• 대한기계학회논문집B
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• 제30권7호
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• pp.646-655
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• 2006

The present research is an experimental study on characteristics of cooling behavior for the adjacent copper blocks with surface roughness or micro-porous coated surface. The experiments were carried out at saturation state or within subcooled states of PF5060. The effects of heater orientation and the intervals between heating surfaces or substrates were investigated under various heat flux conditions. The boiling performance of copper block with micro-porous coated surface was better than that of copper block with surface roughness. It is understood that the bubble sweeping enhances boiling performance for the heaters with inclinations of $\theta=45^{\circ}\;and\;\theta=90^{\circ}$, where as the bubble flattening decreases boiling performance for the heaters with inclinations of $\theta=135^{\circ}\;and\;\theta=180^{\circ}$. In comparison to upper heater and below heater with orientation, the upper heater has lower superheat temperature than the below heater due to the bubble sweeping. It is also found that boiling performance decreases in the case of adjacent double heaters with only 0.2cm substrate interval.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1851-1859
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• 2022

Experimental research on flow characteristics of open natural circulation system was performed, to figure out the mechanism of the open natural circulation behaviors. The influence factors, such as the heating power, the inlet subcooled and the level of cooling tank on the flow characteristics of the system were examined. It was shown that within the scope of the experimental conditions, there are five flow types: single-phase stable flow, flash and geyser coexisting unstable flow, flash stable flow, flash unstable flow, and flash and boiling coexisting unstable flow. The geyser flow in flash and geyser coexisting unstable flow is different from classic geysers flow. The flow oscillation period and amplitude of the former are more regular, is a newly discovered flow pattern. By drawing the flow instability boundary diagram and sorting out the flow types, it is found that the two-phase unstable flow is mainly characterized by boiling and flash, which determine the behavior of open natural circulation respectively or jointly. Moreover, compared with full liquid level system, non-full liquid level system is more prone to boiling phenomenon, and the range of heat flux density and undercooling degree corresponding to unstable flow is larger.

• 한국가스학회지
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• 제21권4호
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• pp.92-102
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• 2017

피셔-트롭쉬 합성반응은 CO와 H2의 혼합가스로 이루어진 합성가스를 부가가치가 높은 탄화수소 제품으로 변환시킨다. 본 논문에서는 저온 피셔-트롭쉬 합성반응과 단일, 다중 마이크로채널 반응기에 패킹시킨 촉매를 기반으로 강화된 반응조건의 열전달을 고려하여 전산유체역학 기반의 시뮬레이션을 진행하고 분석하였다. 단일채널모델을 통하여 CO 전환률이 ~65% 이상, $C_{5+}$ 선택도가 ~74% 이상을 달성하면서도 Co 기반의 super-active 촉매를 통해 GHSV를 $30000hr^{-1}$을 달성할 수 있음을 보였다. 다중 마이크로채널 반응기모델에서는 열전달 시뮬레이션을 동시에 해석하여, 3가지의 다른 반응기구조에 대해서, 직교류 wall boiling 냉매를 사용시 ${\Delta}T_{max}$가 23 K였으며 평행유동 subcooled 냉매와 평행유동 wall boiling 냉매의 경우 각각 15 K와 13 K의 ${\Delta}T_{max}$를 보였다. 반응기 전체적으로 498 - 521 K에서 온도제어가 가능했으며 계산된 사슬성장 가능성은 저온 피셔-트롭쉬 합성에 적합한 것으로 보인다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.337-342
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• 1996

The pool-boiling critical heat flux (CHF) of water on small flat plates has been experimentally investigated focusing on the effects of the inclination angle and size of the heated surface under near atmospheric pressure. The second-phase experiment was accomplished to find out the general CHF behavior for over-all inclination angles from -90$^{\circ}$ to 90$^{\circ}$using two plate-type test sections (30$\times$150 mm and 40$\times$150 mm) submerged in a slightly subcooled water pool. Test results generally confirm the first-phase findings and show little effect of inclination angle for inclined upward-facing cases. CHF position moves to lower position with the increase of the heater characteristic size and inclination angle(from -30$^{\circ}$to 60$^{\circ}$).

• 한국수소및신에너지학회논문집
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• 제29권1호
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• pp.105-110
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• 2018

The flow boiling heat transfer of water was experimentally investigated on plain and sintered microporous surfaces in a mini-channel. The effects of microporous coating on flow boiling heat transfer of subcooled water were investigated in a 300 mm long mini-channel with a cross section of $20{\times}10mm^2$. The test section has sufficiently long entrance length of 300 mm which provides a fully-developed flow before the channel inlet. The bottom side of the channel was heated by a copper block assembled with a high-density cartridge heater and other sides of the channel were insulated. The microporous surface was fabricated by sintering copper particles with the average particle size of $50{\mu}m$ on the top side of the copper block. Heat transfer measurement was conducted at the mass flux of $208kg/m^2s$ and the heat flux up to $500kW/m^2$. Microporous coated surface showed an earlier boiling incipience compared with plain surface regardless of the mass flux. Microporous coating were significantly attributed to local wall temperature and local heat transfer coefficient for flow boiling.

