• 대한기계학회논문집B
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• 제29권2호
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• pp.239-246
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• 2005

Effects of subcooling on pool boiling heat transfer in vertical annuli with closed bottoms have been investigated experimentally. For the test, a tube of 19.1mm diameter and the water at atmospheric pressure have been used. Three annular gaps of 7.05, 18.15, and 28.20 have been tested in the subcooled water and results of the annuli are compared with the data of a single unrestricted tube. The increase in pool subcooling results in much change in heat transfer coefficients. At highly subcooled regions, heat transfer coefficients for the annuli are much larger than those of a single tube. As the heat flux increases and subcooling decrease, a deterioration of heat transfer coefficients is observed at the annulus of 7.05mm gap. Single-phase natural convection and liquid agitation are the governing mechanisms for the single tube while liquid agitation and bubble coalescence are the major factors at the bottom closed annuli.

• Nuclear Engineering and Technology
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• 제32권1호
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• pp.17-33
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• 2000

A mechanistic model based on wall-attached bubble coalescence, previously developed by the authors, was extended to predict a vow high critical heat flux (CHF）in highly subcooled flow boiling, especially for high mass flux and small tube diameter conditions. In order to take into account the enhanced condensation due to high subcooling and high mass velocity in small diameter tubes, a mechanistic approach was adopted to evaluate the non-equilibrium flow quality and void fraction in the subcooled water flow boiling, with preserving the structure of the previous CHF model. Comparison of the model predictions against highly subcooled water CHF data showed relatively good agreement over a wide range of parameters. The significance of the proposed CHF model lies in its generality in applying over the entire subcooled flow boiling regime including the operating conditions of fission and fusion reactors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.449-456
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• 1998

A theoretical critical heat flux (CLE) model. based on lateral bubble coalescence on the heated wall, is proposed to predict the subcooled flow boiling CHF in a uniformly heated vertical tube. The model is based on the concept that a single layer of bubbles contacted to the heated wall events a bulk liquid from reaching the wall at near CHF condition. Comparisons between the model predictions and experimental data result in satisfactory agreement within less than 9.73 % root-mean-square error by the appropriate choice of the critical void fraction in the bubbly layer. The present model shows comparable performance with the CHF look-up table of Groeneveld et al.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.552-557
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• 1997

Based on the bubble coalescence adjacent to the heated wall as a flow structure for CHF condition, Chang and Lee developed a mechanistic critical heat flux (CHF) model for subcooled flow boiling. In this paper, improvements of Chang-Lee model are implemented with more solid theoretical bases for subcooled and low-quality flow boiling in tubes. Nedderman-Shearer's equations for the skin friction factor and universal velocity profile models are employed. Slip effect of movable bubbly layer is implemented to improve the predictability of low mass flow. Also, mechanistic subcooled flow boiling model is used to predict the flow quality and void fraction. The performance of the present model is verified using the KAIST CHF database of water in uniformly heated tubes. It is found that the present model can give a satisfactory agreement with experimental data within less than 9% RMS error.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.911-920
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• 2013

Experimental studies on both subcooled and saturated pool boiling of water were performed to obtain local heat transfer coefficients on a $3^{\circ}$ inclined tube of 50.8 mm diameter at atmospheric pressure. The local values were determined at every $45^{\circ}$ from the very bottom to the uppermost of the tube periphery. The maximum and minimum local coefficients were observed at the azimuthal angles of $0^{\circ}$ and $180^{\circ}$, respectively, in saturated water. The locations of the maxima and the minima were dependent on the inclination angle of the tube as well as the degree of subcooling. The major heat transfer mechanisms were considered to be liquid agitation generated by the sliding bubbles and the creation of big size bubbles through bubble coalescence. As a way of quantifying the heat transfer coefficients, an empirical correlation was suggested.

• 대한기계학회논문집
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• 제19권9호
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• pp.2316-2327
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• 1995

Experimental results from nucleate pool boiling heat transfer with various finned tubes in CFC11, HCF123 and HCFC141b are reported. One plain tube and four low fin tubes of various fin densities were tested in an attempt to find out the optimum fin density in the heat flux range of 10-60 kW/m$^{[-992]}$ at near atmospheric pressure. The results indicated that CFC11 showed the highest heat transfer coefficients. Its alternatives, HCFC123 and HCFC141b, showed 3-5% lower heat transfer coefficients than those of CFC11 at the same heat flux. As the fin density increases, so does the heat transfer surface area. Measured heat transfer coefficients, however, do not necessarily always increase as the fin density increases. This unique phenomenon seems to be caused by the coalescence of the bubblers that prevent the cool liquid from entering into the fin valleys. For all the refrigerants tested, the optimum fin density yielding the highest performance was 28 fins per inch confirming the previous results by other researchers.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2771-2782
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• 2022

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4-2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as interrelated phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS.

