• 반도체디스플레이기술학회지
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• 제21권4호
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• pp.18-22
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• 2022

Photovoltaics (PV) power generation efficiency is affected by meteorological factors such as temperature and wind speed. In general, it is known that the power generation amount decreases because photovoltaics panel temperature rises and the power generation efficiency decreases in summer. Photovoltaics Thermal (PVT) power generation has the ad-vantage of being able to produce heat together with power, as well as preventing the reduction in power generation efficien-cy and output due to the temperature rise of the panel. In this study, the amount of heat collected by season and time was calculated for photovoltaics thermal modules using the International Weather for Energy Calculations (IWEC) data provided by the American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE). Based on this, we propose a method of predicting the temperature of the photovoltaics panel using thermal analysis and then calculating the flow rate of coolant to improve power generation efficiency. As the results, the photovoltaics efficiencies versus time on January, April, July, and October in Jeju of the Republic of Korea were calculated to the range of 15.06% to 17.83%, and the maxi-mum cooling load and flow rate for the photovoltaics thermal module were calculated to 121.16 W and 45 cc/min, respec-tively. Though this study, it could be concluded that the photovoltaics thermal system can be composed of up to 53 modules with targeting the Jeju, since the maximum capacity of the coolant circulation pump of the photovoltaics thermal system applied in this study is 2,400 cc/min.

• 한국전산유체공학회지
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• 제8권3호
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• pp.32-40
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• 2003

CFD analyses of the three-dimensional turbulent flow in the impeller and diffuser of an axial flow pump including suction and discharge parts are presented and compared with experimental data. The purpose of the current study is to validate the CFD method for the performance analysis of the main coolant pump for SMART and to investigate the effect of suction and discharge shapes on the pump performance. To generate a performance curve, not only the design point but also the off-design points were computed. The results were compared with available experimental data in terms of head generated. At the design point, the analysis accurately predicts the experimental head value. In the range of the higher flow rates, the results are also in very good agreement with the experimental data, in magnitude but also in terms of slope of variation. For lower flow rates, the results shows that the analysis considering the suction and discharge well describe the typical S-shape performance curve of the axial pump.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.271-276
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• 1996

Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in LMR core steady state thermal-hydraulic performance analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each pin bundle, thus pin cladding damage accrual and pin reliability. The flow orificing analysis for conceptual design will be performed with Excel spreadsheet program ORFCE which was set up and tested, using the calibration factors based on available analyses data. For the verification of this program, flow orificing calculation for the MDP 840MWth core was performed. The calculational results are satisfactory compared to those of CRIEPI calculation.

• 한국전산유체공학회지
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• 제16권3호
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• pp.95-103
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• 2011

In the block type VHTR core, there are inevitable gaps among core blocks for the installation and refueling of the fuel blocks. These gaps are called bypass gap and the bypass flow is defined as a coolant flows through the bypass gap. Distribution of core bypass flow varies according to the reactor operation since the graphite core blocks are deformed by the fast neutron irradiation and thermal expansion. Furthermore, the cross-flow through an interfacial gap between the stacked blocks causes flow mixing between the coolant holes and bypass gap, so that complicated flow distribution occurs in the core. Since the bypass flow affects core thermal margin and reactor efficiency, accurate prediction and evaluation of the core bypass flow are very important. In this regard, experimental and computational studies were carried out to evaluate the core bypass flow distribution. A multi-block experimental apparatus was constructed to measure flow and pressure distribution. Multi-block effect such as cross flow phenomenon was investigated in the experiment. The experimental data were used to validate a CFD model foranalysis of bypass flow characteristics in detail.

• 한국자동차공학회논문집
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• 제16권1호
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• pp.64-70
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• 2008

Temperatures of engine head and liner depend on many factors such as spray and combustion process, coolant passage flow and engine related structures. To estimate the temperature distribution of engine structure, multi-dimensional computational fluid dynamics (CFD) codes have been mainly adopted. In this case, it is of great importance to obtain the realistic wall temperature distribution of entire engine structure. In the present work, a CFD-FEM coupling methodology was presented to address this demand. This approach was applied to a real large-size marine diesel engine. CFD combustion and coolant flow simulations were coupled to FEM temperature analysis. Wall heat flux and wall temperature data were interfaced between combustion simulation and solid component temperature analysis via translator by a commercial CFD package named FIRE by AVL. Heat transfer coefficient and surface temperature data were exchanged and mapped between coolant flow simulation and FEM temperature analysis. Results indicate that there exists the optimum cell thickness near combustion chamber wall to reasonably predict the wall heat flux during combustion period. The present study also shows that the effect of cell refining on predicting in-cylinder pressure during combustion is negligible. Hence, the basic guidance on obtaining the wall heat flux needed for the reasonable CFD-FEM coupling analysis has been established. It is expected that this coupling methodology is a robust tool for practical engine design and can be applied to further assessment of the temperature distribution of other engine components.

• 한국가시화정보학회지
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• 제18권3호
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• pp.23-34
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• 2020

Two-phase flow and heat transfer characteristics during the reflood phase of a single heated rod in the KHU reflood experimental facility were examined. Two-phase flow behavior during the reflooding experiment was carefully visualized along with transient temperature measurement at a point inside the heated rod. By numerically solving one-dimensional inverse heat conduction equation using the measured temperature data, time-resolved wall heat flux and temperature histories at the interface of the heated rod and coolant were obtained. Once water coolant was injected into the test section from the bottom to reflood the heated rod of >700℃, vast vapor bubbles and droplets were generated near the reflood front and dispersed flow film boiling consisted of continuous vapor flow and tiny liquid droplets appeared in the upper part. Following the dispersed flow film boiling, inverted annular/slug/churn flow film boiling regimes were sequentially observed and the wall temperature gradually decreased. When so-called minimum film boiling temperature reached, the stable vapor film between the heated rod and coolant was suddenly collapsed, resulting in the quenching transition from film boiling into nucleate boiling. The moving speed of the quench front measured in the present study showed a good agreement with prediction by a correlation in literature. The obtained results revealed that typical two-phase flow and heat transfer behaviors during the reflood phase of overheated fuel rods in light water nuclear reactors are well reproduced in the KHU facility. Thus, the verified reflood experimental facility can be used to explore the effects of other affecting parameters, such as CRUD, on the reflood heat transfer behaviors in practical nuclear reactors.

