• 제목/요약/키워드: Vertical Downward flow

검색결과 54건 처리시간 0.027초

비비등 수직 하향 유동의 대류 열전달 특성 (The Characteristics of Convective Heat Transfer in Non Boiling Vertical Downard Flow)

  • 이동상;김재근;양희준;오율권;차경옥
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2000년도 추계학술대회논문집B
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    • pp.118-123
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    • 2000
  • This experimental study was conducted to figure out the characteristics of convective heat transfer in non boiling vertical downward flow with polymer additives. This experiment was studied in 26mm diameter, 800mm heating length and $1{\times}10^5W/m^2$ heat flux. The polymer concentration ranged from 0PPM to 500PPM with corresponding from Reynolds number $3.3{\times}10^4$ to $6.8{\times}10^4$ in non boiling vertical downward flow. Experimental results show that the characteristics of convective heat transfer was a strong function of polymer concentration and it has decreased with increasing the polymer concentration in non boiling vertical downward flow.

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Flow regime transition criteria for vertical downward two-phase flow in rectangular channel

  • Chalgeri, Vikrant Siddharudh;Jeong, Ji Hwan
    • Nuclear Engineering and Technology
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    • 제54권2호
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    • pp.546-553
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    • 2022
  • Narrow rectangular channels are employed in nuclear research reactors that use plate-type nuclear fuels, high heat-flux compact heat exchangers, and high-performance micro-electronics cooling systems. Two-phase flow in narrow rectangular channels is important, and it needs to be better understood because it is considerably different than that in round tubes. In this study, mechanistic models were developed for the flow regime transition criteria for various flow regimes in co-current air-water two-phase flow for vertical downward flow inside a narrow rectangular channel. The newly developed criteria were compared to a flow regime map of downward air-water two-phase flow inside a narrow rectangular channel with a 2.35-mm gap width under ambient temperature and pressure conditions. Overall, the proposed model showed good agreement with the experimental data.

2상류의 장거리 수송시 효율적인 열관리에 관한 실험적 연구 (Experimental Study on the Efficient Control of Heat of Lc Distance Transport for Two- Phase Fluid)

  • 김재호;김재근;오율권;차경옥
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2001년도 춘계학술대회논문집E
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    • pp.119-124
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    • 2001
  • This experimental study was conducted to figure out the characteristics of convective heat transfer non boiling vertical downward flow with polymer additives. This experiment was studied in diameter, 800mm heating length and $1{\times}10^5 W/m^2$ heat flux. The polymer concentration ranged 0ppm to 500ppm with corresponding from superficial liquid velocity 1.25m/s to 2.5m/s in non bo vertical up and downward flow. Experimental results show that the characteristics of convective transfer was a strong function of polymer concentration and it has decreased with increasing polymer concentration in non boiling up and vertical downward flow.

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고분자 물질 첨가에 의한 2상 유동의 마찰 항력 감소와 대류 열전달 특성 (The Drag Reduction and Convective Heat Transfer Characteristics of Two-Phase Flow with Polymer Additives)

  • 이동상;김재근;차경옥
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2000년도 춘계학술대회논문집B
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    • pp.71-76
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    • 2000
  • This experimental study was conducted to figure out the drag reduction and convective heat transfer in vertical downward two-phase flow with polymer additives. The drag reduction effect were analyzed by using the difference of the pressure drop between the flow with polymer additives and without it. Experimental results show that the pressure drop with polymer additives is less than the pressure drop without polymer in vertical downward two-phase flow. And the convective heat transfer has decreased with increasing the polymer concentration in vertical downward two-phase flow.

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Development of a one-dimensional system code for the analysis of downward air-water two-phase flow in large vertical pipes

  • Donkoan Hwang;Soon Ho Kang;Nakjun Choi;HangJin Jo
    • Nuclear Engineering and Technology
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    • 제56권1호
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    • pp.19-33
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    • 2024
  • In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphon-breaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as / increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.

Plant-scale experiments of an air inflow accident under sub-atmospheric pressure by pipe break in an open-pool type research reactor

  • Donkoan Hwang;Nakjun Choi;WooHyun Jung;Taeil Kim;Yohan Lee;HangJin Jo
    • Nuclear Engineering and Technology
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    • 제55권5호
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    • pp.1604-1615
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    • 2023
  • In an open-pool type research reactor with a downward forced flow in the core, pipes can be under sub-atmospheric pressure because of the large pressure drop at the reactor core in the atmospheric pool. Sub-atmospheric pressure can result in air inflow into the pipe from the pressure difference between the atmosphere and the inside of the pipe, which in a postulated pipe break scenario can lead to the breakdown of the cooling pump. In this study, a plant-scale experiment was conducted to study air inflow in large piping systems by considering the actual operational conditions of an advanced research reactor. The air inflow rate was measured, and the entrained air was visualized to investigate the behavior of air inflow and flow regime depending on the pipe break size. In addition, the developed drift-flux model for a large vertical pipe with a diameter of 600 mm was compared with other correlations. The flow regime transition in a large vertical pipe under downward flow was also studied using the newly developed drift-flux model. Consequently, the characteristics of two-phase flow in a large vertical pipe were found to differ from those in small vertical pipes where liquid recirculation was not dominant.

