• 대한기계학회논문집B
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• 제37권4호
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• pp.343-352
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• 2013

본 연구에서는 열전달 표면의 형상과 그 위에서의 유동 속도의 변화에 따른 풀 비등 열전달계수의 변화를 살펴보기 위해 평판, 낮은 핀, Thermoexcel-E, Turbo-B 표면을 사용하여 유동 속도를 변화시켜가며 임계 열유속까지 열전달계수를 측정하였다. 작동 유체로는 증류수를 사용하였고 사각 평면 히터($9.53{\times}9.53mm$)를 이용하여 네 가지 표면에서 임계 열유속까지의 데이터를 얻을 수 있도록 장치를 제작하였고 $60^{\circ}C$에서 유동 속도를 0, 0.1, 0.15, 0.2m/s로 변화시켜가며 데이터를 취했다. 실험 데이터를 보면 모든 표면에서 유동이 있을 때의 임계 열유속은 유동이 없을 때에 비해 높은 것으로 나타났다. 또한 표면적의 증가와 기포 이탈에 충분한 핀 간격 등으로 인해 낮은 핀 표면의 임계 열유속은 평판이나 Turbo-B, Thermoexcel-E 표면보다 훨씬 놓았고 평판에 비해서는 무려 5배 정도의 향상을 보였다. 한편 대형 냉동기의 증발기용으로 개발된 Turbo-B와 Thermoexcel-E 표면은 물에서 기포의 이탈 지름이 크므로 열전달계수와 임계 열유속 모두 예상보다 큰 효과를 나타내지 않았다. $50kW/m^2$이하의 저열유속에서는 모든 표면에 대해 유동 속도 증가에 따라 열전달계수가 증가하였다. 결론적으로 핵발전소의 증기발생기에 적용하기에는 낮은 핀 형상의 표면이 가장 좋은 것으로 나타났다.

• 대한기계학회논문집B
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• 제41권9호
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• pp.597-603
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• 2017

다양한 온도 조건의 계면활성제 수용액 내에서 급속 냉각되는 고온 수직 금속봉의 최소막비등점에 대한 실험을 수행하였다. 액체로는 Triton X-100 수용액(100 wppm)과 순수(pure water)를 이용하였고, 액체의 온도는 $77^{\circ}C{\sim}100^{\circ}C$ 영역이었다. 고체 시편으로는 시편 중심의 초기 온도가 $500^{\circ}C$인 스테인레스 스틸(stainless steel) 수직봉을 이용하였다. Triton X-100 수용액과 순수에서, 액체의 온도가 감소함에 따라 최소막비등점의 도달시간은 감소하였고, 온도 및 열유속은 증가하였다. 한편, 본 실험 온도 영역에서, Triton X-100 수용액의 경우가 순수의 경우보다 최소막비등점의 도달시간은 길었고, 온도 및 열유속은 감소하는 경향을 나타냈다. 본 실험데이터를 토대로 고온 수직 금속봉에 대해서 Triton X-100 수용액과 순수에서의 최소막비등 온도에 대한 실험식을 제안하였다.

• 설비공학논문집
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• 제8권1호
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• pp.1-12
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• 1996

In lots of application to heat exchanger systems, closed two-phase thermosyphons are tilted from a horizontal. If the tilt angle, especially, is less than 30$^{\circ}$, the operational performances of thermosyphon are highly dependent on tilt angle. The present study was conducted to better understand such operational behaviors as mech-anni는 of phase change, and flow patterns inside a tilted thermosyphon. For experiment, an ethanol thermosyphon with a 35% of fill charge rate was designed and manufactured, using a copper tube with a diameter 19mm and a length 1500mm. Through a series of test, the tilt angle was kept constant at each of 4 different values in the range 10~25deg. and the heat supply to the evaporator was stepwisely increased up to 30㎾/$m^2$. When a steady state was established to the thermosyphon for each step of thermal loads, the wall temperature distribution and vapor temperature at the condenser were measured. The wall temperature distributions demonstrated a formation of dry patch in the top end zone of the evaporator, with a values of temperature 20~4$0^{\circ}C$ higher than the wetted surface for a moderate heat flux q≒20㎾/$m^2$. Inspite of the presence of hot dry patch, however, the mean values of boiling heat transfer coefficient at the evaporator wall were still in a good agreement with those predicted by Rohsenow's formula, which was based on nucleate boiling. For the condenser, the wall temperatures were practically uniform, and the measured values of condensation heat transfer coefficient were 1.7 times higher than the predicted values obtained from Nusselt's film condensation theory on tilted plate. Using those two expressions, a correlation was formulated as a function of heat flux and tilt angle, to determine the total thermal resistance of a tilted thermosyphon. The correlation formula showed a good agreement with the experimental data within 20%.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.561-574
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• 2009

First ex-vessel core catcher has been applied to the practical design of NPPs with VVER-1000 reactors built in China (Tyanvan) and India (Kudankulam) for severe accident management (SAM) and mitigation of SA consequences. The paper presents the concept and basic design of this crucible-type core catcher as well as an evaluation of its efficiency. The important role of oxidic sacrificial material is discussed. Insight into the behaviour of the molten pool, which forms in the catcher after core relocation from the reactor vessel, is provided. It is shown that heat loads on the water-cooled vessel walls are kept within acceptable limits and that the necessary margins for departure from nucleate boiling (DNB) and of vessel failure caused by thermo-mechanical stress are satisfactorily provided for.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.1219-1224
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• 2004

