• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.11-25
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• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.336-343
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• 1995

The Self Tuning Regulator(STR) method which is an approach of adaptive control theory, is ap-plied to design the fully automatic power controller of the nonlinear reactor model. The adaptive control represent a proper approach to design the suboptimal controller for nonlinear, time-varying stochastic systems. The control system is based on a third­order linear model with unknown, time-varying parameters. The updating of the parameter estimates is achieved by the recursive extended least square method with a variable forgetting factor. Based on the estimated parameters, the output (average coolant temperature) is predicted one-step ahead. And then, a weighted one-step ahead controller is designed so that the difference between the output and the desired output is minimized and the variation of the control rod position is small. Also, an integral action is added in order to remove the steady­state error. A nonlinear M plant model was used to simulate the proposed controller of reactor power which covers a wide operating range. From the simulation result, the performances of this controller for ramp input (increase or decrease) are proved to be successful. However, for step input this controller leaves something to be desired.

• Korean Journal of Materials Research
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• v.25 no.10
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• pp.547-551
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• 2015

The present study prepared molybdenum trioxide ($MoO_3$), the most important intermediate of molybdenum metal, by using a fluidized bed reactor for the thermal decomposition of ammonium molybdate (AM) in the presence of an air flow. During the process of fluidizing the sample inside the reactor, the reaction time and temperature were optimized with a close analysis of the X-ray diffraction (XRD) data and with thermogravimetric analysis (TGA). In particular, the temperature level, at which the AM decomposition is completed, is very important as a primary operating parameter. The analysis of the XRD and TGA data showed that the AM decomposition is almost completed at ${\sim}350^{\circ}C$ with a reaction time of 30 min. A shorter reaction time of 10 min. required a higher reaction temperature of ${\sim}500^{\circ}C$ with the same air flow rate to complete the AM decomposition. A sharp rise in the decomposition efficiency at a temperature ranging between 320 and $350^{\circ}C$ indicated a threshold for the AM decomposition. The operating conditions determined in this study can be used for future scale-ups of the process.

• Composites Research
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• v.34 no.4
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• pp.249-256
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• 2021

In this paper, the densification of a carbon/carbon composite during a chemical vapor infiltration (CVI) process is studied using a chemo-mechanical model. The multi-physics numerical model, developed in the previous research, couples computational fluid dynamics and major chemical reactions in the reactor. The model is especially utilized to study the effect of flow behavior around the preform on the densification. Four different types of "flow-guide" structures are placed to alter the gas flow around the preform. It is shown that the flow pattern and speed around the preform can be controlled by the guide structures. The process simulations demonstrate that the average density and/or density distribution of the preform can be improved by controlling the gas flow around the perform. In this study, a full industrial-scale reactor and process parameter were used.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.34 no.6
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• pp.385-392
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• 2021

Interest in the seismic performance of nuclear facilities under strong earthquakes has increased because their nonlinear response is important. In this paper, we proposed appropriate parameters for the nonlinear finite element analysis of a concrete material model, for a reinforced concrete (RC) shear wall in nuclear facilities: maximum tensile strength, dilation angle, and damage parameter. The study of the effects of the important parameters, on the nonlinear behavior and shear failure mode of the RC shear wall having low aspect ratio, was conducted using ABAQUS finite element analysis program. Based on the study results the nonlinear response of a nuclear reactor containment building (RCB) subjected to a strong earthquake was evaluated using nonlinear time-history analysis.

• Journal of Korea Water Resources Association
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• v.47 no.5
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• pp.483-490
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• 2014

The design procedures for a biological reactor and a secondary settling tank (SST) of an activated sludge system are based on the steady state design method (Ekama et al., 1986; WRC, 1984) and the 1-D flux theory design method (Ekama et al., 1997), respectively. This study combined both of the design procedures, to determine the optimum sludge concentration in the reactor and the best design with the lowest cost. The best design of the reactor volume and the SST diameter at the optimum sludge concentration were specified with varying wastewater and sludge characteristics, temperature, sludge retention time (SRT) and peak flow rate. The effects of the influent wastewater characteristics, such as substrate concentration and unbiodegradable particulate fraction, were found to be considerable, but the effect of unbiodegradable soluble fraction was to be negligible. The effects of sludge settling characteristics, were also significant. SRT, as an operating parameter, was found to be an important factor for determining the optimum sludge concentration. However, the effect of temperature was found to be small. Furthermore, for designing a large scale wastewater treatment plant, the number of reactors or SSTs could be estimated, by dividing the total reactor volume or SST area. The new combined design procedure, proposed in this research, will be able to allow engineers to provide the best design of an activated sludge system with the lowest cost.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.6
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• pp.1069-1076
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• 2002

