• Proceedings of the KIEE Conference
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• 1999.07a
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• pp.403-405
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• 1999

Dissoved oxygen(DO) control by deaerator has a great effect on the integrity of S/G in nuclear power plant. The goal of this study based on the theoretical basis and the extensive surveys is to identify the effect of deaerator in feedwater system on steam generator to clear the need of installation of deaerator. In addition, this paper discusses the review to understand the mechanism of DO formation as well as removal. The conclusion is that the installation of deaerator improve the integrity of S/G and is contributed to the whole nuclear power plant safety.

• Proceedings of the KIEE Conference
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• 1991.11a
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• pp.349-352
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• 1991

Fuzzy modelling is a useful method when the variation of plant dynamics is large. In the fuzzy modelling by parameter identification, a new method is proposed in the part of premise parameters identification and in expanding MISO system into MIMO system. Using the proposed method, a fuzzy model of the drum boiler of the thermal power plant can be derived. In addition, feedwater control of the drum by fuzzy controller using the fuzzy model, is simulated.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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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.

• 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.

• Proceedings of the KIEE Conference
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• 2006.07d
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• pp.1885-1886
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• 2006

The thermal energy from nuclear fission is transferred to the steam generator which is a kind of a large heat exchanger. After the feedwater is injected into the steam generator and absorbs the thermal energy, it is converted into the steam. This steam goes into the turbine. The balance between the generated energy and the consumed energy is required for the nuclear power plant to be stable. For the purpose of which, the feed water, a parameter for energy transfer, should be controlled in stability. Usually, the nuclear power plants are operated in base load in the view of power system for the stability of fission system. Therefore, though there will be almost no unbalance, there can be some instability from unbalance in case of startup/shutdown or disturbance. In this case, the controllability of feedwater pump is very important for the quick recover of stability.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2004.05a
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• pp.490-494
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• 2004

The purpose of this paper was to identify the mechanism that caused abnormal vibration and noise on the piping system connected to discharge flow of BFP(Boiler Feed water Pump) in a coal fired power plant, and to develop the device that can reduce the level of abnormal vibration and noise. Major results of this project can be summarized as follows: First, we analyzed the acoustic mode for the discharge piping of BFP to trace a path of the noise, and assumed that noise and vibration on the piping system can be related with length of pipe. Second, a minimized model of the piping system was set up to simulate abnormal vibration and noise within the specific range of operating frequencies, and as a result we confirmed that the acoustic mode affected the piping system considerably. Finally the test device which can reduce the level of abnormal noise and vibration was built to verify validity applying for the piping system. Then we concluded that the noise and vibration generated from the piping system was attributed to the acoustic resonance in piping system, and so developed new device which can reduce the level of noise and vibration under 40%. Put Abstract here.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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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.

• Corrosion Science and Technology
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• v.15 no.4
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• pp.182-188
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• 2016

Significant piping wall thinning caused by Flow-Accelerated Corrosion (FAC) and Erosion-Corrosion (EC) continues to occur, even after the Mihama Power Station unit 3 secondary pipe rupture in 2004, in which workers were seriously injured or died. Nuclear power plants in many countries have experienced FAC and EC-related cases in steam cycle piping systems. Korea has also experienced piping wall thinning cases including thinning in the downstream straight pipe of a check valve in a feedwater pump line, the downstream elbow of a control valve in a feedwater flow control line, and failure of the straight pipe downstream of an orifice in an auxiliary steam return line. Cause analyses were performed by reviewing thickness data using Ultrasonic Techniques (UT) and, Scanning Electron Microscope (SEM) images for the failed pipe, and numerical simulation results for FAC and EC cases in Korea Nuclear Power Plants. It was concluded that the main cause of wall thinning for the downstream pipe of a check valve is FAC caused by water vortex flow due to the internal flow shape of a check valve, the main cause of wall thinning for the downstream elbow of a control valve is FAC caused by a thickness difference with the upstream pipe, and the main cause of wall thinning for the downstream pipe of an orifice is FAC and EC caused by liquid droplets and vortex flow. In order to investigate more cases, additional analyses were performed with the review of a lot of thickness data for inspected pipes. The results showed that pipe wall thinning was also affected by the operating condition of upstream equipment. Management of FAC and EC based on these cases will focus on the downstream piping of abnormal or unusual operated equipment.

• Journal of Institute of Control, Robotics and Systems
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• v.5 no.3
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• pp.363-369
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• 1999

In this paper, we design the robust $H_{\infty}$ controller for water level control of steam generator using a mixed $H_{\infty}$ optimization with model-matching method. Firstly we choose the desired model which has good disturbance rejection performance. Secondly we design a stabilizing controller to keep the model-matching error small and also provide sufficiently large stability margin against additive perturbations of the nominal plant. Simulation results show that proposed robust $H_{\infty}$ controller at specific power operation has satisfactory performances against the variations of load power, steam flow rate, primary circuit coolant temperature, and feedwater temperature. It can be also observed that the proposed robust $H_{\infty}$ controller exhibits better robust stability than conventional PI controller.

• Proceedings of the KIEE Conference
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• 2006.07d
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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.