• 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.

• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.745-750
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• 1996

An analysis of a loss of residual heat removal system (RHRS) event during midloop operation after reactor shutdown was performed using the RELAP5/MOD3 thermal-hydraulic computer code. The experimental data of a 5% cold leg break test conducted at the ROSA-IV Large Scale Test Facility (LSTF) to simulate a main coolant pump shaft seal removal event during midloop operation of a Westinghouse-type PWR were used in the analysis. The predicted core boiling time and the peak primary system pressure showed good agreements with the measured data. Some differences between the calculational results and the experimental results were, however, found in areas of the timing of loop seal clearing and the temperature distribution in a pressurizer. Other calculational problems identified were discussed as well.

• 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).

• 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.

• Journal of Electrical Engineering and Technology
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• 제1권3호
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• pp.359-365
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• 2006

Since the start of the KSTAR (Korea Superconducting Tokamak Advanced Research) project, Instrumentation and Control (I&C) of the Neutral Beam Test Facility (NB-TF) has been striving to answer diverse requests arising from various facets during the project's development and construction phases. Hard-wired electrical circuits have been designed, tested, fabricated, and finally installed to the relevant parts of the system. In relation to the vacuum system I&C, controlling functions for the rotary pumps, a Roots pump, two turbomolecular pumps, and four cryosorption pumps have been constructed. I&C for the ion source operation are the temperature and flow rate signal monitoring, Langmuir probe signal measurements, gradient grid current measurements, and arc detector circuit. For the huge power system to be monitored or safely operated, many temperature measurement functions have also been implemented for the beam line components like the neutralizer, bending magnet, ion dump, and calorimeter. Nearly all of the control and probe signals between the NB test stand and the control room were made to be transmitted through the optical cables. Failures of coolant flow or beam line vacuum pressure were made to be safely blocked from influencing the system by an appropriate interlock circuit that will shut down the extraction voltage application to the system or prevent damages to the vacuum components. Preliminary estimation of the beam power through the calorimetric measurement shows that 87.9% of the total power of the 60kV/18A beam with 200 seconds duration is absorbed by the calorimeter surface. Most of these I&C results would be highly appropriate for the construction of the main NBI facility for the KSTAR national fusion research project.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2005년도 동계학술발표대회 논문집
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• pp.353-358
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• 2005

The heat transfer and pressure drop of supercritical $CO_2$ cooled in a helically coiled tube was investigated experimentally. The experiments were conducted without oil in the refrigerant loop. The experimental apparatus of the refrigerant loop consist of receiver, a variable speed pump, a mass flowmeter, a pre-heater, a gas cooler(test section) and an isothermal tank. The test section is a helically coiled tube in tube counter flow heat exchanger with $CO_2$ flowed inside the inner tube and coolant( water) flowed along the outside annular passage, It was made of it copper tube with the inner diameter of 4.55[mm]. the outer diameter of 6.35 [mm] and length of 10000 [mm]. The refrigerant mass fluxes were $200^{\sim}600$ [kg/m2s] and the inlet pressure of gas cooler varied from 7.5 [MPa] to 10.0 [MPa]. The main results are summarized as follows : The heat transfer coefficient of supercritical $CO_2$ increases, as the cooling pressure of gas cooler decreases. And the heat transfer coefficient increases with the increase of the refrigerant mass flux. The pressure drop decreases in increase of the gas cooler pressure and increases with increase the refrigerant mass flux.

• 한국소음진동공학회논문집
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• 제25권9호
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• pp.599-605
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• 2015

U.S. Nuclear Regulatory Commission(NRC), Regulatory Guide(RG) 1.20(Rev.3, 2007년)은 원자로 냉각재계통, 주증기, 주급수 및 복수시스템의 주요 배관 및 기기에 대하여 압력변동 및 진동에 의한 잠재적 유해효과에 대한 평가를 요구한다. 그러나 증기발생기와 연결된 주증기, 주급수 및 복수시스템의 주요 배관 전체에 대하여 상세 해석하는 것은 매우 복잡하여 한계가 있다. 이 논문은 APR1400 원전의 종합진동평가(comprehensive vibration assessment program, CVAP)를 수행하기 위하여 증기발생기에 연결된 2차측 주요 배관의 음향공진과 펌프유발진동을 위한 간이평가 방법에 관한 것이다. 이 논문에서는 이러한 배관시스템의 잠재적 진동 원인이 무엇인지, 음향공진과 펌프유발진동의 가능성을 예방하기 위한 간이평가 방법은 무엇인지를 고찰하고자 한다. 이 논문은 APR1400 원전 증기발생기와 연결된 주증기 및 주급수 배관의 유동유발진동 간이평가를 위해 사용될 것으로 판단된다.

• 한국산학기술학회논문지
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• 제20권1호
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• pp.464-469
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• 2019

본 논문에서는 화생방 방호시설에 사용되는 가스 차단밸브를 비롯한 각종 밸브 및 출입통제 챔버를 제어할 수 있는 원격 감시패널과 제어시스템을 개발하였다. 원격 감시패널은 화생방 통제실에 설치되는 메인 패널과 청정 기계실에 설치되는 보조 패널로 구성하였으며, 그 크기를 순수 제어용과 CCTV를 포함한 제어용으로 구분하여 개발하였다. 본 시스템은 방폭문 및 가스 차단문이 전시 및 평시에 발생할 수 있는 상황별 상태에 따라 원격 감시 및 제어가 가능하며, 평시모드와 전시모드로 구분하여 각 모드 작동시 추가적인 활성창을 통하여 제어 정보를 표시한다. 특히, 화생방 방호시설 내부의 각종 밸브 및 센서 그리고 각종 여과기의 동작상태를 주기적으로 센싱하여 각각의 기구 및 기기들의 정상동작 여부를 파악하며, 수리 및 교체가 필요할 경우 이를 원격으로 상황 근무자에게 경보함으로써 위급상황에서 각각의 기기들의 동작이상으로 인한 피해를 미연에 방지한다. 이를 통하여 화생방 상황 발생시 차단밸브 및 양압밸브의 상태에 따른 송풍기, 급기 및 배기 댐퍼의 작동, 비상 발전기 및 냉각수 펌프 등의 제어가 가능하여 재래식 무기 및 핵폭발에 의한 폭풍압의 급격한 유입으로 인한 피해를 예방할 수 있다.