The objective of this study is to propose an experimental calibration facility in which a heat flux sensor can be calibrated under conductive condition by using helium gas. The heat flux calibration facility was designed, simulated and manufactured for use in a high heat transfer condition. It delivers heat fluxes up to a maximum of 35 KW $m^{-2}$. A copper block heated electrically with 3.5 KW power is designed to produce uniform temperature up to 600 K across its face. High heat fluxes are provided between hot plate and cold plate by 1 mm height helium filled gap. A cold plate is maintained around 300 K through pool boiling using a refrigerant and water-cooled heat exchanger. A simulation was conducted to verify the design of the main test section. To verify the performance of calibration facility, a heat flux sensor was examined. The measured heat fluxes were compared to the calculated one.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.409-414
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• 1997

Reflood model in RELAP5/MOD3.1 are modified to improve the unrealistic prediction results of the model. In the new method, the modified Zuber pool boiling critical heat flux (CHF) correlation is adopted. The reflood drop size is characterized by the use of We=1.5 and the minimum drop size of 0.0007 m for $p^{*}\;{\leq}\;0.025$. To describe the wall to vapor heat transfer at low pressure and low flow condition, the Webb-Chen correlation is utilized . The suggested method has been verified through the simulations of the Lehigh University rod bundle reflood tests. Through sensitivity study it is shown that the effect of drag coefficients is dominant in the reflood model. It is proved that the present modifications result in much more improved quench behavior and accurate wan and vapor temperature predictions.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.1512-1517
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• 2003

Augmentation of CHF by ultrasonic is experimentally studied under subcooling pool boiling condition. Experiment is carried out for downward-facing plate with and without the ultrasonic. The working fluid is distilled water. Experimental apparatus is composed of a bath, power supply, test section, ultrasonic generator, DAQ system. Experiment is performed with the subcooling temperature of $5^{\circ}C$, $20^{\circ}C$, $40^{\circ}C$ and the inclined angle of $0^{\circ}$, $10^{\circ}$, $20^{\circ}$, $45^{\circ}$, 90. From the experimental results, it is found that ultrasonic effect enhances CHF of the downward-facing plate. As increasing the degree of subcooling, the rate of CHF increase is enhanced. As increasing the inclined angle, the rate of CHF increase decreases. Also, we can see that the heat transfer mechanism of CHF augmentation is closely connected with the dynamic behavior of bubble generation and departure.

• 한국자동차공학회논문집
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• 제4권3호
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• pp.176-186
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• 1996

The phenomena of drag reduction using small quantities of a liner macromolecules has attracted the attention of many experimental investigations. On the other hand drag reduction in two phase flow can be applied to the transport of crude oil, phase change system such as chemical reactor, pool and boiling flow, and to flow with cavitation which occurs pump impellers. But the research on dragreduction in two phase flow is not sufficient. The purpose of the present work is to evaluate the drag reduction by measuring pressure drop, void fraction, mean liquid velocity and turbulent intensity whether polymer additives a horizontal single and two phase system or not. Flow pattern of air-water two phase flow was classified by electrical conductivity probe signal. Velocities and turbulent intensities of signal were measured simultaneously with a Hot-film anemometer.

• Nuclear Engineering and Technology
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• 제42권6호
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• pp.636-655
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• 2010

A thermal-hydraulic code, named CUPID, has been being developed for the realistic analysis of transient two-phase flows in nuclear reactor components. The CUPID code development was motivated from very practical needs, including the analyses of a downcomer boiling, a two-phase flow mixing in a pool, and a two-phase flow in a direct vessel injection system. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations are solved over unstructured grids with a semi-implicit two-step method. This paper presents an overview of the CUPID code development and assessment strategy. It also presents the code couplings with a system code, MARS, and, a three-dimensional reactor kinetics code, MASTER.

• Nuclear Engineering and Technology
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• 제27권4호
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• pp.544-554
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• 1995

폴리머 수용액의 증기폭발 억제 효과에 대한 물리적 현상을 이해하기 위해 폴리에틸렌옥사이드 수용액에서의 풀비등 특성을 실험적으로 관찰하였다. 본 실험에서는 22.2mm와 9.5mm 직경의 두 구형 체를 가열하여 여러가지 농도의 3$0^{\circ}C$ 수용액에서 냉각시켰다. 그 결과, 순수한 물에서는 7$0^{\circ}C$ 이상인 최소막비등온도($\Delta$ $T_{MFB}$)가 300ppm농도의 폴리머 수용액에서 22.2mm구의 경우 15$0^{\circ}C$ 까지, 9.5mm구의 경우 35$0^{\circ}C$까지 낮아짐을 알 수 있었다. 이러한 폴리머 수용액에서 최소막비등온도가 크게 낮아지는 현상은 이 수용액에서 중기폭발이 억제되는 이유로 해석될 수 있다. 또한, 외부 압력파의 막비등에 대한 영향을 관찰한 결과, 수용액의 농도가 클수록 증기막의 안정도가 커짐을 알 수 있었다. 이러한 폴리머 수용액에서의 비등 특성과 증기폭발 억제에 대한 실험 결과들은 원자로 비상냉각수에 폴리에틸렌옥사이드와 같은 폴리머를 최소 300ppm 정도 소량 첨가하는 방법으로 중대사고시 폭발적 FCI 반응을 방지 또는 완화할 수 있음을 제시한다.다.