• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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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 $50^{\circ}C$.

• 한국압력기기공학회 논문집
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• 제19권1호
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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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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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.

• 한국압력기기공학회 논문집
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• 제17권1호
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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.

• 한국안전학회지
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• 제17권2호
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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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• 제36권3호
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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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• 제54권10호
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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.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 1990년도 추계학술대회논문집; 한양대학교, 서울; 24 Nov. 1990
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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.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2008년도 춘계학술대회논문집
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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.

• 대한기계학회논문집B
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• 제37권11호
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• pp.1053-1060
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• 2013

본 연구의 목적은 원자력 발전소에서 사용되는 동력구동밸브의 패킹 마찰력 예측에 대한 적절한 방법을 찾는 것이다. 밸브의 패킹은 스템을 통한 누설을 방지하는 것이 목적이나 누설방지를 위하여 글랜드 너트 강하게 조여줄 경우 패킹의 마찰력으로 인하여 밸브 작동성이 떨어지게 된다. 본 연구에서 패킹 마찰력 예측을 위하여 실험적 방법과 통계적 방법을 사용하였다. 실험적 방법은 패킹 조합의 fY값 측정후 현장 밸브에 적용하였으며, 통계적 방법은 과거 10년간의 안전관련 밸브의 진단시험 데이터의 패킹 마찰력값을 표본으로 신뢰구간을 만들어 분석에 이용하였다. 두 방법 모두 실측값과는 상당한 차이를 보였으나 fY값 측정을 통한 실험적 방법 보다는 과거 진단시험 데이터를 이용한 예측이 좀 더 정확한 결과를 보였다.