• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.85-90
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• 1996

In this paper, the self-organizing fuzzy logic controller is developed for water level control of steam generator. In comparison with conventional fuzzy logic controllers, this controller performs control task with no control rules at initial and creates control rules as control behavior goes on, and also modifies its control structure when uncertain disturbance is suspected. Selected parameters in the fuzzy logic controller are updated on-line by the gradient descent loaming algorithm based on the performance cost function. This control algorithm is applied to water level control of steam generator model developed by Lee, et al. The computer simulation results confirm good performance of this control algorithm in all power ranges. This control algorithm can be expected to be used for automatic control of feedwater control system in the nuclear power plant with digital instrumentation and control systems.

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.14-20
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• 2012

Full load rejection capability of nuclear power plant depends primarily on steam dump capacity (SDCAP) and steam generator level control capability. Recently, Ulchin Units 1&2 have performed stretched power uprate (SPU) and replacement steam generator (RSG) projects, which increase the power by 4.5 percent. They change major design or operating parameters and especially reduces steam dump capacity at full power due to increase of the steam flow. The reduction of SDC after SPU results in degradation of heat removal capability in full load rejection transients. Therefore, we should perform evaluation to determine whether reactor trips occur in large load rejection transients. Uchin Units 1&2 have experienced full load rejection (FLR) three times from 2004 to 2010. Operating data from the plant occurrence of FLR at Ulchin Units 1&2 showed that steam generator (SG) level transients were limiting in point of reactor trip. However the plant had never reached reactor trip in the FLR and successfully continued in house load operation. The parameters and setpoints for the SG will be changed if the SG is replaced. Therefore, we evaluated the appropriateness of steam dump, main feedwater and steam generator water level control system preventing the plant from reactor trip in case of FLR by the parameter sensitivity study whether SG water level operated smoothly after SPU and RSG projects.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2006년도 제37회 하계학술대회 논문집 D
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• pp.1871-1872
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• 2006

발전소를 운영함에 있어 보일러 제어시스템에는 수 많은 제어루프가 있으며, 이들 중에서 급수제어는 발전소 보일러제어 시점에서 매우 중요한 역할을 한다. 본 논문은 필리핀 일리한 복합발전소의 시운전 당시부터 고질적인 급수제어 불안정에 대해 원인과 분석을 통하여 근본적인 원인을 찾아내고 해결하였으며, 복잡한 제어로직을 해독하고 개선하여 원활한 제어특성을 갖도록 하여 발전소 운전에 안정성과 신뢰성을 확보한 내용을 기술하였다.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.96-99
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• 2014

The aim of the study was to evaluate the water treatment of pressurized water reactor secondary side by the mixed amine of ammonia and ethanolamine, from the standpoint of corrosion control, as compared with all volatile treatment of ammonia. The pressurized water reactor systems have switched a secondary side pH control agent to minimize the corrosion in the moisture separator/reheater and feedwater heater systems and the transport of corrosion products into steam generator. As results of field test, pH was increased in the steam generator and the wet steam area of moisture separator/reheater and the concentration of Fe were decreased by more than 50% as compared with water treatment of ammonia.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2011년도 제42회 하계학술대회
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• pp.1804-1805
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• 2011

In recent year there has been an increasing interest in the dynamic simulation of complex systems. This study uses a large-scale forty-seventh order fossil fuel power plant. Twenty-three state variables are associated with the physical processes and twenty-four state variables associated with the control system. The plant model is expected to predict all dominant effects in a steady and transient state. In this study, the power plant model is reorganized into four subsystems, each with its controller, and the four connected to each other through a manager, which is a fifth part to the system. The four parts of the unit are the boiler system, steam turbine system, condenser system, and feedwater system.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.235-241
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• 1995

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor Power Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.514-519
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• 1997

The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control roe drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.617-621
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• 2003

The purpose of this paper is to analyze piping failure trend of safety pipings In domestic nuclear power plants. First, database for the piping failure was constructed with 105 data fields. The database includes plant population data, event data, and service history data. 7 kinds of piping failures in domestic NPPs were investigated. Among the 7 cases, detailed root causes were investigated for 3 cases. The first one is pipe wall thinning in main feedwater pipings of Westinghouse 3 loop type plants. The root cause of the wall thinning was flow accelerated corrosion near welding area. The next one is leak event in chemical and volume control system(CVCS) due to vibration. Some cracks occurred in socket welding area. The events showed that the integrity or socket weld is very vulnerable to vibration. The last one is also a leak event in primary sampling line in Korean standard reactor due to thermal fatigue. Although the structural integrity was not maintained by the events, there was no effect on nuclear safety in the above 3 piping failure eases.

• 조명전기설비학회논문지
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• 제12권4호
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• pp.6-13
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• 1998

본 연구에서는 증기유량, 중기온도 주 급수 온도 및 유량둥과 같은 외란에 의해 증기발생기의 웅축 및 팽창효 과가 발생하여 수위조절에 어려웅이 발생하는 문제를 신경망-PI제어기를 이용해 효과적으로 제어하는 연구를 하였다. 종래의 PI제어기는 저 출력에서 효과적으로 제어 할 수 없어 기동시는 숙련자의 노련한 기술이 필요한데 반해 본 연구에서 적용한 신경망 PI제어기는 Kp,Ti 파라매터를 외란에 영향을 줄 수 있는 파라메터를 고려하여 신경망을 이용해 튜닝함으로서 웅축 및 팽창효과를 줄이고 외란 및 설정치 변경에 대해 효과적으로 제어 될 수 있음을 실험과 시뮬레이션을 통해 확인 할 수 있었다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1998년도 하계학술대회 논문집 A
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• pp.189-192
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• 1998

This paper presents the basis for the need of feed-water deaerator installation in nuclear power plant and has been conducting reviews to understand the mechanism of dissolved oxygen(DO) formation as well as removal. DO is produced by feedwater make-up, air inleakage in vacuum system and condensation in condenser, etc. and removed by mechanical method and chemical method. DO has an influence on S/G in the form of denting, pitting, fretting, etc. DO control by deaerator has an great effect on the integrity of S/G in nuclear power plant.