• 대한기계학회논문집B
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• 제35권3호
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• pp.237-244
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• 2011

경사각이 대기압하의 포화상태인 물의 풀비등에 미치는 영향을 조사하기 위해 실험을 통한 연구를 수행하였다. 연구를 위하여 5mm의 틈새간격을 가지는 하부유로개폐 상태인 환상공간을 고려하였다. 환상공간의 내부에 설치된 튜브를 가열하였으며 튜브의 직경과 길이는 각각 25.4mm와 500mm이다. 경사각은 수평부터 수직까지 변경하였다. 본 실험의 결과를 틈새간격이 더 큰 환상공간 및 단일튜브에 대한 결과와 서로 비교하였다. 작은 틈새간격을 가지는 환상공간의 경우 경사각이 열전달에 미치는 영향은 그다지 크지 않음을 확인하였다. 그러나 환상공간이 수평상태인 경우 80kW/$m^2$에서 임계열유속이 관찰되었다. 액체 교란의 정도와 기포군집형성이 환상공간 내부 풀비등의 주된 열전달 기구로 이해된다.

• 대한기계학회논문집B
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• 제38권1호
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• pp.81-88
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• 2014

수평에 가깝게 설치된 튜브 원주면에 대해 평균 열전달계수를 결정하기 위한 실험적 연구를 수행하였다. 실험을 위하여 대기압 상태하의 물속에 잠긴 50.8 mm의 스테인리스강 튜브를 사용하였다. 과냉 및 포화 풀비등 조건을 모두 고려하였으며, 튜브 경사각은 수평으로부터 $9^{\circ}$까지 $3^{\circ}$ 간격으로 변경하였다. 포화상태에서는 튜브의 최하부로부터의 방위각이 $90^{\circ}$인 위치에서 측정한 국소비등열전달계수가 평균값으로 취급될 수 있으며, 이러한 경향은 튜브 경사각과는 무관함을 확인하였다. 그러나 물이 과냉상태인 경우, 평균 열전달계수의 위치는 경사각과 열유속에 의존한다. 열전달을 변화시키는 주된 열전달 기구는 액체교란 강도 및 기포군집에 의한 큰 기포 덩어리의 형성과 밀접한 관계가 있는 것으로 설명된다.

• Nuclear Engineering and Technology
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• 제28권1호
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• pp.1-16
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• 1996

축소한 격납용기 내부 핵연료재장전수저장탱크의 안쪽에 설치한 열교환기 튜브의 주요 매개변수들이 풀핵비등 열전달에 미치는 복합적인 영향을 극명하기 위해 튜브 외경, 표면 거칠기, 그리고 튜브 설치 방향에 대한 다양한 조합들을 환용하여 열유속 q'quot;와 과열 온도 차이 $\Delta$T 간의 관계에 대한 총 1,966 개의 실험값을 취득하였다. 이 실험 결과들에 의하면, (1) 표면 거칠기 증가는 수평 및 수직 튜브 모두에 대해 열전달을 향상시키고, (2) 기포 생성에 따른 두가지 열전달 기구인 주변 액체 운동증가에 의한 열전달 향상과 기포층 및 기포 군집 형성에 의한 열전달 감소는 50㎾/$m^2$의 열유속을 경계로 낮은 열유속과 높은 열유속 영 역 에서 서로 다르게 관찰되는데, 이것은 튜브 설치 방향과 표면 거칠기의 크기와 관련이 있으며, (3) 튜브 외경 증가는 수평 및 수직 튜브 모두에 대해 열전달을 감소시키는데, 그 영향정도는 수평보다 수직구조에서 더 크다. 수평 및 수직 튜브들에 대해 열유속 q'quot;와 표면 거칠기 ($\varepsilon$) 및 튜브 외경 (D) 사이의 관계를 결정하는 두 가지 실험식을 개발하였다. 그리고, q'quot;만의 함수로된 풀핵비등 열전달계수( $h_{b}$ 에 대한 간단한 실험식도 부가적으로 개발하였다. 실험식도 부가적으로 개발하였다.'quot;만의 함수로된 풀핵비등 열전달계수($h_{b}$ 에 대한 간단한 실험식도 부가적으로 개발하였다.