• Journal of computational fluids engineering
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• v.12 no.4
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• pp.44-53
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• 2007

A three-dimensional (3D) unstructured hydrodynamic solver for transient two-phase flows has been developed for a 3D component of a nuclear system code and a component-scale analysis tool. A two-fluid three-field model is used for the two-phase flows. The three fields represent a continuous liquid, an entrained liquid, and a vapour field. An unstructured grid is adopted for realistic simulations of the flows in a complicated geometry. The semi-implicit ICE (Implicit Continuous-fluid Eulerian) numerical scheme has been applied to the unstructured non-staggered grid. This paper presents the numerical method and the preliminary results of the calculations. The results show that the modified numerical scheme is robust and predicts the phase change and the flow transitions due to boiling and flashing very well.

• Progress in Superconductivity and Cryogenics
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• v.12 no.1
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• pp.66-70
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• 2010

The operational temperature range of thermosiphon is generally limited from the critical point to the triple point of the working fluid to maintain two-phase state. Thermosiphon with mixed working fluid has a potential to widen the operational temperature range. In this study, the physical behavior of mixed working fluid during the transient operation of thermosiphon was analyzed with temperature-mole fraction diagram. The condenser and the evaporator temperature variations were explained by the dew line and the bubble line of the mixture. It is encouraging that the thermosiphon operation commences early with larger fraction of high boiling point component, but the temperature gap between the condenser and the evaporator due to the separation of two components has a negative effect on the officient cool down process.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.47-58
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• 2020

Most of the reactor trips in Korean NPPs related to core protection systems were caused not because of proximity of boiling crisis and, consequently, a damage in the core, but due to particular miscalculations or component failures related to the core protection system. The most common core protection system applied in Korean NPPs is the Core Protection Calculator System (CPCS), which is installed in OPR1000 and APR1400 plants. It generates a trip signal to scram the reactor in case of low Departure from Nucleate Boiling Ratio (DNBR) or high Local Power Density (LPD). However, is a reactor trip necessary to protect the core? Or could a fast power reduction be enough to recover the DNBR/LPD without a scram? In order to analyze the online calculation of DNBR/LPD, and the use of fast power reduction as trip avoidance methodology, a concept of CPCS with fast power reduction function was developed in Matlab® Simulink using systems engineering approach. The system was validated with maximum of 0.2% deviation from the reference and the dynamic deviation was maximum of 12.65% for DNBR and 6.72% for LPD during a transient of 16,000 seconds.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.31 no.9
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• pp.764-771
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• 2007

The objective of the present work is to evaluate the evaporation heat flux of deposited droplet on the hot surface by using of inverse heat transfer technique. On the basis of measured temperature, a integral form solution is determined for the transient temperatures beyond the two positions by using Green's function technique. This method first approximates the temperature data with a half polynomial series of time. we compared this result with constant radius model in single phase regime, nucleate boiling regime, film boiling regime respectively. this paper performed the experiments as following conditions: (a)the surface temperature is within the range between $80^{\circ}C\;and\;160^{\circ}C$ in the conduction, (b) droplet diameter are 2.4 and 3.0mm. (c) surface roughness is $0.18{\mu}m$.

• Journal of Advanced Marine Engineering and Technology
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• v.28 no.3
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• pp.466-474
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• 2004

The characteristics of liquefied butane spray are expected to be different from conventional diesel fuel spray, because a flash boiling spray is expected when the surround pressure is below the saturated vapor pressure of the butane(0.23MPa 98K). The axial velocities. radial velocities. and size distributions in butane sprays were measured with PDPA(Phase Doppler Particle Analyzer) system. Sprays were macroscopically observed by using the high speed camera in case that the surround pressure is 0.37MPa and 0.15MPa. respectively. Compared with the conventional spray. the reversed results were investigated when the surround pressure is below the saturated vapor pressure of the butane.

• Journal of ILASS-Korea
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• v.5 no.1
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• pp.30-40
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• 2000

The characteristics of liquefied butane spray are expected to be different from conventional diesel fuel spray, because a kind of flash boiling spray is expected when the back pressure is below the saturated vapor pressure of the butane(0.23MPa at 298K). The ambient pressure was held at a pressure above(0.37MPa) and below(0.15MPa) the fuel vapor pressure. The axial velocities, radial velocities, and size distributions in butane sprays were measured with PDPA(Phase Doppler Particle Analyzer) system. The PDPA measurement showed a smaller SMD at the 0.15MPa chamber pressure, compared to the 0.37MPa case. Log-hyperbolic density function for the droplets size distribution can be fitted to the experimental results of a liquefied butane spray.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.831-846
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• 2012

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

• Journal of Advanced Marine Engineering and Technology
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• v.32 no.5
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• pp.707-716
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• 2008

The objective of this study was to analyze the effect of mixed fuel composition and mass fraction on spray inner structure in evaporating transient spray under the variant ambient conditions. Spray structure and spatial distribution of liquid phase concentration were investigated using a thin laser sheet illumination technique on the three component mixed fuels. A pulsed Nd:YAG laser was used as a light source. The experiments were conducted in a constant volume vessel with optical access. Fuel was injected into the vessel with electronically controlled common rail injector. Used fuel contains i-octane($C_8H_{18}$), n-dodecane($C_{12}H_{26}$) and n-hexadecane($C_{16}H_{34}$) that were selected as low-, middle- and high-boiling point fuel, respectively. Experimental conditions are 42 MPa, 72 MPa and 112 MPa in injection pressure, $5\;kg/m^3$, $15kg/m^3$ and $30kg/m^3$ in ambient gas density, 300 K, 500 K, 600 K and 700 K in ambient gas temperature, 300 K and 368 K in fuel temperature and different fuel mass fraction. Experimental results indicated that the multi-component fuels made two phase region mixed vapor and liquid so that it would are helpful to improve combustion, for the fuels of high boiling point component could accelerate evaporation very much according as low boiling point fuel was added to high boiling point fuel.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.252-259
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• 2001

It is derived for the temperature profile in a cylindrical and planar shape capillary pumped loop evaporators subject to a uniform heat flux prior to the initiation of boiling using the finite difference method. The results of the analysis allow for the determination of applied power levels for which nucleation is likely to occur only within the vapor grooves of the evaporator while maintaining subcooling in the liquid core, thereby increasing the likelihood of a successful startup. Also, limits are found for which additional increases in the applied heat flux do not increase the temperature difference between the vapor grooves and the wick-liquid core interface. Several advantages of larger diameter evaporators observed experimentally in startup are explained and quantified by the model. This analysis is appropriate for standard capillary pumped loop evaporators during a fully-flooded startup as well as starter pump designs and loop heat pipes.

• 한국전산유체공학회:학술대회논문집
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• 2007.10a
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• pp.243-248
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• 2007

A three-dimensional (3D) unstructured hydrodynamic solver for transient two-phase flows has been developed. A two-fluid three-field model was adopted for the two-phase flows. The three fields represent a continuous liquid, an entrained liquid, and a vapour field. The hydrodynamic solver is for the 3D component of a nuclear system code and the component-scale analysis tools for transient two-phase flows. The finite volume method and unstructured grid are adopted, which are useful for the flows in a complicated geometry. The semi-implicit ICE (Implicit Continuous-fluid Eulerian) numerical scheme has been adapted to the unstructured non-staggered grid. This paper presents the numerical method and the preliminary results of the calculations. The results show that the numerical scheme is robust and predicts the phase change and the flow transitions due to boiling and flashing problems well.