deposited on the surface of two specimens made of the same material increased the heat flux in the nanoparticles with high conductivity, and there was no difference in the critical heat flux when the same material nanoparticles were deposited on the two different specimen surfaces.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2303-2308
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• 2007

Experiments were performed to determine the thermal (or turbulent) diffusion coefficient (TDC) and to investigate the critical heat flux (CHF) performance in the 5${\times}$5 rod bundle with 5 unheated rods which are supported by Hybrid Mixing Vane. In this study, HFC-134a fluid was used as working fluid and the fluid temperature were measured in the important subchannels. To determine the TDC value, the measured fluid temperatures were compared with the predicted values obtained from the MATRA code. The best optimized value of ${\beta}$ was found to be 0.02 by considering prediction statistics, i.e., average and standard deviations of the differences between the experimental results and code calculations. Using the best optimized value of ${\beta}$ as 0.02, the MATRA code predicts the test results of the fluid temperature within ${\pm}$1.0 % of error. According to the experimental results on CHF of 5 non-heating guide tubes, the case with non-heating guide tube showed a little good performance in terms of CHF.