• 센서학회지
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• 제13권2호
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• pp.139-143
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• 2004

This paper presents a method that the engine oil condition is detected using a natural convection heat transfer in a engine oil. A sensor circuit maintains a constant temperature difference between a heat plate and engine oil for detecting a natural convection heat transfer rate on the constant temperature. The natural convection heat transfer rate is measured by a current through the heat plate of the sensor circuit. The sensor is tested by a fresh oil. 6,000 km and 10,000 km driven oil in the oil temperature range from $20^{\circ}C$ to $100^{\circ}C$. In the experimental result, when the current through the heat plate is altered by variation of a engine oil temperature and flows driven oil more than fresh oil, the sensor could inform a engine oil deterioration to a car driver.

• Nuclear Engineering and Technology
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• 제40권2호
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• pp.155-162
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• 2008

Heat transfer experiments in an annulus passage were performed using SPHINX(Supercritical Pressure Heat Transfer Investigation for NeXt Generation), which was constructed at KAERI(Korea Atomic Energy Research Institute), to investigate the heat transfer behaviors of supercritical $CO_{2}$. $CO_{2}$ was selected as the working fluid to utilize its low critical pressure and temperature when compared with water. The mass flux was in the range of 400 to 1200 $kg/m^{2}s$ and the heat flux was chosen at rates up to 150 $kW/m^{2}$. The selected pressures were 7.75 and 8.12 MPa. At lower mass fluxes, heat transfer deterioration occurs if the heat flux increases beyond a certain value. Comparison with the tube test results showed that the degree of heat transfer deterioration in the heat flux was smaller than that in the tube. In addition, the Nusselt number correlation for a normal heat transfer mode is presented.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3395-3400
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• 2007

Heat transfer experiments at a vertical annulus passage were carried out in the SPHINX(Supercritical Pressure Heat Transfer Investigation for NeXt Generation) to investigate the heat transfer behaviors of supercritical $CO_2$. The collected test data are to be used for the reactor core design of the SCWR (SuperCritical Water-cooled Reactor). The mass flux was in the range of 400${\sim}$1200 kg/$m^2$s and the heat flux was chosen up to 150 kW/$m^2$. The selected pressures were 7.75 and 8.12 MPa. The heat transfer data were analyzed and compared with the previous tube test data. The test results showed that the heat transfer characteristics were similar to those of the tube in case of a normal heat transfer mode and degree of heat transfer deterioration became smaller than that in the tube. Comparison of the experimental heat transfer coefficients with the predicted ones by the existing correlations showed that there was not a distinct difference between the correlations.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.103-112
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• 2017

Application of the supercritical condition in reactor core cooling needs to be properly justified based on the extreme level of parameters involved. Therefore, a numerical study is presented to compare the thermalhydraulic performance of supercritical and single-phase natural circulation loops under low-to-intermediate power levels. Carbon dioxide and water are selected as respective working fluids, operating under an identical set of conditions. Accordingly, a three-dimensional computational model was developed, and solved with an appropriate turbulence model and equations of state. Large asymmetry in velocity and temperature profiles was observed in a single cross section due to local buoyancy effect, which is more prominent for supercritical fluids. Mass flow rate in a supercritical loop increases with power until a maximum is reached, which subsequently corresponds to a rapid deterioration in heat transfer coefficient. That can be identified as the limit of operation for such loops to avoid a high temperature, and therefore, the use of a supercritical loop is suggested only until the appearance of such maxima. Flow-induced heat transfer deterioration can be delayed by increasing system pressure or lowering sink temperature. Bulk temperature level throughout the loop with water as working fluid is higher than supercritical carbon dioxide. This is until the heat transfer deterioration, and hence the use of a single-phase loop is prescribed beyond that limit.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1100-1108
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• 2021

Supercritical water has great potential as a coolant for nuclear reactor. Its use will lead to higher thermal efficiency of Rankine cycle. However, in certain conditions heat transfer may get deteriorated which may lead to undesirable high clad surface temperature. It is necessary to estimate the operating conditions in which heat transfer deterioration (HTD) will take place, so as to establish thermal margins for safe reactor operation. In the present work, the heat flux corresponding to onset of HTD for vertically upward flow of supercritical water in a pipe is obtained over a wide range of system parameters, namely pressure, mass flux, and pipe diameter. This is done by performing large number of simulations using an in-house CFD code, which is especially developed and validated for this purpose. The identification of HTD is based on observance of one or more peak/s in the computed wall temperature profile. The existing correlations for predicting the onset of HTD are compared against the results obtained by present simulations as well as available sets of experimental data. It is found that the prediction accuracy of the correlation proposed by Dongliang et al. is best among the existing correlations.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.292-295
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• 2008

