• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.160-160
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• 1999

• Nuclear Engineering and Technology
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• v.40 no.2
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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.

• 한국전산유체공학회:학술대회논문집
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• 2003.10a
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• pp.281-282
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• 2003

Numerical investigations are carried out to study the mass flux and temperature distribution in a calandria using a 3-D RANS code. The computations made for simulations of flow and convective heat transfer with near-to working conditions. The work provides an estimate of the safe working limits of the heat dissipation by virtue of prediction of the 'hot spots' in the calandria. The work assumes significance for preliminary designs of the reactors and for detailed critical parametric analysis that would be otherwise more expensive.

• Fire Science and Engineering
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• v.21 no.4
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• pp.25-31
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• 2007

The combustion characteristics and combustion heat of sandwich panel cores were analysed using variable external irradiation level. The characteristics such as ignition time, critical heat flux, ignition temperature and surface temperature profile were measured. Fuel samples were exposed to incident heat fluxes from 15 to $50\;kW/m^2$. For the measurement of various combustion characteristics, the size of specimen was $100\;mm\;{\times}\;100\;mm\;{\times}\;50\;mm$ and the samples were 3 different kinds. As results, Type B showed the best characteristics in measurement of combustion heat and ignition temperature and Type C showd the best characteristics in critical heat flux and surface temperature profile than that of the other two. In conclusion, we knew that Type C had the best performance in fire safety from all data of this study.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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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.

• Fire Science and Engineering
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• v.26 no.4
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• pp.18-23
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• 2012

After produce the extracted wood to silicic acid gel, sodium carbonate and silicon dioxide with application of the making method for carbon dioxide, ignition time, ignition temperature, mass loss rate and critical heat flux are measured according to external radiation source (20, 25, 35 and 50 $kW/m^2$). From the results, pressure impregnation wood to use with water glass and carbon dioxide has fire retardant performance at heat flux (less than 20 $kW/m^2$) of Pre-Flashover fires. If we find out the excellent maxing ratio through continuously study, it might be decided to be able to be utilized as fire-retardant wood.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1914-1928
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• 2022

To improve the heat transfer efficiency of the reactor fuel assembly, it is necessary to accurately calculate the two-phase flow boiling characteristics and the critical heat flux (CHF) in the fuel assembly. In this paper, a Eulerian two-fluid model combined with the extended wall boiling model was used to numerically simulate the 5 × 5 fuel rod bundle with spacer grids (four sets of mixing vane grids and four sets of simple support grids without mixing vanes). We calculated and analyzed 11 experimental conditions under different pressure, inlet temperature, and mass flux. After comparing the CHF and the location of departure from the nucleate boiling obtained by the numerical simulation with the experimental results, we confirmed the reliability of computational fluid dynamic analysis for the prediction of the CHF of the rod bundle and the boiling characteristics of the two-phase flow. Subsequently, we analyzed the influence of the spacer grid and mixing vanes on the void fraction, liquid temperature, and secondary flow distribution. The research in this article provides theoretical support for the design of fuel assemblies.

• Nuclear Engineering and Technology
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• v.29 no.1
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• pp.35-44
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• 1997

Critical too-Phase flow rates of subcooled water through Short Pipes (L 140039n) with small diameters (D$\leq$7.15 min) have been experimentally investigated for wide ranges of subcooling (0~199$^{\circ}C$) and pressure (0.5~2.0 MPa). To examine the effects of various parameters (i.e., the location of flashing inception, the degree of subcooling, the stagnation temperature and pressure, and the pipe size) on the critical two-phase flow rates of subcooled water through short pipes with small diameters, a total of 135 runs were made for various combinations of test parameters using four different L/D test sections. Experimental results that show effect of various parameters on subcooled critical two phase flow rates are presented in the form of graphs such as the dimensionless mass flux ( $G^{*}$) versus the dimensionless subcooling ( $T_{sub}$$^{*}$) curve. An empirical correlation expressed in terms of a dimensionless subcooling is also obtained for subcooled two-phase flow rates through present test sections. Comparisons between the mass fluxes calculated by present correlation and a total of 755 selected experimental data points of 9 different investigators show that the agreement is fairly good except for very low subcooling data obtained from small L/D (less than 10) orifices.s.s.s.

• Journal of the Korean Society for Precision Engineering
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• v.26 no.3
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• pp.40-46
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• 2009

Toroidally-wound brushless direct-current (BLOC) machines are compact, highly efficient, and can work across a large magnetic gap. For these reasons, they have been used in pumps, flywheel energy storage systems and left ventricular assist devices among others. The common feature of these systems is a spinning rotor supported by a set of (either mechanical or magnetic) bearings. From the view point of dynamics, it is desirable to increase the first critical speed of the rotor so that it can run at a higher operating speed. The first critical speed of the rotor is determined by the radial stiffnesses of the bearings and the rotor mass. The motor also affects the first critical speed if the rotor is displaced from the rotating center. In this paper, we analytically derive the flux density distribution in a toroidally-wound BLOC machine and also derive the negative stiffness of the motor, based on the assumption that the rotor displacement perturbs the flux density distribution linearly. The estimated negative stiffness is validated by finite element analyses.