The high temperature creep behavior of heat machine systems such as aircraft engines, boilers and turbines in power plants and nuclear reactor components have been considered as an important and needful fact. There are considerable research results available for the design of high temperature tube materials in power plants. However, few studies on the Initial Strain Method (ISM) capable of securing repair, maintenance, cost loss and life loss have been made. In this method, 3 long time prediction Of high temperature creep characteristics can be dramatically induced through a short time experiment. The purpose of present study is to investigate the high temperature creep lift of Udimet 720, SCM 440-STD61 and 1Cr-0.5Mo steel using the ISM. The creep test was performed at 40$0^{\circ}C$ to $700^{\circ}C$ under a pure loading. In the prediction of creep life for each materials, the equation of ISM was superior of Larson-Miller Parameter(LMP). Especially, the long time prediction of creep life was identified to improve the reliability.

• 한국신재생에너지학회:학술대회논문집
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• 2011.05a
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• pp.156.1-156.1
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• 2011

It is of great importance to predict operating parameter characteristics of an integrated fuel processor by the increased life-time and system performance. In this study, computational analysis is performed to gain fundamental insights on transport phenomena and chemical reactions in reformer which consists of preheating, steam reforming, and water gas shift reaction beds. Also, a top-fired burner locates inside of the reforming system. The combustor is providing thermal energy necessary for the steam reforming bed which is a endothermic catalytic reactor. Two-dimensional numerical model of the integrated fuel processing system is introduced for the analysis of heat and mass transport phenomena as well as surface kinetics and catalytic process. A kinetic model was developed and then computational results were compared with the experimental data available in the literature. Subsequently, parameter study using the validated steam methane reforming model was conducted by considering operating parameters, i.e. steam to carbon ratio and temperature.

• Transactions on Control, Automation and Systems Engineering
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• v.1 no.2
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• pp.141-146
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• 1999

This paper describes the communication method of an auto-startup system. The Auto-Startup system is designed to operate a nuclear power plant automatically during the startup operation . In general , the operations during startup in existing plant have only been manually controlled by the operator. The manual operation caused to the operator mistake. The Auto-Startup system consists of the Distributed Control System(DCS) and G2 (Expert System). Also, Functional Test Facility(FTF) provides the plant's real-data for an Auto-Startup system. So, it is necessary to develop the communication method between these systems. We developed two methods ; one is a network and the other is a hardwire line. To communicate between these systems (DCS-G2 and DCS-FTF) , we developed the communication program. In case of DCS-FTF, we used the TCP/IP and VXI. BUt, in case of DCS-G2 , we , what it called , developed the bridge program using the GSI(G2 Standard Interface). We test to check the function of the important parameter, in time, for analysis of the developed communication method. The results are a good performance when we check the communication time of important parameter. We conclude that Auto-startup system could save heat-up time about at least 5 hours and reduced the change of the reactor operation and trip.

• Journal of the Korean Institute of Telematics and Electronics
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• v.8 no.6
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• pp.8-16
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• 1971

This paper describes the transistorized linear counting ratemeter which can be uses to indicate on the meter or record the counting rates of the nuclear radiations produced from the atomic reactor or from the radio isotopes. Tte feature of this ratemeter is the use of the transistor chopper for good stabilization. At the input stage of the a. c. amplifier a composite emitter follower buffer stage has been used to give the high input impedance. A hybrid parameter equivalent circuit was modeled for the analysis of this buffer stage. The counting rates can be linearly measured from few CPS up to 100KCPS in 4 ranges. The resolution is less than 0.5$\mu$sec and the output drift at the room temperature with 7-hour continuous operation is in the order of $\pm$0.5$\mu$A.