• Transactions of the Korean Society of Mechanical Engineers B
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• v.26 no.10
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• pp.1419-1426
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• 2002

An experimental study was made to elucidate critical heat flux(CHF) characteristics in a two-phase concentric-tube thermosyphon. The experiment was performed by using saturated water, over the experimental range of configuration: inner diameter of heated outer tube D=12mm, outer diameter of unheated inner tube do=3 to 10mm and heated tube length L=100 to 1000mm. The experiment shows that the CHF is enhanced with increase in the inner tube diameter, and that the CHF decreases beyond a certain diameter of the inner tube. There is an optimum diameter for inner tube that maximizes the CHF, for each tube length and test liquid. The CHF maximum is about two to eight times as large as that without an inner tube. For a large inner tube, the CHF characteristics is similar to that for natural convective boiling in a vertical annular tube.

• Nuclear Engineering and Technology
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• v.29 no.6
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• pp.49-59
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• 1997

본 기술 보고는 임계 열유속(Critical Heat Flux; CHF)을 예측하기 위해 사용되고 있는 상관식의 형태와 적용 방법, 이에 따른 예측 오차와 여유도의 변화 등을 종합적으로 분석한다. CHF 현상에 대해서는 지난 반 세기 동안 발생 메커니즘, 예측 모델, 설계에의 적용 방법 등에 대한 연구가 광범위하게 수행되어 대부분의 운전 조건에 대해 신뢰할만한 예측 모델들이 확립되어 있다. 그러나 예측 모델의 이용에서 가장 중요한 기준이 되는 예측 오차의 의미가 잘못 이해되는 경우가 많으므로, 이 글에서는 예측 모델의 형태 및 적용 방법에 따라 예측 오차가 달라지는 원인을 명확하게 해석하고, 실제 계산을 통하여 예시하였다. 그리고 상관식 형태 및 이용 방법에 따라 임계 열유속비(Critical Heat Flux Ratio: CHFR)와 임계 출력비(Critical Power Ratio; CPR)가 어떠한 관계를 갖는가를 논의하였다.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.292-295
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• 2008

In a SCWR (SuperCritical pressure Water cooled Reactor), the coolant temperature initially at below the pseudo-critical temperature at the bottom of a reactor core increases as the coolant flows upward through the sub-channels of the fuel assemblies, and it finally becomes higher than the pseudo-critical temperature when it leaves the reactor core. At certain conditions, heat transfer deterioration occurs near the pseudo-critical temperature and it may cause a drastic rise of the fuel surface temperature resulting a fuel failure. Therefore, an accurate estimation of the heat transfer coefficient is very important for the thermal-hydraulic design of a reactor core. An experiment on heat transfer to the vertically upward flowing $CO_2$ at a supercritical pressure in a circular tube were performed at KAERI. The internal diameter of the test section is 6.32 mm, which corresponds to the hydraulic diameter of a sub-channel in the conceptional design proposed by KAERI. The test range of the mass flux is 285 to 1200 kg/m$^2$s and the maximum heat flux is 170 kW/m$^2$. The inlet pressure is maintained at 8.12 MPa, which is 1.1 times the critical pressure. A new correlation, which covers both the normal and deterioration heat transfer regimes was proposed and compared with the estimations by exiting correlations.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.08a
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• pp.172-172
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• 2010

$MgB_2$ doped with $TiO_2$ was prepared by the in-situ solid state reaction to study the effects of $TiO_2$ dopant on the flux pinning behavior of $MgB_2$ superconductor. From the field-cooled and the zero-field-cooled temperature dependences of magnetization, the realms of vortex-glass and vortex-liquid states of $TiO_2$-doped $MgB_2$ were determined in the H-T diagram (the temperature dependence of upper critical magnetic field and irreversibility line). The critical current density was estimated from the width of hysteresis loops in the framework of Beam's model at different temperatures. The results indicate that nano-scale $TiO_2$ inclusions play a role of the effective pinning centers and lead to the enhanced upper critical field and critical current density. It is suggested that the grain-boundary pinning mechanism is realized in $TiO_2$-doped $MgB_2$ superconductor.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2396-2401
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• 2007

Freon CHF experiments are carried out to investigate the CHF enhancements by mixing vane shapes of spacer grids in nuclear fuel assembly. The experiments were performed for a wide range mass flux, 50$\sim}$3000 kg/$m^2s$. Three kinds of spacer grids in 5${\times}$5 rod bundles are tested: no mixing vane grids, hybrid mixing vane grids, and split mixing vane grids. The CHF performances are compared along with the data belong to the PWR operating conditions based on a water equivalence through a fluid-to-fluid modeling method. The average of the data in this range is 16.4% for 37 data of hybrid vane grid and 12.5% for 24 data of split vane. In the lower mass flux, however, the split vane grid shows slightly higher performance than the hybrid vane grid.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.379-394
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• 2000

