• 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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• 2011.07a
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• pp.1810-1811
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• 2011

화력발전소의 증기발생기는 급수, 공기, 연료를 적절히 제어하여 터빈에서 필요로 하는 증기를 생성하고, 발전기 출력에 필요한 터빈속도는 증기발생기에서 발생한 증기유량을 제어하여 조절한다. 본 논문에서는 보일러 주제어기로부터 신호를 입력받아 급수 주제어기를 생성하는 알고리즘에 대하여 기술하였다. 이 알고리즘은 발전소의 다른 제어 알고리즘과 함께 기 운용중인 500MW급 석탄화력발전소 시뮬레이터에서 성능을 검증하고 기능을 보완하여 국내에서 개발 중인 분산제어 시스템에 설치하여 실제 발전소에 적용할 예정이다.

• Proceedings of the KIEE Conference
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• 2011.07a
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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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.

• Proceedings of KOSOMES biannual meeting
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• 2009.06a
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• pp.177-178
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• 2009

Hydraulic Transients would be occurred since pressure is increased or decreased when water speed inside of pipeline is rapidly changed A study on water hammer has become more important because the pumping stations were big and the systems conveying the fluid through the large and long transmission pipelines were complex. In this study, the method of characteristic line was adopted to evaluate the valve-induced water hammer phenomena in a pumps feedwater system.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.420-424
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• 1994

This paper presents an improved water level control scheme for Pressurized Water Reactor(PWR) Steam Generator(S/G) at the low power operation and transient states. To reduce fluctuations of the water level by the swell and shrink phenomena, the scheme adds feedforward terms considering S/G pressure and the feedwater temperature into the conventional proportional-integral feedback controller. The simulation results using the Compact Nuclear Simulator show that smaller level errors and much faster settling time than those of the conventional scheme can be obtained. The proposed algorithm is easily implementable and has a potential for the real applications.

• The KSFM Journal of Fluid Machinery
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• v.13 no.3
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• pp.12-17
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• 2010

The orifice establishment which is improper does to change the entity differential pressure and occurs an error in flow measurement data. Because of this, the thermal power of nuclear power plant could be evaluated excessively and the safety margin could be decreased. In this paper, characters of orifice which is established abnormally was investigated. Specially, the orifice plate which is established in opposition case was modeled and analyzed. Finally, 14.4% was lowly measured differential pressure, when being established in the resultant opposition. And this result with EPRI and NRC experiences was similar.

• 제어로봇시스템학회:학술대회논문집
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• 2002.10a
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• pp.101.3-101
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• 2002

Contents 1 Abstract- In the thermal power plant, there are six manipulated variables; main steam flow, feedwater flow, air flow, spray flow, fuel flow, and gas recirculation flow. Therefore, the thermal power plant control system is a multi-input and output system. In the control system, the main steam temperature typically is regulated by the fuel flow rate and the spray flow rate, and the reheater steam temperature is regulated by the gas recirculation flow rate. Up to the present time, the PID controller has been used to operate this system. This paper focuses on the characteristic comparison of the PID controller, the modified 2-DOF PID Controller on the DCS, in order to design an optimal...

• Proceedings of the KSME Conference
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• 2003.04a
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.213-218
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• 1996

A simulation model for the transient behavior of CANDU U-tube steam generator(UTSG) has been developed for application to the simulation of operational transient behavior of CANDU nuclear power plant. For application to CANDU UTSG. tile design characteristics of CANDU UTSG such as Wolsong Units, feedwater inlet near the tube sheet. is approximated. For realistic prediction of thermal hydraulic behavior of and tube bundle region is divided into two separate control volumes, subcooled region and saturated region. and the variation of thermal hydraulic properties within a control volume is considered. Test results for typical CANDU operational transient case show reasonable transient behavior of steam generator and considered to be applicable to the simulation of overall plant.