• 대한기계학회논문집A
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• 제36권12호
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• pp.1653-1661
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• 2012

본 논문에서는 냉각유동에 의한 SMART 핵연료집합체의 압력강하변화 및 구조특성을 연구하였다. 난류 모델인 BSL 레이놀즈 응력 모델로서 냉각수의 유동을 모델링하여 유체고체연계 해석을 수행하였다. 우선, 지지격자체에 지지된 핵연료봉의 진동해석을 수행하여 실험 결과와 비교하였는데 실험에서의 고유진동수는 48 Hz 로서 시뮬레이션 값과 2% 의 오차를 발생하였다. 핵연료집합체의 압력강하는 한국원자력연구원에서 수행한 실험적 값과 비교하여 8%의 오차가 발생하였고 해석의 타당성을 증명하였다. 유체해석에서는 집합체를 통과하는 각 구간의 유체 속도와 이차유동에 의한 와류생성과정을 관찰하였다. 마지막으로 진동해석과 유체해석의 연계를 통하여 유체유발진동에 의한 연료봉의 변위 값을 관찰하고 최대 변위가 발생하는 곳의 변위 PSD 를 계산하였다.

• Nuclear Engineering and Technology
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• 제18권4호
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• pp.273-280
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• 1986

이 연구의 목적은 가혹한 사고후 손상된 원자로심을 모의한 열을 발생하는 데브리층에서의 강제대류 드라이아웃 열유속을 실험적으로 얻고져 한 것이다. 이 연구에서 냉각재 순환장치를 사용하여 대기압하에서 냉각재가 상향 강제대류하는 유도 가열된 강구 입자층에서의 드라이아웃 열유속을 얻었다. 이 실험에서는 주로 강제대류 드라이아웃에 대한 질량유속.입자크기.입자층의 높이 및 냉각재의 서브클링의 영향이 관찰되었다. 실험은 입자직경이 1.5, 2.5, 3.0 및 4.0mm의 탄소강입자를 55mm 내경의 Pyrex 유리용기에 넣어 고주파유도 전류를 통해 가열하여 이루어졌다. 냉각재로서 증류수를 질량유속 0~3.5kg/$m^2$s로 변화시키어 공금하고 층의 높이는 55mm와 110mm, 냉각재 유입온도는 2$0^{\circ}C$와 8$0^{\circ}C$로 변화시켰다. 주요 실험결과는 다음과 같다. (1) 드라이아웃 열유속은 상향 강제 대류 질량유속과 입자크기내의 증가에 따라 증가한다. (2) 질량유속이 없는 경우 드라이아웃 열유속은 기존 연구결과와 같이 입자직경에 의존한다. (3) 얕은 입자층에서의 드라이아웃 열유속은 깊은 입자층의 것보다 얼마간 높다.

• Journal of Advanced Marine Engineering and Technology
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• 제17권3호
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• pp.24-32
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• 1993

Experimental data on the heat transfer characteristics of HCFC-123 and CFC-11 during condensation in horizontal smooth tube are presented. The experimental apparatus consisted of a closed working fluid loop, coolant loop, and measuring system. The major components of the working fluid loop made of a refrigerant pump, boiler, superheater, refrigerant flow meter, receiver and test section. The tube-in-tube type test section was made of smooth tube which were constructed form 9.52 mm outer diameter of smooth copper tube with 50 mm outside diameter of PVC tube duct. The ranges of parameter, such as refrigerant mass velocity, coolant flow rate, and quality were 90-325kg/($m^2$.s), 60-360kg/h, 5-95% respectively. Data were obtained under steady state condition for annular flow. As a result of these, the condensation heat transfer coefficients for HCFC-123 were slightly lower than those of CFC-11 from 8% to 15% inside horizontal smooth tube. Furthermore, a new generalized correlation for the heat transfer coefficients of HCFC-123 and CFC-11 during condensation inside horizontal smooth tube is proposed.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.104-104
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• 1999

The local heat transfer coefficient is experimentally investigated for the reflux condensation in a countercurrent flow between the steam-air mixture and the condensate. A single vertical tube has a geometry which is a length of 2.4m, inner diameter of 16.56mm and outer diameter of 19.05mm and is made of stainless steel. Air is used as a noncondensible gas. The secondary side is installed in the form of coolant block around vertical tube and the heat by primary condensation is transferred to the coolant water. The local temperatures are measured at 15 locations in the vertical direction and each location has 3 measurement points in the radial direction, which are installed at the tube center, at the outer wall and at the coolant side. In three different pressures, the 27 sets of data are obtained in the range of inlet steam flow rate 1.348 -3.282kg/hr, of inlet air mass fraction 11.8 -55.0%. The local heat transfer coefficient increases as the increase of inlet steam flow rate and decreases as the decrease of inlet air mass fraction. As an increase of the system pressure, the active condensing region is contracted and the heat transfer capability in this region is magnified. The empirical correlation is developed represented with the 165 sets of local heat transfer data. As a result, the Jacob number and film Reynolds number are dominant parameters to govern the local heat transfer coefficient. The rms error is 17. 7% between the results by the experiment and by the correlation.