In a SCWR (SuperCritical pressure Water cooled Reactor), the coolant temperature initially at below the pseudo-critical temperature at the bottom of a reactor core increases as the coolant flows upward through the sub-channels of the fuel assemblies, and it finally becomes higher than the pseudo-critical temperature when it leaves the reactor core. At certain conditions, heat transfer deterioration occurs near the pseudo-critical temperature and it may cause a drastic rise of the fuel surface temperature resulting a fuel failure. Therefore, an accurate estimation of the heat transfer coefficient is very important for the thermal-hydraulic design of a reactor core. An experiment on heat transfer to the vertically upward flowing $CO_2$ at a supercritical pressure in a circular tube were performed at KAERI. The internal diameter of the test section is 6.32 mm, which corresponds to the hydraulic diameter of a sub-channel in the conceptional design proposed by KAERI. The test range of the mass flux is 285 to 1200 kg/m$^2$s and the maximum heat flux is 170 kW/m$^2$. The inlet pressure is maintained at 8.12 MPa, which is 1.1 times the critical pressure. A new correlation, which covers both the normal and deterioration heat transfer regimes was proposed and compared with the estimations by exiting correlations.

• 대한기계학회논문집B
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• 제28권11호
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• pp.1302-1314
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• 2004

Turbulent heat transfer to $CO_2$ at supercritical pressure flowing in vertical tubes is investigated using direct numerical simulation (DNS). A conservative space-time discretization scheme for variable-density flows at low Mach numbers is adopted in the present study to treat steep variations of fluid properties at supercritical pressure just above the thermodynamic critical point. The fluid properties at these conditions are obtained using PROPATH and used in the form of tables in the simulations. The buoyancy influence induced by strong variation of density across the pseudo-critical temperature proved to play a major role in turbulent heat transfer at supercritical state. Depending on the degree of buoyancy influence, turbulent heat transfer may be enhanced or significantly deteriorated, resulting in local hot spots along the heated surface. Based on the results of the present DNS combined with theoretical considerations, the physical mechanism of this local heat transfer deterioration is elucidated.

• 대한기계학회논문집B
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• 제35권1호
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• pp.23-31
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• 2011

원자로의 안전을 확보하고 운전성능을 정확히 예측하기 위해서는 핵연료에서 초임계압 냉각재로 전달되는 열전달량을 예측하는 것이 매우 중요하다. 초임계압 유체에서 열전달은 유사임계온도 부근에서 급격하게 변화하는 물성의 영향을 크게 받아 열전달량이 증가하기도 하고, 특정한 질량유속과 열유속 조건에서는 열전달량이 감소하는 현상을 보이기도 한다. 본 연구에서는 초임계압수냉각로(SCWR)의 핵연료집합체 부수로의 수력 직경과 유사한 내경 4.57 mm 원형관 안에서 상 하향 방향으로 흐르는 $CO_2$로 전달되는 열전달량(실제로는 원형관 외벽온도)을 다양한 조건에서 측정하고 분석하였다. 총 7,250개의 실험결과를 기존의 초임계압 강제대류 상관식들로 예측한 결과와 비교하여 그 상관식들의 정확성을 검토하였으며, 부력매개변수 $\overline{Gr}_b/(Re^{2.7}_b\overline{Pr}^{0.5}_b})$를 도입하여 혼합대류 영역에서도 적용할 수 있도록 기존의 상관식을 확장하였다. 그리고 기존의 열전달열화 발생 조건을 본 실험의 열전달열화 조건과 비교하여 적용성을 평가하였다.

• 대한기계학회논문집B
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• 제33권4호
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• pp.235-241
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• 2009

Experimental studies on single-phase toroidal circulation loop(thermosyphon) have been performed in the present study with Ag-nanofluids as a working fluids. The present paper deals with an experimental study on the heat transfer behavior of single-phase toroidal loop. Toroidal loop charged with nanofluid has been constructed and a number of tests have been carried out. Different geometric parameter, e.g., orientation has been investigated. The tests were conducted employing two fluids: distilled water and Ag-nanofluid of various volume concentrations. The experiments at Rayleigh number from $10^5$ to $10^6$ showed a systematic and slight deterioration in natural convective heat transfer. It was observed that the deterioration due to the particle concentration was in the range of 5-10%. At a given particle concentration of 0.05%, abrupt decrease in the Nusselt number and the Raleigh number was observed. The present study with toroidal loop shows that the application of nanofluids for heat transfer intensification should not be decided only by the effective thermal conductivity with increasing particle concentration.

• 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.