• 대한기계학회논문집
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• 제9권5호
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• pp.638-646
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• 1985

본 논문에서는 펌프가 동시에 전원차단 되었을 때 송출밸브의 운전조건(개도, 폐쇄속도), 계통의 기하학적 형상(취수 및 배수관로의 길이, 단면적, 재질, 콘덴서의 높이등)과 해면상태(간막의 차에 의한 수위, 파도)의 변화가 계통내 수격현상에 미치 는 영향에 대하여 수치해석 하였다.

• 동력기계공학회지
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• 제9권4호
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• pp.175-180
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• 2005

Incorrections between model and plant are parameter, system order uncertainties and modeling error due to disturbance like friction. Therefore to achieve a good tracking performance, adaptive discrete time sliding mode tracking controller is used under time-varying desired position. Based on the diophantine equation, a new discrete time sliding function is defined and utilized for the control law. Robustness is increased by using both a recursive least-square method and a sliding function-based nonlinear feedback. The effectiveness of the proposed control algorithm is proved by the results of simulation and experiment.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1994년도 추계학술발표회 초록집
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• pp.135-140
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• 1994

An improved steam generator model has been developed for the DSNP simulation of normal operational transient behavior of CANDU nuclear power plant. For more realistic prediction of steam generator behavior during transient, tube bundle region is divided into two separate control volumes, subcooled region and saturated region, and the variation of thermal hydraulic properties in the control volume is accounted for more realistic estimates of outlet enthalpy of each control volume. Test results for typical CANDU operational transient case show reasonable transient behavior of steam generator with overall CANDU operation and improved operational characteristics of steam generator with power variation.

• 대한조선학회 특별논문집
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• 대한조선학회 2009년도 특별논문집
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• pp.71-79
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• 2009

The world-wide ship construction market is forecast that the considerable portion of shipbuilding and oceanic plant industry will be transferred consequently in China after 5 or 10 years. This point of view where the Korean ship construction industry seizes the initiative from the world-wide ship construction/oceanic field, we must cultivate technical power of base technology, and focus our interests on the development of core parts. In this study, our proprietary remotely operated valve actuator system with power line technology was developed to enhance the installation and commissioning process by our own technology. This paper describes the new design and functions of the remotely operated valve system for shipbuilding and offshore market especially for FPSO.

• 터널과지하공간
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• 제21권1호
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• pp.11-19
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• 2011

지열에너지는 여러 신재생에너지원 중에서도 기저부하를 담당할 수 있는 중요한 자원으로 인식되고 있다. 국내에서도 천부지열을 이용한 지열냉난방은 효율 높은 신재생에너지 활용 사업으로 그 보급이 활성화 되어 있다. 반면, 전세계적으로 지열 발전 기술이 진일보하고, 그 시장이 크게 확대되고 있는 상황에서 아직까지 국내의 심부 지열을 이용한 지열 발전 기술은 낮은 단계에 머무르고 있다. 이러한 조건에서 2010년 12월에 국내 최초의 EGS(Enhanced Geothermal System) 지열 발전 상용화 기술 개발 과제가 착수되었다. 총 5개년의 기간으로 수행되는 이 과제는 2단계로 구분되어 진행될 계획이다. 처음 2년의 1단계에서는 3 km 심도에서 최소 $100^{\circ}C$의 지열저류층 온도를 확인하는 것을 주요 과제 내용으로 하여 지중 지열수 순환시스템의 설계가 이루어질 예정이다. 이후 3년을 통해 수행될 2단계에서는 5 km 심도의 생산정과 주입정 등 두 개의 지열발전정을 설치하고, 수리자극을 통하여 온도 $180^{\circ}C$의 지열저류층에서 유량 40 kg/s 이상의 지열수를 활용하는 MW급 지열발전소를 건립 운영하게 된다. 이 사업을 성공적으로 추진하기 위하여 현재 지질, 수리지질, 지구물리, 암석역학, 플랜트 엔지니어링 등 다양한 분야의 산학연 연구 기관 등이 망라되어 연구진을 구성한 상태이며, 이후 관심있는 여러 기관과 연구자들의 지원과 참여를 기대하고 있다.

• Nuclear Engineering and Technology
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• 제47권1호
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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.

• 비파괴검사학회지
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• 제31권5호
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• pp.543-551
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• 2011

Steam generator tubes for nuclear power plants contain the local shape transitions on their inner or outer surface such as dent, bulge, over-expansion, eccentricity, deflection, and so on by the application of physical force during the tube manufacturing and steam generator assembling and by the sludge (that is, corrosion products) produced during the plant operation. The structural integrity of tubes will be degraded by generating the corrosive crack at that location. The profilometry using the traditional bobbin probes which are currently applied for measuring the profile change of tubes gives us basic information such as axial locations and average magnitudes of deformations. However, the three-dimensional quantitative evaluation on circumferential locations, distributional angle, and size of deformations will have to be conducted to understand the effects of residual stresses increased by local deformations on corrosive cracking of tubes. Steam generator tubes of Korean standard nuclear power plants expanded within their tube-sheets by the explosive expansion method and suffered from corrosive cracks in the early stage of power operation. Thus, local deformations of steam generator tubes at the top of tube-sheet were measured with an advanced rotating probe and a laser profiling system for the two cases where the tubes expanded by the explosive expansion method and hydraulic expansion. Also, the trends of eccentricity, deflection, and over-expansion of tubes were evaluated. The advanced eddy current profilometry was confirmed to provide accurate information of local deformations compared with laser profilometry.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.301-308
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• 2011

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.

• 한국유체기계학회 논문집
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• 제18권6호
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• pp.45-56
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• 2015

This paper aims to develop a submerged propeller turbine for micro hydropower plant which allows to sustain high values of efficiency in a broad range of hydrological conditions (H=2~6 m, $Q=0.15{\sim}0.39m^3/s$). The two aspects to be considered in this development are mechanical simplicity and high-efficiency operation. Unlike conventional turbines that have spiral casing and gear box, this is directing driving and no spiral casing. A 10 kW class turbine which has the most high potential of the power generation has been developed. The most important element in the design of turbine is the runner blade. The initial blade is designed using inverse design method and then the runner geometry is modified by classical hydraulic method. The design process is carried out in two steps. First, the blade shape is fix and then other components of submerged propeller turbine are designed. Computational fluid dynamics analyses based on the Navier-Stokes equations have been used to obtain overall performance data for the blade and the full turbine, respectively. The results generated by performance parameters(head, guide vane opening angle and rotational speed) variations are theoretically analysed. The evaluation criteria for the blade and the turbine performances are the pressure distribution and flow's behavior on the runner blades and turbine. The results of simulation reveals an efficiency of 91.5% and power generation of 10.5kW at the best efficiency point at the head of 4m and a discharge of $0.3m^3/s$.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.