• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.437-442
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• 1997

To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are peformed. The methodology is developed by evaluating "decrease in secondary heat removal" events such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling, The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. pressure.

• Journal of the Korean Society of Safety
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• v.25 no.2
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• pp.24-28
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• 2010

The nuclear power Plant onsite AC electrical power sources are required to supply power to the engineering safety facility buses if the offsite power source is lost. Typically, Diesel Generators are used as the onsite power source. The 125 VAC buses are part of the onsite Class 1E AC and DC electrical power distribution system. The DC power distribution system ensure the availability of DC electrical power for system required to shutdown the reactor and maintain it in a safety condition after an anticipated operational occurrence or a postulated Design Base Accident. Recently, onsite DC power supply system trip occurs the loss of system function. To obtain the performance such as reliability and availability, we analyzed the cause of battery charger trip and described the improvement of DC power supply system reliability. Finally, we provide reliability performance criteria of charger in order to ensure the probabilistic goals for the safety of the nuclear power plants.

• Journal of Korea Society of Digital Industry and Information Management
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• v.19 no.3
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• pp.25-38
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• 2023

Instrumentation and control systems measure and control various variables of nuclear facilities to operate nuclear power plants safely. A diverse protection system, a representative instrumentation and control system, generates a reactor trip and turbine trip signal by high pressure in a pressurizer and containment to satisfy the design requirements 10CFR50.62. Also, it generates an auxiliary feedwater actuation signal by low water levels in steam generators. Cybersecurity has become more critical as digital technology is gradually applied to solve problems such as performance degradation due to aging of analog equipment, increased maintenance costs, and product discontinuation. This paper analyzed possible cybersecurity threat scenarios in the diverse protection system using attack trees. Based on the analyzed cybersecurity threat scenario, we calculated the probability of attack occurrence and confirmed the cybersecurity risk in connection with the asset value.

• 한국신재생에너지학회:학술대회논문집
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• 2008.05a
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• pp.364-367
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• 2008

This present paper describes concepts of the real-time operation monitoring system as a tool for enhancing the reliability and raising the availability of the first Korea IGCC (Integrated Gasification Combined Cycle) power plant. This system consists of five (5) modules : (1) Data Validation Module, (2) Performance Calculation Module (3) Performance Diagnostic Module, (4) Trip Information Module, and (5) Statistics Analysis Module. Among these modules, Performance Calculation Module is explained in more detail. The objective of this module is to continuously evaluate the degradation (decrease in performance) of the IGCC plant and its equipment in order to provide plant operators additional information to help them identify problems, improve performance.

• Proceedings of the KIPE Conference
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• 2007.07a
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• pp.479-481
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• 2007

This paper presented a actual conditions on a false-tripping of ELB(Earth Leakage Breakers) and an assessment of the harmonics due to non-linear electronic equipment such as personal computers, rectifier, electric furnace and test equipments in plant installations. Actual conditions on a failure of ELB were surveyed and analyzed by making up a question to safety managers of plant installations. Also, power quality measurements such as harmonic currents, voltage and current waveform and current imbalance were made at various plant installations where nuisance tripping of ELB is often occurred. The results of this study can be used in making decisions regarding cause of ELB trip and operating and tripping characteristics test of ELB when subjected to distorted current waveform.

• Proceedings of the KIEE Conference
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• 1996.07b
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• pp.1275-1277
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• 1996

The Sequence Of Event (SOE) system used in nuclear power plants is a part of the Plant Data Acquisition System (PDAS). The SOE system of the existing nuclear power plant share the computer H/W and S/W, and required more complicated structure to provide the events or trip signals. Moreover, there are high potential of collision between synchronization signals and data transmitted to the Plant Computer System(PCS) when the synchronization signals are sent from PCS to the three SOE processors. When this collision happens the SOE system will break down, thus it is not possible to analyze the trend of events or trips. This paper issued the limitation items of the existing SOE system and proposed the revised SOE system.

• Plant Journal
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• v.11 no.1
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• pp.50-57
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• 2015

This study presents prediction of maximum fly ash conveying capacity of fly ash system in a power plant. The mixture ratio and pressure drop characteristics of air-fly ash flow in piping system are not well understood due to the complexity of particle motion mechanism. In this paper, the researcher investigated the optimum mixture ratio when the pressure drop of fly ash conveying system is equal to maximum static pressure of displacement fly ash transport blower and the capacity of fly ash transport according to the optimum mixture ratio by experimenting the fly ash conveying system of domestic D coal thermal power plants, which is currently in operation. The experiment results showed that the maximum fly ash conveying capacity of fly ash system were founded under the condition of maximum air volume 5,040 m3/h, static pressure of trip condition 1,163 mmH2O. In addition, it was predicted maximum mixture ratio of the air-fly ash was 8.66 and maximum capacity of fly ash conveying was 52,600 kg/h under these conditions.

• Proceedings of the KIEE Conference
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• 2001.07b
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• pp.1088-1090
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• 2001

In general, Power SCR(Silicon Controlled Rectifier) is most widely used in Power Plant as well as Industrial field. It has been controlled and operated according to its own control method. Especially, in case of Power plant, it plays a major role in AVR(Automatic Voltage Regulator) or electro chlorination control circuits. Generally, they used in Analog control system at above field. But each SCR current value is different because of load unbalance or switching characteristic variations, it may cause power plant unit trip or system disorder according to SCR element burn out or bad operating condition. Therefore, in this paper a development of 3 phase current balance control unit is described, it gets over the past analog control system limit, uses DSP(Digital signal processor) had high speed response, controls SCR gate firing angle for 3 phase current balance.

• 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.

• Journal of Power System Engineering
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• v.12 no.5
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• pp.65-70
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• 2008

The purpose of this study is to verify availability of the acoustic emission in-situ monitoring method to the internal leak and operating conditions of the turbine major valves relating to safety for turbine operating and prevention of turbine trouble at nuclear power plants. In this study, acoustic emission tests are performed when the pressurized electro-hydraulic control oil flowed through turbine electro-hydraulic controller oil check valve and turbine power/trip fluid solenoid valve in the condition of actual turbine operating. The acoustic emission method was applied to the valves at the site, and the background noise was measured far the abnormal plant condition. To judge for the leak existence ell the object valves, voltage analysis and frequency analysis of acoustic signal emitted from infernal leak in the valve operating condition are performed. It was conformed that acoustic emission method could monitor for valve internal leak to high sensitivity.