• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.546-547
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• 2005

• The Transactions of The Korean Institute of Electrical Engineers
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• v.65 no.5
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• pp.836-842
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• 2016

When an accident occurs in the nuclear power plant, the faulted information might mislead to the high possibility of aggravating the accident. At the Fukushima accident, the operators misunderstood that there was no core exposure despite in the processing of core damage, because the instrument information of the reactor water level was provided to the operators optimistically other than the actual situation. Thus, this misunderstanding actually caused to much confusions on the rapid countermeasure on the accident, and then resulted in multiplying the accident propagation. It is necessary to be equipped with the function that informs operators the status of instrument integrity in real time. If plant operators verify that the instruments are working properly during accident conditions, they are able to make a decision more safely. In this study, we have performed various tests for the fault detection sensitivity of an data-driven empirical model to review the usability of the model in the accident conditions. The test was performed by using simulation data from the compact nuclear simulator that is numerically simulated to PWR type nuclear power plant. As a result of the test, the proposed model has shown good performance for detecting the specified instrument faults during normal plant conditions. Although the instrument fault detection sensitivity during plant accident conditions is lower than that during normal condition, the data-drive empirical model can be detected an instrument fault during early stage of plant accidents.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.351.2-351
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• 1999

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.654-659
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• 2020

The reliability of the spent fuel pool instrument is very important for the security of nuclear power plant, especially during the earthquake. The effect of the fluid force on the vibration characteristics of the flexible tube of the spent fuel pool instrument needs comprehensive analysis. In this paper, based on the potential flow theory, the hydrodynamic pressures acting on the flexible tube were obtained. A mathematical model of a flexible tube was constructed to obtain the dynamic response considering the effects of seismic load and fluid force, and a computer code was written. Based on the mathematical model and computer code, the maximum stresses of the flexible tube in both safe shutdown earthquake and operating basis earthquake events on the spent fuel pool with three typical water levels were calculated, respectively. The results show that the fluid force has an obvious effect on the stress and strain of the flexible tube in both safe shutdown earthquake and operating basis earthquake events.

• Journal of Energy Engineering
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• v.23 no.3
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• pp.96-101
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• 2014

The Instrument Channel setpoints of the Reactor Protection System(RPS) and the Engineered Safety Feature Actuation System(ESFAS) ensures the safety of Nuclear Power Plants (NPPs), and the actuation of the protection system should be guaranteed on power change condition. The goal of this study is to verify the appropriateness of the sensor drift and rack drift which are important factors for setpoints evaluation and to improve the setpoints margin using the operation data, design specifications and operation manuals of the NPPS.

• The Transactions of the Korean Institute of Electrical Engineers P
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• v.63 no.2
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• pp.61-68
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• 2014

Function of battery bank stores energy for DC load in general, and DC power system of the nuclear power plant is used to supply DC loads for safety- featured instrumentation and control such as inverter, class 1E power system control and indication, and station annunciation. Class 1E DC power system must provide a power for the design basis accident conditions, and adequate capacity must be available during loss of AC power and subsequent safe shutdown of the plant. In present, batteries of Class 1E DC power system of the nuclear power plant uses lead-acid batteries. Class 1E batteries of nuclear power plants in Korea are summarized in terms of specification, such as capacity, discharge rate, bank configuration and discharge end voltage, etc. This paper summarizes standards of determining battery size for the nuclear power plant, and analyzes duty cycle for the class 1E DC power system of nuclear power plant. Then, battery cell size is calculated as 2613Ah according to the standard. In addition, this paper analyzes performance test results during past 13 years and shows performance degradation in the battery bank. Performance tests in 2001 and 2005 represent that entire battery cells do not reach the discharge-end voltage. Howeyer, the discharge-end voltage is reached in 14.7% of channel A (17 EA), 13.8% of channel B (16 EA), 5.2% of channel C (6 EA) and 16.4% of channel D (19 EA) at 2011 performance test. Based on the performance test results analysis and size calculation, battery capacity and degradation by age in Korearn nuclear power plant is discussed and would be used for new design.

• Korean Journal of Computational Design and Engineering
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• v.12 no.1
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• pp.74-85
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• 2007

During the lifecycle of a nuclear power plant many organizations are involved in KOREA. Korea Plant Engineering Co. (KOPEC) participates in the design stage, Korea Hydraulic and Nuclear Power (KHNP) operates and manages all nuclear power plants in KOREA, Dusan Heavy Industries manufactures the main equipment, and a construction company constructs the plant. Even though each organization has a digital data management system inside and obtains a certain level of automation, data sharing among organizations is poor. KHNP gets drawing and technical specifications from KOPEC in the form of paper. It results in manual re-work of definition and there are potential errors in the process. A data warehouse based on a neutral model has been constructed in order to make an information bridge between design and O&M phases. GPM(generic product model), a data model from Hitachi, Japan is addressed and extended in this study. GPM has a similar architecture with ISO 15926 "life cycle data for process plant". The extension is oriented to nuclear power plants. This paper introduces some of implementation results: 1) 2D piping and instrument diagram (P&ID) and 3D CAD model exchanges and their visualization; 2) Interface between GPM-based data warehouse and KHNP ERP system.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.430-436
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• 1993

Due to the failure of the instrument and control devices of nuclear power plants caused by aging, nuclear power plants occasionally trip. Even a trip of a single nuclear power plant (NPP) causes an extravagant economical loss and deteriorates public acceptance of nuclear power plants. Therefore, the replacement of the instrument and control devices with proper consideration of the aging effect is necessary in order to prevent the inadvertent trip. In this paper we investigated the optimal replacement periods of the control computer's components of Wolsung nuclear power plant Unit 1. We first derived mathematical models for optimal replacement periods to the digital control computer's components of Wolsung NPP Unit 1 and calculated the optimal replacement periods analytically. We compared the periods with the replacement periods currently used at Wolsung NPP Unit 1. The periods used at Wolsung is not based on mathematical analysis, but on empirical knowledge. As a consequence, the optimal replacement periods analytically obtained and those used in the field show a little difference.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.1095-1096
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• 2005

• Journal of the Korean Society of Safety
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• v.26 no.2
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• pp.20-25
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• 2011

The fire event occurred in fire proof zone often causes serious electrical problems such as shorts, ground faults, or open circuits in nuclear power plants. These would be directed to the loss of safe shutdown capabilities performed by safety related systems and equipments. The fire event can treat the basic design principle that safety systems should keep their functions with redundancy and independency. In case of a multi-core cable fire, operators can not perform their mission properly and can misjudge the situation because of spurious operation, wrong indication or instrument. These would deteriorate the plant capabilities of safety shutdown and make disastrous conditions. In this paper, the characteristic of cable fire is investigated and the heat-flux evaluation for cable fire is studied. Moreover, a diagnostic methodology for degraded cable in nuclear power plants is presented.