• Journal of Advanced Marine Engineering and Technology
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• v.28 no.1
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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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• v.36 no.5
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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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.

• Proceedings of the Korea Committee for Ocean Resources and Engineering Conference
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• 2004.05a
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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).

• Proceedings of the KSME Conference
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• 2001.06a
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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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• v.1 no.3
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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.

• Proceedings of the SAREK Conference
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• 2005.11a
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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.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.9
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• pp.599-605
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• 2015

The revised U.S. Nuclear Regulatory Commission(NRC), Regulatory Guide(RG) 1.20, rev.3 requires the evaluation of the potential adverse effects from pressure fluctuations and vibrations on piping and components for the reactor coolant, steam, feedwater, and condensate systems. Detailed vibration analyses for the systems attached to the steam generator are very difficult, because these piping systems are very complicated. This paper suggests a screening method for the flow-induced vibration of acoustic resonances and pump-induced vibration of the piping systems attached to the steam generator in order to conduct the APR1400 comprehensive vibration assessment program. This paper seeks to address the areas such as potential vibration sources, and methods to prevent the occurrence of acoustic resonances and pump-induced vibration of piping systems attached to the steam generator, for conducting the APR1400 comprehensive vibration assessment program. The screening method in this paper will be used to estimate the flow-induced vibration of the piping systems attached to the steam generator for the APR1400.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.20 no.1
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• pp.464-469
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• 2019

A remote monitoring panel and control system was developed to control various valves and access control chambers, including gas shutoff valves used in CBR(Chemical, Biological and Radiological) facilities. The remote monitoring panel consisted of a main panel installed in the NBC (Nuclear, Biological and Chemical) control room and auxiliary panel installed in the clean room, and the size was divided into pure control and control including CCTV. This system can be monitored and controlled remotely according to the situation where an explosion door and gas barrier door can occur during war and during normal times. This system is divided into normal mode and war mode. In particular, it periodically senses the operation status of various valves, sensors, and filters in the CBR facilities to determine if each apparatus and equipment is in normal operation, and remotely alerts situation workers when repair or replacement is necessary. Damage due to the abnormal operation of each device in the situation can be prevented. This enables control of the blower, supply and exhaust damper, emergency generator, and coolant pump according to the state of shutoff valve and positive pressure valve in the occurrence of NBC, and prevents damage caused by abrupt inflow of conventional weapons and nuclear explosions.