• 대한기계학회논문집A
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• 제36권6호
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• pp.637-644
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• 2012

증기 발생기 내부의 U-tube와 지지 구조간의 충격에 의하여 발생하는 프레팅 마모는 원자력 발전소 안전성에 영향을 미치게 된다. 증기발생기의 신뢰성을 향상시키기 위하여 이러한 프레팅 마모 현상을 평가하는 것이 필요하며, 본 연구는 프레팅 마모현상을 정성적, 정량적으로 규명하기 위하여 증기발생기의 실제 상황과 같은 조건의 온도와 압력하에서 실험을 수행하였다. 다양한 실험조건에 대하여 기본적인 실험을 수행하였으며 일률과 마모량의 관계를 온도에 따라 구하였다. $90^{\circ}C$, $200^{\circ}C$, $340^{\circ}C$ 각각의 온도에서의 마모상수는 $9.051{\times}10^{-16}\;Pa^{-1}$, $3.009{\times}10^{-15}\;Pa^{-1}$, $2.235{\times}10^{-15}\;Pa^{-1}$로 구해졌으며 특히 저온 수중상태의 마모상수는 물의 점도의 영향으로 상온 공기중의 값보다 작은 것으로 나타났다.

• 대한기계학회논문집B
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• 제34권4호
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• pp.437-442
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• 2010

고온, 고압의 유체가 흐르는 탄소강 배관에서는 유동가속부식으로 인한 배관감육 현상이 발생할 수 있다. 화력 및 원자력발전소에서 유동가속부식으로 인한 배관 손상시 고비용의 보수와 발전 정지를 유발할 뿐 아니라 발전소 신뢰도 및 안전성에 영향을 미칠 수도 있다. CHECWORKS 프로그램은 국내 발전소에서 유동가속부식에 의한 배관 손상을 예방하기 위하여 배관 두께검사 데이터를 평가하고 검사 계획을 수립하는데 이용되어 왔다. 그러나 상기 프로그램은 원전 2차측 배관 모두를 데이터베이스화한 후에 배관라인 그룹별로 유동가속부식 손상을 예측하기 때문에 국부적으로 감육에 민감한 부위를 찾는데 어려움이 있다. 본 논문에서는 CHECWORKS 프로그램을 이용하여 해석을 수행하고 수치해석을 통하여 검증할 수 있는 방법론을 기술하였다. 또한 국내 원전 2개의 배관 라인그룹에 대하여 CHECWORKS 프로그램을 이용한 유동가속부식 민감 부위를 FLUENT를 이용한 수치해석 결과와 비교하였다.

• 한국시뮬레이션학회논문지
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• 제29권3호
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• pp.9-18
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• 2020

대규모 화력 발전소 제어용 시뮬레이터 개발은 급수/증기 계통, 공기/연소가스 계통, 미분탄 계통 및 터빈/발전기 계통으로 구성되며, 기계적인 터빈/발전기를 제외하고 모든 계통에 대하여 모델링이 가능하다. 현재까지 화력발전의 일부 계통에 대한 신경망 시뮬레이터 개발에 대한 시도는 있었으나 전체 계통에 대한 시뮬레이터 개발은 완성된 적이 없다. 특히 모든 발전사의 핵심 기술 개발중 하나인 오토튜닝은 정확도가 높은 모든 계통에 대한 모델링이 완성되어야 이룰 수 있는 기술이다. 이에 본 논문은 신경망 알고리즘을 이용하여 시스템을 설계할 경우 가장 핵심인 입출력 관계에 대한 변수를 모든 계통에 대하여 정의하였다. 시뮬레이션을 수행한 결과 실제 보일러 계통의 95~99% 이상 정확도를 보임에 따라 본 시뮬레이터에 현장 PID 제어기를 결합할 경우 고장진단이나 오토튜닝에 활용 가능할 것이다.

