• Nuclear Engineering and Technology
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• 제16권1호
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• pp.1-10
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• 1984

2상 유동에 대한 일반적인 시간과 공간에 대해 평균된 보존방정식과 jump condition을 유도했다. 단상난류 유동 방정식에 사용되는 방법을 써서, 한부분에서 순간적으로 이루어지는 평형 방정식 (local instant balance eq.)으로부터 2상유동내 단상영역에 관한 시간에 대해 평균된 방정식을 유체체적에 대해 적분하여 얻어진 결과는 공간적으로 두차례에 걸쳐 평균된다. 즉, 한 유체체적내에서 일차적으로 k번째 상의 단상영역에 대해 평균하고 다음에 k번째 상 전체체적에 대해 평균한다. 질량, 운동량 그리고 에너지 보존 방정식은 일반적인 시간과 공간에 대해 평균된 방정식으로부터 얻어진다. 이 모델의 장점은 Ishii모델, 그리고 Banerjee의 model과 비교하여 설명된다. 마지막으로, THERMIT-6S의 방정식에 포함된 가정과 근사항들에 대해 박혀둔다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집E
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• pp.107-112
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• 2001

The effect of ambient gas (steam) condensation on swirl spray characteristics were studied experimentally for low subcooling condition of the liquid. The configuration of the liquid(water) sheet and the breakup modes were examined. Also variation of the discharge coefficient, breakup length, local and the cross-sectional area-averaged SMD of droplets with the liquid flow(injection) rate were obtained. The perforation breakup mode appears dominant with condensation while the aerodynamic wave breakup mode is dominant without condensation(in the air environment). The discharge coefficient, breakup length and the mean drop sizes decrease in a same manner with increasing of the liquid flow rate for both cases(with and without condensation). The condensation effects are insignificant with the discharge coefficient. However, the local and cross-sectional area-averaged SMD are larger and the breakup length becomes shorter in the steam environment. The spray angle predicted from the volumetric flux distribution along the radial direction of the sprays in the steam environment becomes larger with condensation.

• Nuclear Engineering and Technology
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• 제34권1호
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• 대한기계학회논문집B
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• 제31권4호
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• pp.349-356
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• 2007

The present study investigated local heat/mass transfer characteristics on the tip of the rotating turbine blade with various incoming flow incidence angles. The experiments are conducted in a low speed annular cascade with a single stage turbine. The blade has a flat tip with a mean tip clearance of 2.5% of the blade chord. The incoming flow Reynolds number is $1.5{\times}10^5$ at design condition. To examine the effect of off-design condition, the experiments with various incidence angles ranging between $-15^{\circ}$ and $+7{\circ}$ were conducted. A naphthalene sublimation method is used to measure detailed mass transfer coefficient on the blade. The results indicated that the incidence angle strongly affects the behavior of tip leakage flow around the blade tip and consequently plays an important role in determining heat transfer characteristics on the tip. For negative incidence angles, the heat/mass transfer in the upstream region on the tip decreases by up to 20%. On the contrary, for positive incidence angles, much higher heat transfer coefficients are observed even with small increase of incidence angle.

• 한국압력기기공학회 논문집
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• 제5권1호
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• pp.9-13
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• 2009

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS-56.8(1994) in Korea. Two methods, the make-up flow rate and the pressure decay, are used for LLRT. Though ANSI/ANS-56.8 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the test period for type "C" LLRT is differently applied to each NPPs. Therefore, this study presents a unified test criteria for data stabilization and test duration through experiments to improve the test reliability for type "C".

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 1998년도 춘계 학술대회논문집
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• pp.43-48
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• 1998

The development of unsteady three-dimensional incompressible viscous solver based on unsteady physical curvilinear coordinate system is presented. A 12-point finite analytic scheme based on local uniform grid spacing is extended for nonuniform grid spacing. The formulation of a condition is suggested to avoid the oscillation of the series summations produced by the application of the method of separation of variables. SIMPLER and pressure Poisson equation techniques are used for solving a velocity-pressure coupled problem. The matrix is solved using the Generalized Minimal RESidual (GMRES) method to enhance the convergence rate of unsteady flow solver and the Kinematic boundary condition of a free surface flow. It is demonstrated that the numerical solutions of these equations are less mesh sensitive.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.721-733
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• 2020

