• Nuclear Engineering and Technology
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• 제34권4호
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• pp.269-285
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• 2002

Water heat transfer experiments were carried out in a uniformly heated annulus with a wide range of pressure conditions. The local heat transfer coefficients for saturated water (low boiling have been measured just before the occurrence of the critical heat flux (CHF) along the length of the heated section. The trends of the measured heat transfer coefficients were quite different from the conventional understanding for the heat transfer of saturated flow boiling. This discrepancy was explained from the nucleate boiling in the liquid film of annular flow under high heat flux conditions. The well-known correlations were compared with the measured heat transfer coefficients. The Shah and Kandlikar correlations gave better prediction than the Chen correlation. However, the modified Chen correlation proposed in the present work showed the best agreement with the present data among correlations examined .

• Nuclear Engineering and Technology
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• 제34권2호
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• pp.165-175
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• 2002

Effects of gap sizes (3.9-44.3 mm) of vertical annuli on nucleate pool boiling heat transfer of water at atmospheric pressure have been obtained experimentally. Through the study, tubes of the closed bottom have been investigated and results are compared with those of a single unconfined tube. According to the results, the annular condition gives much increase in heat transfer coefficient at moderate heat fluxes. The increase is much enhanced 3s the gap size decreases. At the same tube wall superheat (about 3.1 K) the heat transfer coefficient for the least gap size (i.e., 3.9 mm) is more than three times greater than that of the unconfined tube. However, deterioration of heat transfer occurs at high heat flux for confined boiling.

• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1205-1214
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• 2009

The critical heat flux (CHF) levels that occurred due to exponential heat inputs for varying periods to a 1.0-mm diameter horizontal cylinder immersed in various liquids were measured to develop an extended database on the effect of high subcoolings for quasi-steady-state and transient maximum heat fluxes. Two main mechanisms of CHF were found. One mechanism is due to the time lag of the hydrodynamic instability (HI) which starts at steady-state CHF upon fully developed nucleate boiling, and the other mechanism is due to the explosive process of heterogeneous spontaneous nucleation (HSN) which occurs at a certain HSN superheat in originally flooded cavities on the cylinder surface. Steady-state CHFs were divided into three regions for lower, intermediate and higher subcooling at pressures resulting from HI, transition and HSN, respectively. HSN consistently occurred in the transient boiling CHF conditions that correspond to a short period. It was also found that the transient boiling CHFs gradually increased, then rapidly decreased and finally increased again as the period became shorter.

• Nuclear Engineering and Technology
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• 제53권9호
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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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• 제13권1호
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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.

• 에너지공학
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• 제24권1호
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• pp.149-163
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• 2015

본 연구에서는 전산유체역학(Computational Fluid Dynamics, CFD) 해석코드를 사용하여 마스트집합체의 열수력적 안전성에 대한 연구를 수행하였다. 이를 위해 자연대류 벤치마크 문제를 선정하여 CFD 코드의 물리모델을 선정 및 해석 능력을 검증하고 이를 이용하여 마스트집합체에 대한 자연대류 열전달 해석을 수행하였다. 본 연구에서는 Betts et al.의 사각 수직공동에서 난류 자연대류 실험결과를 대상으로 CFD 해석을 수행하여 자연대류 조건에 적용하기 위한 난류 모델로 표준 $k-{\omega}$ 모델을 선정하였다. 이렇게 도출된 난류모델을 CFD코드에 적용하여 Bates et al.에 의해 수행된 PNL(Pacific Northwest Laboratory)의 $2{\times}6$ 번들 실험과 이에 대한 Kwon et al.의 MATRA, Fluent 코드의 해석과 비교 계산을 수행하여 CFD코드의 부수로조건 자연대류 열전달 해석 능력을 검증하였다. 최종적으로 도출된 $k-{\omega}$ 난류 모델을 사용하여 마스트집합체 및 핵연료 집합체에 대한 자연대류 해석을 수행하였다. 해석 결과 수조 내부 및 부수로 내에서 안정적인 자연대류 유동이 발생함을 확인하였으며, 본 유동 조건에서 핵비등이탈비를 계산함으로써 열수력적 안전성을 정량적으로 평가하였다.

• 한국산학기술학회논문지
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• 제9권3호
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• pp.574-579
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• 2008

수평관내 $CO_2$의 증발 열전달 계수를 실험적으로 조사하였다. 냉매 순환루프의 주요 구성품은 수액기, 변속냉매 펌프, 질량 유량계, 예열기, 증발기(시험부)로 구성된다. 시험부는 내경 4.57 mm의 수평 평활 스텐레스관이다. 실험은 질량유속 $400{\sim}900kg/m^2s$, 포화온도 $5{\sim}20^{\circ}C$, 열유속 $10{\sim}40kW/m^2$인 조건에서 수행하였다. 실험결과로부터 $CO_2$의 열전달은 대류비등보다는 핵비등에 더 많은 영향을 받는 것을 알 수 있었고, $CO_2$의 질량유속은 핵비등에 많은 영향을 미치지 않는 것으로 나타났다. 실험결과와 종래의 상관식을 비교해 본 결과, 기존의 상관식은 실험데이터를 과소예측하였지만, 정 등의 상관식은 좋은 일치를 보였다. 따라서, 수평관내 $CO_2$의 증발 열전달 계수를 예측할 수 있는 정확한 상관식의 개발이 필요하리라 판단된다.

• 한국생산제조학회지
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• 제8권4호
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• pp.68-78
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• 1999

Heat transfer performance is studied for boiling and condensation of R-11 on integral-fin tubes. Nine tubes with trapezoidal integral-fins having fin densities from 748 to 1654fpm and 10,30 grooves and finned tubes with caves of 0.55 and 0.64 mm height respectively are tested. in case of condensation CFC-11 condensates at saturation stat of 32$^{\circ}C$ on the outside surface cooled by inside cooling water flows. And in case of boiling the refrigerant evaporates at a saturation state of 1 bar on the outside tube surface and heat is supplied by hot water which circulates inside of the tube,. The tube having fin transfer coefficient concerns fin tubes with caves show higher valve than low fin tube having find density of 1299fpm and 30grooves. The overall heat transfer coefficient of fin tube with caves is about 5155 W/mK at 2.8m/s of water velocity, The value is abuot 2.7 times higher than plain tube and 1.3 times higher than low fin tube having fin density of 1299fpm and 30 grooves.

• 한국생산제조학회지
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• 제8권4호
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• pp.65-65
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• 1999

Heat transfer performance is studied for boiling and condensation of R-11 on integral-fin tubes. Nine tubes with trapezoidal integral-fins having fin densities from 748 to 1654fpm and 10,30 grooves and finned tubes with caves of 0.55 and 0.64 mm height respectively are tested. in case of condensation CFC-11 condensates at saturation stat of 32℃ on the outside surface cooled by inside cooling water flows. And in case of boiling the refrigerant evaporates at a saturation state of 1 bar on the outside tube surface and heat is supplied by hot water which circulates inside of the tube,. The tube having fin transfer coefficient concerns fin tubes with caves show higher valve than low fin tube having find density of 1299fpm and 30grooves. The overall heat transfer coefficient of fin tube with caves is about 5155 W/mK at 2.8m/s of water velocity, The value is abuot 2.7 times higher than plain tube and 1.3 times higher than low fin tube having fin density of 1299fpm and 30 grooves.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1388-1395
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• 2017

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal-hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal-hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.