• Journal of Advanced Marine Engineering and Technology
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• 제10권2호
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• pp.156-164
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• 1986

In the case of boiling on high temperature wall, vapor film covers fully or parcially the surface. This phenomenon, film boiling or transition boiling, is very important in the surface heat treatment of metal, design of cryogenic heat exchanger and emergency cooling of nuclear reactor. Mainly supposed hydraulic-thermal accidents in nuclear reactor are LCCA (Loss of Coolant Accident) and PCM (Power-Cooling Mismatch). Recently, world-wide studies on reflooding of high temperature rod bundles after the occurrence of the above accidents focus attention on wall temperature history and required time in transient cooling process, wall superheat at rewet point, heat flux-wall superheat relationship beyond the transition boiling region, and two-phase flow state near the surface. It is considered that the further systematical study in this field will be in need in spite of the previous results in ref. (2), (3), (4). The paper is the study about the fast transient cooling process following the wall temperature excursion under the CHF (Critical Heat Flux) condition in a forced convective subcooled boiling system. The test section is a vertically arranged concentric annulus of 800 mm long and 10 mm hydraulic diameter. The inner tube, SUS 304 of 400 mm long, 8 mm I.D, and 7 mm O.D., is heated uniformly by the low voltage AC power. The wall temperature measurements were performed at the axial distance from the inlet of the heating tube, z=390 mm. 6 chromel- alumel thermocouples of 76 .mu.m were press fitted to the inner surface of the heating tube periphery. To investigate the heat transfer characteristics during the fast transient cooling process, the outer surface (fluid side) temperature and the surface heat flux are computed from the measured inner surface temperature history by means of a numerical method for inverse problems of transient heat conduction. Present cooling (boiling) curve is sufficiently compared with the previous results.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.108-113
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• 1996

Several analytical models for the departure from nucleate boiling (DNB) phenomenon have been developed during the last decade. Among these, Chang & Lee's model based on a bubble crowding mechanism is remarkable in the fundamental features characterized as the formulation of mass, energy, and momentum balance equation at thermal-hydraulic conditions leading to the DNB. However, Bricard and Souyri remarked that the assumption of stagnant bubbly layer at the DNB condition is questionable and the signs on the axial projections of the momentum fluxes at the core/bubbly layer interface in the momentum balance equations are erroneous. From this remark, Chang & Lee's model has been re-examined and modified by correcting the erroneous treatments in the momentum balance equations and removing the spurious assumptions. The revised model predicts well the extensive DNB data of water in uniformly heated tubes at low qualities and shows more accurate prediction compared with the original model.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3754-3767
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• 2023

Turbulent jet discharges in subcooled water pools are essential for safety systems in nuclear power plants, specifically in the pressure suppression pool of boiling water reactors and In-containment Refueling Water Storage Tank of advanced pressurized water reactors. The gas and liquid flow in these systems is investigated using multiphase flow analysis. This field has been extensively examined using a combination of experiments, theoretical models, and Computational Fluid Dynamics (CFD) simulations. ANSYS CFX offers two approaches to model multiphase flow behavior. The non-homogeneous Eulerian-Eulerian Model has been used in this work; it computes global information and is more convenient to study interpenetrated fluids. This study utilized the Large Eddy Simulation Model as the turbulence model, as it is better suited for non-stationary and buoyant flows. The CFD results of this study were validated with experimental data and theoretical results previously obtained. The figures of merit dimensionless penetration length and the dimensionless buoyancy length show good agreement with the experimental measurements. Correlations for these variables were obtained as a function of dimensionless numbers to give generality using only initial boundary conditions. CFD numerical model developed in this research has the capability to simulate the behavior of non-condensable gases discharged in water.

• Journal of Advanced Marine Engineering and Technology
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• 제18권5호
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• pp.12-23
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• 1994

The two-phase flow is observed in power plants, chemical process plants, and refrigeration systems etc., and it is very important to solve the heat transfer mechanism of a boiler, an automic reactor, a condenser and various types of evaporators. Recently, the problem of two phase heat transfer is braught up in many regions with development of energy saving technique. In flow boiling system it is necessary to store data in each condition because the heat transfer characteristics of flow boiling region vary by the change of flow pattern and the magnetude of heat flux to tube length, and be subtly affected by the flow and heating condition. So basic study for knowing flow pattern in heat transfer region and the relation between heat transfer characteristic and flow condition is desired to accumulate data in wide variety of liquid and flow system in the study of heat transfer of two phase flow. In this study R-113 was selected as working fluid whose properties were programmed by least square method, and experiment was conducted in the region of mass flow $1.628{\times}10^6$~$4.884{\times}10^6$/kg/$m^2$hr with inlet subcooling 10~3$0^{\circ}C$, sustaining test section inlet pressure to 1.5kg$_f$/$cm^2$abs.

• 한국안전학회지
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• 제32권2호
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• pp.147-152
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• 2017

The external reactor vessel cooling (ERVC) is well known strategy to mitigate a severe accident at which nuclear fuel inside the reactor vessel is molten. In order to compare the heat removal capacity of ERVC between the nuclear reactor designs quantitatively, numerical method is often used. However, the study for ERVC using computational fluid dynamics (CFD) is still quite scarce. As a validation study on the numerical prediction for ERVC using CFD, the subcooled boiling flow and natural circulation of coolant at the ULPU-V experiment was simulated. The commercially available CFD software ANSYS-CFX was used. Shear stress transport (SST) model and RPI model were used for turbulence closure and wall-boiling, respectively. The averaged flow velocities in the downcomer and the baffle entry under the reactor vessel lower plenum are in good agreement with the available experimental data and recent computational results. Steam generated from the heated wall condenses rapidly and coolant flows maintains single-phase flow until coolant boils again by flashing process due to the decrease of saturation temperature induced by higher elevation. Hence, the flow rate of coolant natural circulation does not vary significantly with the change of heat flux applied at the reactor vessel, which is also consistent with the previous literatures.