• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 proceedings of the second asia international conference on tribology
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• pp.289-290
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• 2002

Stainless steel ball bearings are used in the control element drive mechanism and driving mechanisms such as step motor and gear boxes for the integral nuclear reactor, SMART. The bearings operate in pressurized pure water (primary coolant) at high temperature and should be lubricated with only this water because it is impossible to supply greases or any additional lubricant since the whole nuclear rector system should be perfectly sealed and the coolant cannot contain ingredients for bearing lubrication. Temperature of water changes from room temperature to about 120 degree Celsius and pressure rises up to 15MPa in the nuclear reactor. It can be anticipated that the frictional characteristics of the ball bearings changes according to the operating conditions, however little data are available in the literature. It is found that friction coefficient of 440C stainless steel itself does not change sharply according to temperature variation from the former research, and the friction coefficient is about 0.45 at low speed range. In this research frictional characteristics of the assembled ball bearings are investigated. A special tribometer is used to simulate the axial loading and the bearing operating conditions, temperature and pressure in the driving mechanism in the nuclear reactor. Highly purified water is used as lubricant ‘ and the water is heated up to 120 degree Celsius and pressurized to 15MPa. Friction force is monitored by the torque transducer.

• Design & Manufacturing
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• 제6권1호
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• pp.1-4
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• 2012

Plastic products are producted more than 70% of total processes in the injection molding. The injection molding process has 4 processes such as filling, packing, cooling and ejecting. It spends most of times in the cooling process. Therefore, it is important to control the mold temperature in producing plastic products. The time and system of cooling affect the product's quality and productivity. Especially, cooling time has about 60% of total injection cycle time. Therefore, we can improve a productivity by shortening cooling time. In this study, it was made a comparative study about cooling of linear channels and baffles and observed the variation of mold temperature on the coolant's temperature. As the result, the linear channel's cooling rate had faster than baffles and as coolant's temperature was increased, difference of cooling time was increased. Result of this study will be used widely to design for cooling system of injection mold.

• 한국자동차공학회논문집
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• 제14권2호
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• pp.39-48
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• 2006

An engine cooling system affects overall performances of an engine which has been recently requested higher power in more confined engine room. The design of efficient cooling system demands a great effort to effectively correlate with each components, such as water jacket, radiator, coolant pump, cooling fan, etc. Thus, the aim of this study is to provide the design tool of the cooling system in the early design stage by enabling for the designer to accurately predict the engine cooling performances. This user-friendly design tool has various ways to assemble each components and control the running condition with related database. The present design tool was simulated and compared with experimental data. As a result, the inlet and outlet temperature of the radiator agree very well with experiments. It was concluded that the present design tool could be effectively used for the design of the engine cooling system.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1140-1153
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• 2016

Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

• Bulletin of the Korean Chemical Society
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• 제24권3호
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• pp.285-290
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• 2003

An analytical method for the simultaneous determination of trace Ge, As and Se in biological samples by inductively coupled plasma/mass spectrometry has been investigated. The effects of added organic gas into the coolant argon gas on the analyte signal were studied to improve the detection limit, accuracy and precision. The addition of a small amount of methane (10 mL/min.) into the coolant gas channel improved the ionization of Ge, As and Se. The analytical sensitivity of the proposed Ar/CH₄system was superior by at least two-fold to that of the conventional Ar method. In the present method, the detection limits obtained for Ge, As and Se were 0.014, 0.012 and 0.064 ㎍/L, respectively. The analytical reliability of the proposed method was evaluated by analyzing the certified standard reference materials (SRM). Recoveries of 99.9% for Ge, 103% for As, 96.5% for Se were obtained for NIST SRM of freeze dried urine sample. The proposed method was also applied to the biological samples.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• 설비공학논문집
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• 제26권6호
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• pp.287-293
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• 2014

Internal combustion engines release 30~40% of the energy from fossil fuels into the atmosphere in the form of exhaust gases. By utilizing this waste heat, plenty of energy can be conserved in the auto industry. Thermoelectric generation is one way of transforming the energy from engine's exhaust gases into electricity in a vehicle. The thermoelectric generators located on the exhaust pipe have been developed for vehicle applications. Different experiments with thermoelectric generators have been conducted under various test conditions as following examples: hot gas temperature, hot gas mass flow rate, coolant temperature, and coolant mass flow rate. The experimental results have shown that the generated electrical power increases significantly with the temperature difference between the hot and the cold side of the thermoelectric generator and the gas flow rate of the hot-side heat exchanger. In addition, the gas temperature of the hot-side heat exchanger decreases with the length of the thermoelectric generator, especially at a low gas flow rate.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.549-558
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• 2004

TTo comply with increasingly strict regulations for protection against radiation exposure, many nuclear power plants have been working ceaselessly to reduce and control both the radiation sources within power plants and the radiation exposure experienced by operational and maintenance personnel. Many research studies have shown that deposits of irradiated corrosion products on the surfaces of coolant systems are the main cause of occupational radiation exposure in nuclear power plant. These corrosion product deposits on the fuel-clad surface are also known to be main factors in the onset of axial offset anomaly (AOA). Hence, there is a great deal of ongoing research on water chermistry and corrosion processes. In this study, a magnetic filter with permanent magnets was devised to remove the corrosion products in the coolant stream by taking advantage of the magnetic properties of the corrosion products demonstrated a removal efficiency of over 90% for particles above 5${\mu}m$. This finding led to the construction of an electromagnetic device that causes the metallic particulates to flocculate into larger aggregates of about 5${\mu}m$ in diameter by using a novel application of electromagnetic flocculation on radioactive corrosion products.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1307-1313
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• 2019

Recent analysis results with realistic assumptions provide the variability of operator allowable time for the initiation of aggressive cooldown under small break loss of coolant accident or steam generator tube rupture with total failure of high pressure safety injection. We investigated how plant risk may vary depending on the variability of operators' failure probability of timely initiation of aggressive cooldown. Using a probabilistic safety assessment model of a nuclear power plant, we showed that plant risks had a linear relation with the failure probability of aggressive cooldown and could be reduced by up to 10% as aggressive cooldown is more reliably performed. For individual accident management, we found that core damage potential could be gradually reduced by up to 40.49% and 63.84% after a small break loss of coolant accident or a steam generator tube rupture, respectively. Based on the importance of timely initiation of aggressive cooldown by main control room operators within the success criteria, implications for improvement of emergency operating procedures are discussed. We recommend conducting further detailed analyses of aggressive cooldown, commensurate with its importance in reducing risks in nuclear power plants.