• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.911-918
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• 2012

Steam generator level control at low power is difficult due to its adverse thermal hydraulic properties, and is usually conducted by an operator. The basic model predictive control (MPC) is similar to the action of an operator in that the operator knows the desired reference trajectory for a finite period of time and takes the necessary control actions needed to ensure the desired trajectory. An MPC is based on a model; the performance as well as the efficiency of the MPC depends heavily on the exactness of the model. In this study, steam generator models that can describe in detail its thermal hydraulic behaviors, particularly at low power, are used in the MPC design. The design scope is divided into two parts. First, the MPC feedwater controller of the feedwater station is determined, and then the MPC level controller for the overall system is designed. Because the dynamic properties of a steam generator change with the power levels, a realistic situation is simulated by changing the transfer functions of the steam generator at every time step. The resulting MPC controller shows good performance.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.597-610
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• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

• Journal of the Korean Institute of Telematics and Electronics B
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• v.31B no.10
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• pp.53-66
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• 1994

In this paper, a fuzzy expert system is developed for fault diagnoisis of a drum-type boiler system in fossil power plants. The develped fuzzy espert system is composed of knowledge base, fuzzification module, knowledge base process module, knowledge base management module, inference module, and linguistic approximation module. The main objective of the fuzzy expert system is to check the states of the system including the drum level and detect faults such as the feedwater flow sensor fault, feedwater flow control valve fault, and water wall bube rupture. The fuzzy expert system diagnoses faults using process values, manipulated values, and knowledge base which is built via interviews and questionaries with the experts on the plant operations. Finally, the validity of the developed fuzzy expert system is shown via experiments using the digital simulator for boiler system is Seoul Power Plant Unit 4.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2004.05a
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• pp.975-978
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• 2004

There was transient vibration on the piping system from #4 heater to the deaerator in a power plant. We found it was resulted from resonance between the natural vibration of the piping system and vibration induced by flow of feedwater. We verified it would reduce vibration by increasing stiffness of the piping system. Therefore we concluded that it would be generally better to increase stiffness of the piping system to reduce vibration amplitude of 10Hz low for big sized piping systems.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.188.1-188
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.175.2-175
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• 1999

• 제어로봇시스템학회:학술대회논문집
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• 2000.10a
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• pp.37-37
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• 2000

This paper introduces the Performance Monitoring System(PMS) in a thermal power plant. The purpose of the PMS is to offer the operator current performance information of plant which could be an index of plant status or information to improve plant efficiency. The PMS of Bukcheju thermal power plant unit ＃2&3 is implemented under the SIEMENS DCS which supplies about 150 function blocks for performance calculation and all measured signals. The performance of unit, boiler, turbines, feedwater heaters, condenser, airpreheaters, feedwater pumps will be monitored and updated for every 5 minutes in PMS of Bukcheju TPP.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2017-2022
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• 2004

There are multistage preheaters in the power generation plan to improve the thermal efficiency of the plant and to prevent the components from the thermal shock. The energy source of these heaters comes from the extracted two phase fluid of working system. These two-phase fluid can cause the so-called Flow Accelerated Corrosion(FAC) in the extracting piping and the bubble plate of the heater for example, in case of point Beach Nuclear Power Plant and in the Wolsung Nuclear Power Plant. The FAC is due to the mass transport of the thin oxide layer by the convection. FAC is dependent on many parameters such as the operation temperature, void fraction, the fluid velocity and pH of fluid and so on. Therefore, in this paper velocity was calculated by FLUENT code in order to find out the root cause of the wall thinning of the feedwater heaters. It also includeed in the fluid mixing analysis model are around the number 5A feedwater heater shell including the extraction pipeline. To identify the relation between the local velocities and wall thinning, the local velocities according to the analysis results were compared with distribution of the shell wall thicknes by ultrasonic test.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• 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.