• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.233-240
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• 1996

Analyses of the viscous and end effects on electromagnetic (EM) pumps of annular linear induction type for the sodium coolant circulation in Liquid Metal Fast Breeder Reactors have been carried out based on the MHD laminar flow analysis and the electromagnetic field theory. A one-dimensional MHD analysis for the liquid metal flowing through an annular channel has been performed on the basis of a simplified model of equivalent current sheets instead of three-phase currents in the discrete primary windings. The calculations show that the developed pressure difference resulted from electromagnetic and viscous forces in the liquid metal is expressed in terms of the slip, and that the viscous loss effects are negligible compared with electromagnetic driving forces except in the low-slip region where the pumps operate with very high flow velocities comparable with the synchronous velocity of the electromagnetic fields, which is not applicable to the practical EM pumps. A two-dimensional electromagnetic field analysis based on an equivalent current sheet model has found the vector potentials in closed form by means of the Fourier transform method. The resultant magnetic fields and driving forces exerted on the liquid metal reveal that the end effects due to finiteness of the pump length are formidable. In addition, a two-dimensional numerical analysis for vector potentials has been performed by the SOR iterative method on a realistic EM pump model with discretely-distributed currents in the primary windings. The numerical computations for the distributions of magnetic fields and developed pressure differences along the pump axial length also show considerable end effects at both inlet and outlet ends, especially at high flow velocities. Calculations of each magnetic force contribution indicate that the end effects are originated from the magnetic force caused by the induced current ( u x B ) generated by the liquid metal movement across the magnetic field rather than the one (E) produced by externally applied magnetic fields by three-phase winding currents. It is concluded that since the influences of the end effects in addition to viscous losses are extensive particularly in high-velocity operations of the EM pumps, it is necessary to find ways to suppress them, such as proper selection of the pump parameters and compensation of the end effects.

• 플랜트 저널
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• 제12권4호
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• pp.37-43
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• 2016

원자력발전소의 사고 대처 부하에 전력을 공급하는 전원계통은 다양한 조건에서도 일정 전압이 유지됨을 분석을 통하여 입증한다. 이를 위하여 발전소를 일정부하 운전 상태로 유지하고, 대용량 전동기(원자로냉각재펌프(RCP), 기기냉각수펌프(CCWP))를 각각 기동하여 기동 전 후 안전관련 모선의 전압을 측정하였다. 현장 시험으로 확보된 자료(예, 전압, 전류, 역율 등)는 기존 전력계통해석 모델의 운전 조건으로 재입력하고 재분석을 수행하였다. 이는, 기존 전력계통분석에 사용된 분석기법과 가정들을 실질적인 측정과 결과 분석으로 입증하는 과정이다. 결국, 두 경우의 전압 강하는 발전소 안전에 중요한 기기의 전압이 허용전압 이하로 저하되지 않음과 두 값의 비교 결과가 요구되는 제한치 이내임을 검증한다.

• Journal of Advanced Marine Engineering and Technology
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• 제28권1호
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• pp.65-74
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• 2004

The heat transfer coefficient and pressure drop of $CO_2$(R-744) during gas cooling Process of carbon dioxide in a horizontal tube were investigated experimentally and theoretically. The experiments were conducted without oil in the refrigerant loop. The main components of the refrigerant loop consist of a receiver. a variable-speed pump. a mass flowmeter, an evaporator. and a gas cooler(test section). The main components of the water loop consist of a variable-speed Pump. an constant temperature bath. and a flowmeter. The gas cooler is a counterflow heat exchanger with refrigerant flowing in the inner tube and water flowing in the annulus The test section consists of smooth, horizontal stainless steel tube of 9.53 mm outer diameter and 7.75 mm inner diameter. The length of test section is 6 m. The refrigerant mass fluxes were 200 ~ 300 kg/($m^2{\cdot}s$) and the inlet pressure of the gas cooler varied from 7.5 MPa to 8.5 MPa. The main results were summarized as follows : The predicted correlation can evaluated the R-744 exit temperature from the gas cooler within ${\pm}10%$ for most of the experimental data, given only the inlet conditions. The predicted gas cooley capacity using log mean temperature difference showed relatively food agreement with gas cooler capacity within ${\pm}5%$. The pressure drop predicted by Blasius estimated the pressure drop on the $CO_2$ side within ${\pm}4.3%$. The predicted heat transfer coefficients using Gnielinski's correlation evaluated the heat transfer coefficients on the $CO_2$ side well within the range of experimental error. The predicted heat transfer coefficients using Gao and Honda's correlation estimated the heat transfer coefficients on the coolant side well within ${\pm}10\;%$. Therefore. The predicted equation's usefulness is demonstrated by analyzing data obtained in experiments.