As an in-vessel retention (IVR) design concept in coping with a severe accident in the nuclear power plant during which time a considerable amount of core material may melt, external cooling of the reactor vessel has been suggested to protect the lower head from overheating due to relocated material from the core. The efficiency of the ex-vessel management may be estimated by the thermal margin defined as the ratio of the critical heat flux (CHF）to the actual heat flux from the reactor vessel. Principal factors affecting the thermal margin calculation are the amount of heat to be transferred downward from the molten pool, variation of heat flux with the angular position, and the amount of removable heat by external cooling In this paper a thorough literature survey is made and relevant models and correlations are critically reviewed and applied in terms of their capabilities and uncertainties in estimating the thermal margin to potential failure of the vessel on account of the CHF Results of the thermal margin calculation are statistically treated and the associated uncertainties are quantitatively evaluated to shed light on the issues requiring further attention and study in the near term. Our results indicated a higher thermal margin at the bottom than at the top of the vessel accounting for the natural convection within the hemispherical molten debris pool in the lower plenum. The information obtained from this study will serve as the backbone in identifying the maximum heat removal capability and limitations of the IVR technology called the Cerium Attack Syndrome Immunization Structures (COASISO) being developed for next generation reactors.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.25 no.9
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• pp.1237-1244
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• 2001

An experimental study was carried out to identify the various regimes of natural convective pool boiling and to determine the boiling heat transfer curve and Critical Heat Flux(CHF) on a vertical square surface having a 70mm width and a 70mm height. The heater made of copper block with embedded cartridge heaters is submerged in a water tank at atmospheric pressure. As the heat flux increases from 100kW/㎡ to 1.2MW/㎡, the heat transfer regime migrates from the nucleate boiling to the film boiling. The boiling heat transfer data are fitted by Rohsenow type correlation. An explosive vapor generation on the heated surface, whose size and frequency are characterized by the heat flux, is visualized using a high speed digital imaging system.

• Journal of the Korean Society for Heat Treatment
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• v.18 no.6
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• pp.375-382
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• 2005

This paper presents the results of experimental investigation for the effect of heat conduction on the evaporation cooling of water droplet in the process of heat treatment. The experiments are mainly focused on the surface temperature, the surface roughness and the droplet diameter at aluminum. The range of surface temperature is from $80^{\circ}C$ to $140^{\circ}C$, surface roughness is from $R_a=0.18{\mu}m$ to $R_a=1.36{\mu}m$ and droplet diameter is from 2.4 mm to 3.0 mm. The results show that the total evaporation time is shorter for the larger surface roughness, the time averaged heat flux has maximum value for the larger surface roughness and exist the critical heat flux. The total evaporation time has a big influence on the evaporation region for the smaller droplet size, but the total evaporation time has not influence on the nuclear boiling region.

• Nuclear Engineering and Technology
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• v.40 no.5
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• pp.387-396
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• 2008

Many experimental analyses for annular film dryouts, which is one of the Critical Heat Flux (CHF) mechanisms, have been performed because of their importance. Numerical approaches must also be developed in order to assess the results from experiments and to perform pre-tests before experiments. Various thermal-hydraulic codes, such as RELAP, COBRATF, MARS, etc., have been used in the assessment of the results of dryout experiments and in experimental pre-tests. These thermal-hydraulic codes are general tools intended for the analysis of various phenomena that could appear in nuclear power plants, and many models applying these codes are unnecessarily complex for the focused analysis of dryout phenomena alone. In this study, a numerical model was developed for annular film dryout using the drift-flux model from uniform heated tube geometry. Several candidates of models that strongly affect dryout, such as the entrainment model, deposition model, and the criterion for the dryout point model, were tested as candidates for inclusion in an optimized annular film dryout model. The optimized model was developed by adopting the best combination of these candidate models, as determined through comparison with experimental data. This optimized model showed reasonable results, which were better than those of MARS code.

• KIEE International Transactions on Electrophysics and Applications
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• v.5C no.5
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• pp.204-210
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• 2005

Alternating current (AC) losses of two Bi-2223 ([Bi, Pb]: Sr: Ca: Cu: O = 2:2:2:3) tapes [(Tape I, un-twist-pitch) and the other with a twist-pitch of 10 mm (Tape II)] were measured and compared. These samples, produced by the powder-in-(Ag) tube (PIT) method, are multi-filamentary. Also, it's produced by non-twist and different twist pitch (8, 10, 13, 30, 50 and 70 mm). The critical current measurement was carried out under the environment in liquid Nitrogen and in zero-field by 4-probe method. Susceptibility measurements were conducted while cooling in a magnetic field. Flux loss measurements were conducted as a function of ramping rate, frequency and field direction. The AC flux loss increases as the twist-pitch of the tapes decreased, in agreement with the Norris Equation. Neutron-diffraction measurements have been carried out investigate the crystal structure, magnetic structures, and magnetic phase transitions in Bi-2223([Bi, Pb]:Sr:Ca:Cu:O)