수직 동심 환형관 내의 난류혼합대류 현상에 관한 직접수치모사 (Direct Numerical Simulation of Turbulent Mixed Convection in Heated Vertical Annulus)

  • 전용준;배중헌;유정열
    • 대한기계학회논문집B
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    • 제33권9호
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    • pp.674-681
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    • 2009
  • Turbulent mixed convection in heated vertical annulus is investigated using Direct Numerical Simulation (DNS) technique. The objective of this study is to find out the effect of buoyancy on turbulent mixed convection in heated vertical annulus. Downward and upward flows with bulk Reynolds number 8500, based on hydraulic diameter and mean velocity, have been simulated to investigate turbulent mixed convection by gradually increasing the effect of buoyancy. With increased heat flux, heat transfer coefficient first decreases and then increases in the upward flow due to the effect of buoyancy, but it gradually increases in downward flow. The mean velocity and temperature profiles can not be explained by the wall log laws due to the effect of buoyancy, too. All simulation results are in good quantitative agreement with existing numerical results and in good qualitative agreement with existing experimental results.

수직 동심 환형관 내의 난류혼합대류 현상에 관한 직접수치모사 (Direct numerical simulation of turbulent mixed convection in heated vertical annulus)

  • 전용준;배중헌;유정열
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2008년도 추계학술대회B
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    • pp.2759-2764
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    • 2008
  • Turbulent mixed convection in heated vertical annulus is investigated using Direct Numerical Simulation (DNS) technique. The objective of this study is to find out the effect of buoyancy on turbulent mixed convection in heated vertical annulus. Downward and upward flows with bulk Reynolds number 8500, based on hydraulic diameter and mean velocity, have been simulated to investigate turbulent mixed convection by gradually increasing the effect of buoyancy. With increased heat flux, heat transfer coefficient first decreases and then increases in the upward flow due to the effect of buoyancy, but it gradually increases in downward flow. The mean velocity and temperature profiles can not be explained by the wall log laws due to the effect of buoyancy, too. All simulation results are in good quantitative agreement with existing numerical results and in good qualitative agreement with existing experimental results.

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Critical Heat Flux and Flow Pattern for Water Flow in Annular Geometry

  • Park, Jae-Wook;Baek, Won-Pil;Chang, Soon-Heung
    • 한국원자력학회:학술대회논문집
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    • 한국원자력학회 1996년도 추계학술발표회논문집(1)
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    • pp.224-229
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    • 1996
  • An experimental study on critical heat flux (CHF) and two-phase flow visualization has been performed for water flow in internally-heated, vertical, concentric annuli under near atmospheric pressure. Tests have been done under stable forced- circulation, upward and downward flow conditions with three test sections of relatively large gap widths (heated length = 0.6 m. inner diameter = 19 mm, outer diameter = 29, 35 and 51 mm). The outer wall of the test section was made up of the transparent Pyrex tube to allow the observation of flow patterns near the CHF occurrence. The CHF mechanism was changed in the order of flooding, chum-to-annular flow transition, and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. Observed parametric trends are consistent with the previous understanding except that the CHF for downward flow is considerably lower than that for upward flow.

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수직다발관형 빙축열 탱크내 물의 응고과정시 열전달특성에 관한 연구 (An Experimental Study on the Heat Transfer Characteristics during the Freezing Process of Water in the Vertical Multi Tube Type Ice Storage Tank)

  • 김영기;임장순
    • 태양에너지
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    • 제18권3호
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    • pp.95-105
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    • 1998
  • In this study, basic design data which were required for development of highly efficient ice storage system with low temperature latent heat were experimentally obtained. The ice storage system considered in this study was the one that has been widly used in the developed country and called the ice-on-coil type. Using the system, the ice storage performance for various design parameters which were the flow direction and the inlet temperature of the secondary fluid was tested. In addition, the timewise variation of the interface profiles between the solid and the liquid were visualized, and the heat transfer characteristics of the Phase Change Material(PCM) in the ice storage tank were Investigated. During the freezing processes in the ice storage tank with several vertical tubes, decrease of the heat transfer area and the heat resistance of the ice layer made the increasing rate of ice packing factor(IPF) less. The total freezing energy for the upward flow of the secondary fluid was higher than that for the downward flow. The average ice storage efficiency for the upward flow of the secondary fluid was higher than that for the downward flow.

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