• 한국해양공학회:학술대회논문집
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• 한국해양공학회 2004년도 학술대회지
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• pp.331-336
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• 2004

Dissimilar friction welding were produced using 10mm and 11mm diameter solid bar in Inconel ally(IN X-750) to Stainless steel(STS316L) to investigate their mechanical properties. The main friction welding parameters were selected to endure good quality welds on the basis of visual examination, tensile tests, Virkers hardness surveys of the bond of area and HAZ and macro-structure investigations. The specimens were tested as welded, not heat-treated. The tensile strength of the friction welded steel bars was increased up to $95\%$ of the STS316L base metal under the condition of all heating time. Optimal welding conditions were n=2,000(rpm), $P_1=220(MPa),\;P_2=260(MPa),\;t_1=4(s),\;t_2=4(s)$ when the total upset length is 7(mm).

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2012년도 춘계학술대회 논문집
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• pp.734-738
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• 2012

A comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals is established in accordance with the United States Nuclear Regulatory Commission Regulatory Guide 1.20 Revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results at the inner barrel assembly top plate in the reactor. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at Inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals.

• 한국소음진동공학회논문집
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• 제22권5호
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• pp.474-479
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• 2012

A comprehensive vibration assessment program for the advanced power reactor 1400(APR1400) reactor vessel internals is established in accordance with the united states nuclear regulatory commission regulatory guide 1.20 revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results of the inner barrel assembly top plate in APR1400. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for APR1400 reactor vessel internals.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.151-156
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• 2001

Experimental frictional and wear characteristics of silicon graphite materials is studied in this paper. Those specimens are lubricated with high temperature and highly pressurized water to simulate the same operating condition for the journal bearing and the thrust bearing on the main coolant pump bearing in the newly developing nuclear reactor named SMART(System-integrated Modular Advanced ReacTor). Operating condition of the bearings is realized by the tribometer and the autoclave. Friction coefficient and wear loss are analyzed to choose the best silicon graphite material. Pin on plate test specimens are used and coned disk springs are used to control the applied force on the specimens. Wear loss ana wear width are measured by a precision balance and a micrometer. The friction force is measured by the strain gauge which can be used under high temperature and high pressure. Three kinds of silicon graphite materials are examined and compared with each other, and each material shows similar but different results on frictional and wear characteristics.

• 한국자동차공학회논문집
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• 제11권4호
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• pp.58-67
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• 2003

An automotive engine cooling system is closely related with overall engine performances, such as reduction of fuel consumption, decrease of air pollution, and increase of engine life. Because of complex reaction between each component, the direct experiment, using a vehicle, takes high cost, long time, and slow response to the system change. Therefore, a computer simulation would provide the designer with an inexpensive and effective tool for design, development, and optimization of the engine cooling system over a wide range of operating conditions. In this work, it has been predicted the thermal performance of the engine cooling system in cases of stationary mode, constant speed mode, and city-drive mode by mathematical modelling of each component and numerical analysis. The components are engine, radiator, heater, thermostat, water pump, and cooling fans. Since the engine model is the most important, that is divided into eight sub-sections. The volume mean temperature of eight sub-sections are simultaneously calculated at a time. For detail calculation, the radiator and heater are also divided into many sub-sections like control volumes in finite difference method. Each sub-section is assumed to consist of three parts, coolant, tube with fin, and air. Hence it has been developed the simulation program that can be used in case of design and system configuration changes. The overall performance results obtained by the program were desirable and the time-traced tendencies of the results agreed fairly well with those of actual situations.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.382-395
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• 2002

An experimental study on transient critical heat flux (CHF) under flow coastdown has been performed for the water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady-state CHF The transient CHF experiments have been performed for three kinds of flow transient modes based on the coastdown data of a nuclear power plant reactor coolant pump. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to- CHF becomes large as the heat flux decreases. The critical mass flux has the largest value for slow flow reduction rate. There is a pressure effect on the ratio of the transient CHF data to steady-state CHF data. Except under low system pressure conditions, the flow transient CHF was revealed to be conservative compared with the steady-state CHF data. Bowling CHF correlation and thermal hydraulic system code MARS show promising results for the prediction of CHF occurrence .

• Nuclear Engineering and Technology
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• 제36권5호
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• pp.388-402
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• 2004

A phenomena identification and ranking table(PIRT) was developed for a main steam line break (MSLB) event for the Advanced Power Reactor-1400 (APR-1400). The selectee event was a double-ended steam line break at full power, with the reactor coolant pump running. The developmental panel selected the fuel performance as the primary safety criterion during the ranking process. The plant design data, the results of the APR-1400 safety analysis, and the results of an additional best-estimate analysis by the MARS computer code were used in the development of the PIRT. The period of the transient was composed of three phases: pre-trip, rapid cool-down, and safety injection. Based on the relative importance to the primary evaluation criterion, the ranking of each system, component, and phenomenon/process was performed for each time phase. Finally, the knowledge-level for each important process for certain components was ranked in terms of existing knowledge. The PIRT can be used as a guide for planning cost-effective experimental programs and for code development efforts, especially for the quantification of those processes and/or phenomena that are highly important, but not well understood.