• 한국대기환경학회지
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• 제31권5호
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• pp.461-471
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• 2015

Base on emission factors derived from National Institute of Environmental Research, Particulate matter from combined cycle power plants (CCPPs) has been estimated to be a important source of $PM_{10}$. Generally there is no serious emission of particulate matter in CCPPs. because the fuel of them is natural gas. But emission gas after long shut down season has very high dust content. Therefore $PM_{10}$ emission rate is dependent on its operation mode. In this study, particulate dispersion study for new city near CCPPs complex has performed using CALPUFF model for three case. $PM_{10}$ concentration has big difference between normal operation and 2 case start-up condition after long shutdown. In normal operating conditions, daily $0.32{\sim}0.50{\mu}g/m^3$ influence on of the surrounding area. But when 1~2 aerobic high concentration discharged conditions, average concentration is higher about $9.2{\sim}34.1{\mu}g/m^3$ than normal operating conditions.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1005-1014
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• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1215-1222
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• 2009

The present work pertains to a research program to study Molten Fuel-Coolant Interactions (MFCI), which may occur in a nuclear power plant during a hypothetical severe accident. Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant) were investigated in the MISTEE (Micro-Interactions in Steam Explosion Experiments) facility by performing well-controlled, externally triggered, single-droplet experiments, using a high-speed visualization system with synchronized digital cinematography and continuous X-ray radiography. The current study is concerned with the MISTEE-NCG test campaign, in which a considerable amount of non-condensable gases (NCG) are present in the film that enfolds the molten droplet. The SHARP images for the MISTEE-NCG tests were analyzed and special attention was given to the morphology (aspect ratio) and dynamics of the air/ vapor bubble, as well as the melt drop preconditioning. Energetics of the vapor explosion (conversion ratio) were also evaluated. The MISTEE-NCG tests showed two main aspects when compared to the MISTEE test series (without entrapped air). First, analysis showed that the melt preconditioning still strongly depends on the coolant subcooling. Second, in respect to the energetics, the tests consistently showed a reduced conversion ratio compared to that of the MISTEE test series.

• 시스템엔지니어링학술지
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• 제14권2호
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• pp.58-66
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• 2018

This paper represents a systematic approach aimed at improving the performance of the proportional integral (PI) controller for the Advanced Power Reactor (APR) 1400 Feedwater Control System (FWCS). When the performance of the PI controller offers superior control and enhanced robustness, the steam generator (SG) level is properly controlled. This leads to the safe operation and increased the availability of the nuclear power plant. In this paper, a systems engineering approach is used in order to design a novel PI controller for the FWCS. In the reverse engineering stage, the existing FWCS configuration, especially the characteristics of the feedwater controller as well as the feedwater flow path to each SG from the FWCS, were reviewed and analysed. The overall block diagram of the FWCS and the SG was also developed in the reverse engineering process. In the re-engineering stage, the actual design of the feedwater PI controller was carried out using a genetic algorithm (GA). Lastly, in the validation and verification phase, the existing PI controller and the PI controller designed using GA method were simulated in Simulink/Matlab. From the simulation results, the GA-PI controller was found to exhibit greater stability than the current controller of the FWCS.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1955-1962
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• 2020

In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.

• 한국소음진동공학회논문집
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• 제20권1호
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• pp.29-35
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• 2010

This paper studies causes of the L-1 blade damage of a low pressure turbine, which was found during the scheduled maintenance, in 500 MW fossil power plants. Many failures of turbine blades are caused by the coupling of aerodynamic forcing with bladed-disk vibration characteristics. In this study the coupled vibration characteristics of the L-1 turbine bladed-disk in a fossil power plant is shown for the purpose of identifying the root cause of the damage and confirming equipment integrity. First, analytic and experimental modal analysis for the bladed-disk at zero rpm as well as a single blade were performed and analyzed in order to verify the finite element model, and then steady stresses, natural frequencies and corresponding mode shapes, dynamic stresses were calculated for the bladed-disk under operation. Centrifugal force and steady steam force were considered in calculation of steady and dynamic stress. The proximity of modes to sources of excitation was assessed by means of an interference diagram to examine resonances. In addition, fatigue analysis was done for the dangerous modes of operation by a local strain approach. It is expected that these dynamic characteristics will be used effectively to identify the root causes of blade failures and to perform prompt maintenance.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1436-1445
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• 2021

To obtain the dynamic characteristics of reactor secondary circuit under transient conditions, the system analysis program was developed in this study, where dynamic models of secondary circuit were established. The heat transfer process and the mechanical energy transfer process are modularized. Models of main equipment were built, including main turbine, condenser, steam pipe and feedwater system. The established models were verified by design value. The simulation of the secondary circuit system was conducted based on the verified models. The system response and characteristics were investigated based on the parameter transients under emergency shutdown and overload. Various operating conditions like turbine emergency shutdown and overspeed, condenser high water level, ejector failures were studied. The secondary circuit system ensures sufficient design margin to withstand the pressure and flow fluctuations. The adjustment of exhaust valve group could maintain the system pressure within a safe range, at the expense of steam quality. The condenser could rapidly take out most heat to avoid overpressure.