• 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 $50^{\circ}C$.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.52-59
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• 2023

The welding part of the valve needs immediate action when leakage occurs due to cracks or damage. In order to repair leakage of the welding part, the valve must be separated from the pipe or replaced with a new valve. However, it is difficult to remove the valve while operating the power plant. This study presents a method to remove leakage by precisely processing the gap between the clamp and the incision part within 0.1mm while installed in the pipe system. If the external leakage is removed using a clamp on the welding part without removing the valve during operation, the time and cost required for maintenance can be reduced.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.903-908
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• 2001

A web-based client/server program, MOVIDIK(Motor-Operated-Valve Integrated Database Information of KEPCO) has been developed to perform a design basis safety evaluation for a motor-operated-valve(MOV) in the nuclear power plant. The MOVIDIK consists of seven analysis modules and one administrative module. The analysis module calculates a differential pressure on the valve disk, thrust/torque acting at a valve stem, maximum allowable stress, thermal-overload-relay selection, voltage degradation, actuator output and margin. In addition, the administrative module manages user information, approval system and code information. MOVIDIK controls a huge amount of evaluation data and piles up the safety information of safety-related MOV. The MOVIDIK will improve the efficiency of safety evaluation work and standardize the analysis process for the MOV.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.8-17
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• 2021

The seat leakage test is required after the maintenance work on the valve seat. Either the test has been performed outside of the plant after cutting the valve from the pipe system or the simplified test has been performed so far. It was unable to perform the test at the plant site because it is hard to make a steady pressure on the valve inlet when it is installed in the pipe. This research aims to perform the leakage test in the nuclear power plant while it is installed in the pipe system. The mock-up test is performed by pressurizing the leak-off pipe on the valve body. The result is compared with traditional test result by pressurizing the valve inlet. Furthermore the chamber mock-up tests are performed under various conditions. The leak rate by the developed test using the leak-off pipe is found to be similar but greater than the leak rate by the existing test method. It implies that the test using the leak-off pipe is more conservative than the existing test. The methodology and the equipment which this paper suggests that on-site seat test is possible and the application of the technology could reduce the time and cost for the valve maintenance work significantly.

• Journal of the Korean Society of Safety
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• v.17 no.2
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• pp.92-99
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• 2002

Safety-related motor operated valve(MOV) safety significance for Ulchin Unit 3 was categorized. The safety evaluation of MOV of domestic nuclear power plants affects the generic data used for the quantification of MOV common cause failure(CCF) events in Ulchin Units 3&4 PSA. Therefore, in this paper, MGL(multiple greek letter)parameter ${\beta}$, used for the evaluation of MOV CCF probabilities in Ulchin Units 3&4 probabilistic safety assessment(PSA), was re-estimated and the MOV safety significance was categorized. The re-estimation results of MGL parameter show that the value of(is decreased by 30% compared with the current value used in Ulchin Unit 3&4 PSA. The categorization results of MOV safety significance using the changed value of MGL parameter(show that the number of HSSCs(high safety significant components) is decreased by 54.5% compared with those using the current value of it used in Ulchin Units 3&4 PSA.

• 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.54 no.10
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• pp.3745-3765
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• 2022

The concrete structures related to nuclear safety are threatened by accidental impact loadings, mainly including the low-velocity drop-weight impact (e.g., spent fuel cask and assembly, etc. with the velocity less than 20 m/s) and high-speed projectile impact (e.g., steel pipe, valve, turbine bucket, etc. with the velocity higher than 20 m/s), while the existing studies are still limited in the impact resistant design of nuclear power plant (NPP), especially the primary RC slab. This paper aims to propose the numerical simulation and theoretical approaches to assist the impact-resistant design of RC slab in NPP. Firstly, the continuous surface cap (CSC) model parameters for concrete with the compressive strength of 20-70 MPa are fully calibrated and verified, and the refined numerical simulation approach is proposed. Secondly, the two-degree freedom (TDOF) model with considering the mutual effect of flexural and shear resistance of RC slab are developed. Furthermore, based on the low-velocity drop hammer tests and high-speed soft/hard projectile impact tests on RC slabs, the adopted numerical simulation and TDOF model approaches are fully validated by the flexural and punching shear damage, deflection, and impact force time-histories of RC slabs. Finally, as for the two low-velocity impact scenarios, the design procedure of RC slab based on TDOF model is validated and recommended. Meanwhile, as for the four actual high-speed impact scenarios, the impact-resistant design specification in Chinese code NB/T 20012-2019 is evaluated, the over conservation of which is found, and the proposed numerical approach is recommended. The present work could beneficially guide the impact-resistant design and safety assessment of NPPs against the accidental impact loadings.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1990.10a
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• pp.113-124
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• 1990

Tap lines are small branch piping generally less than two inches in diameter. They typically branch off of header piping having a much larger diameter. An example of a common tap line is a 3/4 inch size high point vent or low point drain. Most tap lines have at least one valve near the header tap connection to provide isolation. Two valves are often required for double isolation. A light water reactor(LWR) nuclear power plant will have several hundred tap lines. These lines come in many sizes and shapes and serve numerous functions. A single process piping valve may have three different tap lines associated with it (figure 1). Table 1 delineates the different categories of tap lines. Vibration failures of tap lines are a common occurrence in all industrial plants including nuclear and fossil power plants. These types of failures constitute a significant percentage of all piping related failures. An unscheduled plant shutdown or outage resulting from the failure of a tap line decreases plant reliability and may have a detrimental effect on plant safety. Most tap line vibration failures can be avoided through the use of appropriate routing and support techniques. Standardized designs can be developed for use in a myriad of applications. These designs will not only minimize failures but will also reduce the necessary analysis and installation efforts.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2008.04a
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• pp.416-419
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• 2008

In order to ensure the safety, the performance evaluation of the safety-related components in a nuclear power plant such as air-operated valves. In this paper, the diagnostic system(MOVIDS $A^+$) for the performance of air-operated valves was developed. For this purpose, the characteristics of their operation and the methods of the diagnostic tests were reviewed. The setup and diagnostic functions of the system were mentioned. Its applicability was validated through the diagnostic tests of air-operated valves in nuclear power plants. This diagnostic system is now applied in nuclear power plants for performance tests.

• 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.