In nuclear engineering, the occurrence of critical heat flux (CHF) is complicated for rod bundle, and it is much more difficult to predict the CHF when it is in natural circulation under motion condition. In this paper, the dryout-type CHF is investigated for the rod bundle in a natural circulation loop under rolling motion condition based on the coupled analysis of subchannel method, a one-dimensional system analysis method and a CHF mechanism model, namely the three-fluid model for annular flow. In order to consider the rolling effect of the natural circulation loop, the subchannel model is connected to the one-dimensional system code at the inlet and outlet of the rod bundle. The subchannel analysis provides the local thermal hydraulic parameters as input for the CHF mechanism model to calculate the occurrence of CHF. The rolling motion is modeled by additional motion forces in the momentum equation. First, the calculation methods of the natural circulation and CHF are validated by a published natural circulation experiment data and a CHF empirical correlation, respectively. Then, the CHF of the rod bundle in a natural circulation loop under both the stationary and rolling motion condition is predicted and analyzed. According to the calculation results, CHF under stationary condition is smaller than that under rolling motion condition. Besides, the CHF decreases with the increase of the rolling period and angular acceleration amplitude within the range of inlet subcooling and mass flux adopted in the current research. This paper can provide useful information for the prediction of CHF in natural circulation under motion condition, which is important for the nuclear reactor design improvement and safety analysis.

• 한국해양공학회지
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• 제10권3호
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• pp.138-152
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• 1996

Many studies of heat transfer on the swirling flow or unswirled flow in a abrupt pipe expansion are widely carried out. The mechanism is not fully found evidently due to the instabilities of flow in a sudden change of the shape and appearance of turbulent shear layers in a recirculation region and secondary vortex near the corner. The purpose of this study is to obtain data through an experimental study of the swirling flow and heat transfer downstream of an abrupt expansion in a circular pipe with uniform heat flux. Experiments were carried out for the turbulent flow nd heat transfer downstream of an abrupt circular pipe expansion. The uniform heat flux condition was imposed to the downstream of the abrupt expansion by using an electrically heated pipe. Experimental data are presented for local heat transfer rates and local axial velocities in the tube downstream of an abrupt 3:1 & 2:1 expansion. Air was used as the working fluid in the upstream tube, the Reynolds number was varied from 60, 00 to 120, 000 and the swirl number range (based on the swirl chamber geometry, i.e. L/d ratio) in which the experiments were conducted were L/d=0, 8 and 16. Axial velocity increased rapidly at r/R=0.35 in the abrupt concentric expansion turbulent flow through the test tube in unswirled flow. It showed that with increasing axial distance the highest axial velocities move toward the tube wall in the case of the swirling flow abrupt expansion. A uniform wall heat flux boundary condition was employed, which resulted in wall-to-bulk temperatures ranging from 24.deg. C to 71.deg. C. In swirling flow, the wall temperature showed a greater increase at L/d=16 than any other L/d. The bulk temperature showed a minimum value at the pipe inlet, it also exhibited a linear increase with axial distance along the pipe. As swirl intensity increased, the location of peak Nu numbers was observed to shift from 4 to 1 step heights downstream of the expansion. This upstream movement of the maximum Nusselt number was accompanied by an increase in its magnitude from 2.2 to 8.8 times larger than fully developed tube flow values.

• 대한기계학회논문집
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• 제19권10호
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• pp.2586-2594
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• 1995

Two phase flow phenomena are observed in many industrial facilities and make much importance of optimum design for nuclear power plant and various heat exchangers. This experimental study has been investigated the classification of the flow pattern, the local void distribution and convective heat transfer in swirl and non-swirl two phase flow under the isothermal and nonisothermal conditions. The convective heat transfer coefficients in the single phase water flow were measured and compared with the calculated results from the Sieder-Tate correlation. These coefficients were used for comparisons with the two-phase heat transfer coefficients in the flow orientations. The experimental results indicate, that the void probe signal and probability density function of void distribution can used into classify the flow patterns, no significant difference in voidage distribution was observed between isothermal and non-isothermal condition in non-swirl flow, the values of two phase heat transfer coefficients increase when superficial air velocities increase, and the enhancement of the values is observed to be most pronounced at the highest superficial water velocity in non-swirl flow. Also two phase heat transfer coefficients in swirl flow are increased when the twist ratios are decreased.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2009년 춘계학술대회논문집
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• pp.87-89
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• 2009

In the present paper, transition turbulence model is applied to the NACA64(3)618 and detailed flow features are studied. The turbulence model is sensitive to the boundary layer grid quality and y+ of the grid was limited to 1. The prediction of the transition region is dependent on the local flow condition. The pressure coefficient distribution of the transition turbulence model is compared with that of the fully turbulent mode and the drag distribution of the transition turbulence model was compared with that of the wind tunnel test.