• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.265-273
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• 1996

Among those tests performed during the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3&4) Power Ascension Test period, the Loss of a Main Feedwater Pump test at l00% power is one of the major test which characterize the capability of YGN 3&4. In this event, one of the two normally operating main feedwater pumps is tripped resulting in a 50% reduction in the feedwater flow. Unless the NSSS and Turbine/Generator control systems actuate properly, the reactor will be tripped on low SG water level or high pressurizer pressure. The test performed at Unit 3 was successful by meeting all acceptance criteria, and the plant was stabilized at a reduced power level without reactor trip. The measured test data for the major plant parameters are compared with the predictions made by the KISPAC computer code, an updated best-estimate plant performance analysis code, to verify and validate its applicability. The comparison results showed good agreement in the magnitude as well as the trends of the major plant parameters. Therefore, the KISPAC code can be utilized for the best-estimate nuclear power plant design and simulation tool after a further verification using other plant test data.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.291-292
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.296-302
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• 1996

The Loss of a Main Feedwater Pump test at 100% Power for YGN 4 was simulated in order to verify and validate the KSNPA. The comparison of the test data with the KSNPA prediction results showed reasonable agreement in the trends of the major plant parameters. All plant control systems including NSSS and T/G control systems are properly actuated and stabilized the plant conditions to a new steady state conditions in the KSNPA. From the comparison results, the KSNPA showed its capability to simulate the LOMFP event for the Korean Standard Nuclear Power Plant.

• Transactions of the Korean Society of Mechanical Engineers
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• v.16 no.4
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• pp.753-764
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• 1992

The steam generator water level is difficult to control at low power due to its reversed responses to the feedwater flow, which are well known as the shrink and swell phenomena. With regard to this problem a new control scheme has been studied by which the level transients could be kept within permissible ranges at low power. The relations between the various input conditions to steam generator and the level transients have been examined to be expressed in the form of process transfer functions. Analog filters have been incorporated to be expressed in the process with proper control constants. This control scheme allows the prediction of level variation together with the corresponding feedwater rate and results in mider transients with good stabilites.

• 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.

• Nuclear Engineering and Technology
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• v.32 no.2
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• pp.157-168
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• 2000

The nuclear steam generator level control system is designed by robust control methods. The feedwater controller is designed by three methods of the H$\infty$, the mixed weight sensitivity and the structured singular value. Then the controller located on the feedback loop of the level control system is designed. For the system performance, the controller of simple PID whose coefficients vary with the power is selected. The simulations show that the system has a good performance with proper stability margins.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.21 no.8
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• pp.19-24
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• 2022

We developed a prototype from the design of a trim, which is the most important in the localization development of a 1500 Ib high-differential pressure-control valve used for boiler feedwater, and measured the flow coefficient, the most basic design data for valves. The following conclusions were drawn. The comparison of the design values of the flow coefficients for the existing X-trim and the multicore trim designed for localization development showed that they were almost identical, and the X-trim value was slightly lower. The comparison of the X-trim and multicore trim based on the valve flow coefficient test showed that they were generally similar, indicating no problem with the design. In the future, we plan to compare and analyze the flow paths for the X-trim and multicore trim via flow analysis.

• 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.

• 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.