• Nuclear Engineering and Technology
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• v.36 no.3
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• pp.210-218
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• 2004

In the interest of nuclear power plant safety, a self-diagnostic monitoring system (SDMS) is needed to monitor defects in safety-related components. An air-operated valve (AOV) is one of the components to be monitored since the failure of its operation could potentially have catastrophic consequences. In this paper, a model of the AOV is developed with the parameters that affect the operational characteristics. The model is useful for both understanding the operation and correlating parameters and defects. Various defects are introduced in the experiments to construct a fault library, which will be used in a pattern recognition approach. Finally, the validity of the fault library is examined.

• Nuclear Engineering and Technology
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• v.36 no.3
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• pp.219-228
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• 2004

[1] and [2] present an approach to diagnosing possible defects in the mechanical systems of a nuclear power plant. In this paper, by using a fault library as a database and training data, we develop a diagnostic algorithm 1) to decide whether an Air Operated Valve system is sound or not and 2) to identify the defect from which an Air-Operated Valve system suffers, if any. This algorithm is composed of three stages: a neural net stage, a non-neural net stage, and an integration stage. The neural net stage is a simple perceptron, a pattern-recognition module, using a neural net. The non-neural net stage is a simple pattern-matching algorithm, which translates the degree of matching into a corresponding number. The integration stage collects each output and makes a decision. We present a simulation result and confirm that the developed algorithm works accurately, if the input matches one in the database.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.419-422
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• 1995

The Atmospheric Dump Valve (ADV) and Turbine Bypass Valve (TBV) capacity test was performed at 30% power level during the YGN 3 Power Ascension Test period. In this test, several plant data were measured to calculate the ADV and TBV capacity considering that critical condition was developed through the ADV and TBV. The test results show that the test acceptance criteria are met.

• The Journal of the Acoustical Society of Korea
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• v.16 no.1E
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• pp.9-14
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• 1997

Check valves used in industrial and Nuclear Power Plant safety systems are susceptible to failure modes generally associated with wear of internal parts. Specifically, hinge pins, disc studs, pistons, and other mechanical parts may degrade over time, and in some cases, may which might produce a disabling event leading to plant or process shutdown. The primary diagnostic technique in the past has been to disassemble the valves. This procedure is costly, time consuming, and in the nuclear industry, it can lead to radiation exposure in some situations. Additionally repair and reassembly of a valve does not ensure proper operation. Non-intrusive diagnostic technologies including acoustics and magnetics with a digital signal analysis allow to evaluate check valve performance without a disassembly and is able to help the user detect degraded valve conditions.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.6
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• pp.683-691
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• 2011

In order to develop technology for an equipment qualification (EQ) test, which is an important process in localizing solenoid-operated valves used in nuclear power plants, we analyzed related regulations, test procedures, conditions, equipment, and acceptance criteria. EQ regulations for the solenoid-operated valve are classified as law, guide, and standard, and are subdivided according to test specimens and contents. The EQ test is composed of functional, normal-, and accident- condition tests. The solenoid-operated valve is aged under normal and accident conditions, which are predicted in the design conditions of a nuclear power plant, and the performance of the valve is measured by a functional test. The test method and procedure analyzed in this paper might be very useful for manufacturers as well as EQ testers.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.11
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• pp.1053-1060
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• 2013

The purpose of this study is to estimate the packing friction of power-operated valve in nuclear power plants. The roll of packing in valve is preventing leakage through stem. Packing friction is highly depend on gland nut tightness which means higher reliability in sealing is lower operability. For the estimation of friction, we used statistical analysis and experimental analysis. In experimental approach, we have performed packing fY test and applied it to valve field test. In statistical approach, we have used 10 years DB of safety-related valve in nuclear power plant and analyzed packing friction based on confidence interval of sample. The comparison of two results shows that statistical analysis for packing friction are more accurate than fY analysis even though both approach have error compared to measured value but we confirmed that statistical approach is proper way to estimate packing friction.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.110-114
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• 2001

A numerical analysis has been perfonned to estimate the effect of turbulent penetration and thermal stratified flow in the branch lines piping. This phenomenon of thermal stratification are usually observed in the piping lines of the safety related systems and may be identified as the source of fatigue in the piping system due to the thermal stress loading which are associated with plant operating modes. The turbulent penetration length reaches to $1^{st}$ valve in safety injection piping from reactor coolant system (RCS) at normal operation for nuclear power plant when a coolant does not leak out through valve. At the time, therefore, the thermal stratification does not appear in the piping between RCS piping and $1^{st}$ valve of safety injection piping. When a coolant leak out through the $1^{st}$ valve by any damage, however, the thermal stratification can occur in the safety injection piping. At that time, the maximum temperature difference of fluid between top and bottom in the piping is estimated about $50^{\circ}C$.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.851-852
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.178.1-178
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.559-560
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• 2004