• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.392-397
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• 1997

고리 2호기 주급수 차단밸브의 경우 다른 호기와는 달리 구동용 질소압력 측정스위치에 의한 닫힘 연동신호가 설치되어 있다. 이러한 연동신호는 벨브 구동용 반구내 질소 압력 스위치 "고" 오동작에 의한 밸브차단 가능성이 있으며 이 경우 질소압력스위치가 저/고 경보창에 COMMON되어 있어 원인 규명에 어려움이 있다. 또한 질소압력 스위치 고장 및 질소가스 누설시 작업수행이 어렵고 위험이 따른다. 이러한 70년대 발전소 설계의 과잉설비를 제거하므로서 최적운전과 경제성 향상에 기여할 수 있으며 아울러 유지정비의 용이성과 밸브 불시닫힘을 미연에 예방할 수 있으리라 예상된다. 이와 관련하여 발전소 계통안전성, 기기안전성을 평가한 결과 기존의 안전해석결과가 유효하며 또한 FSAR 수정없이 이러한 설비변경이 가능하다는 결론에 도달하였다.결론에 도달하였다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.39-44
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• 2015

The fracture toughness of 2.25Cr-1Mo cast steel (SA217-WC9) samples which were taken from the check valves of feed water piping of Korean Standard Nuclear Power Plant(KSNPP) was measured by Master Curve method. The measured $T_0$ reference temperature of SA217-WC9 steel was $-30^{\circ}C$. The obtained $T_0$ was compared to the derived value from Charpy impact test data following to SINTEP procedure. The heat-to-heat variation in fracture toughness of SA217-WC9 steel was observed. It was found that the low toughness of a heat of SA217-WC9 steel was attributed to the coarse MnS inclusion originated by high sulfur content as the results of microanalyses.

• Journal of Energy Engineering
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• v.21 no.1
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• pp.75-80
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• 2012

In this paper a realistic bounding method for flooding analysis following rupture of large size of thanks and piping is defined. Mass and energy release during main feedwater line break accident is analyzed with RETRAN code. It is modeled that the main feed water control valve is closed in 5.0 seconds after reactor trip. In result of the analysis, largest mass and energy is discharged at 70% reactor power. The flood sources for main feedwater room are calculated when piping failure occurs in the high energy line and medium energy line. Based on the result of flood level (1.43m), it is investigated that all of the safety-related environmental qualification equipments are well located above the flood level.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.74-78
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• 2015

Nuclear power plant steam generator that is one of the main component has several thousands of thin tubes. And the steam generator tube is subject to damage because of the severe operation conditions such as the high temperature and pressure. Therefore periodic inspections are conducted to ensure the integrity of steam generator component. Hanul unit 3 also has been inspected in accordance with in-service inspection program and is scheduled to be replaced for exceeding the plugging rate which was recommended by manufacturer. During the steam generator replacement activity, we found several clustered porosity on inner surface of main feed water pipe. Additionally crack-like indications were found at weld interface between base material and weld of main feed water pipe. This paper describes the field experience and visual testing results for inner surface of main feed water pipes. The destructive test result had shown that these indications were porosities which were caused by manufacturing process not by operation service.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.04a
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• pp.648-654
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• 2011

The feed-water piping system constitutes a complex flow impedance network incorporating dynamic transfer characteristics which will amplify some pulsation frequencies. Understanding pressure pulsation waves for the feed-water recirculation piping system with cavitation problem of flow control valve is very important to prevent acoustic resonance. Feed water recirculation piping system is excited by potential sources of the shock pulse waves by cavitation of flow control valve. The pulsation becomes the source of structural vibration at the piping system. If it coincides with the natural frequency of the pipe system, excessive vibration results. High-level vibration due to the pressure pulsation affects the reliability of the plant piping system. This paper discusses the piping vibration due to the effect of shock pulsation by the cavitation of the flow control valves for the recirculation piping of feed-water pump system in combined cycle power plants.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.19 no.8
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• pp.843-849
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• 2009

The reliable operation of a swing check valve in the main feed water system of a power plant is most essential for successful shout-down process. A failure to close the valve at proper time often leads to the instability of the main feed water system, or even to an emergency stop of the power plant. In reality it is a very difficult task to monitor the behavior of a swing check valve. Furthermore it is impossible to see the motion of the valve. In this work two measurements were carried out simultaneously to determine the precise valve closure time. The dynamic pressure measurements were made at the inlet and outlet regions of the swing check valve. The transient vibration of the valve housing in the direction of water flow was also measured, which enabled the measurement of the transient vibration of the valve housing near valve closure. By comparing the results produced from these measurements the precise valve closure time could be determined. By carrying out order tracking technique using the dynamic pressure signals and pump rpm signal, the complicated dynamic problems inside the main feed water system can be more easily dealt with. This measurement scheme might be implemented in a power plant on a real-time basis without much difficulty. If this could be implemented, valuable information essential for shut-down operations can readily be passed on to the main control room. The feasibility of this implementation was demonstrated by this experimental work.

• Journal of the Korean Institute of Intelligent Systems
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• v.3 no.2
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• pp.18-35
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• 1993

The water level control system of the steam generator in a pressurized water reactor and its control problems are analysed. In this work the stable control strategy during the low power operation and transient states is studied. To solve the problem, a fuzzy logic control method is applied as a basic algorithm of the controller. The control algorithm is based on the operator's knowledges and the experiences of manual operation for water level control at the compact nuclear simulator set up in Korea Atomic Energy Research Institute. From a viewpoint of the system realization, the control variables and rules are established considering simpler tuning and the input-output relation. The control strategy includes the dynamic tuning method and employs a substitutional information using the bypass valve opening instead of incorrectly measured signal at the low flow rate as the fuzzy variable of the flow rate during the pressure control mode of the steam generator. It also involves the switching algorithm between the control valves to suppress the perturbation of water level. The simulation results show that both of the fine control action at the small level error and the quick response at the large level error can be obtained and that the performance of the controller is improved.