• 한국지진공학회논문집
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• 제26권1호
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• pp.39-48
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• 2022

Many piping systems installed in the power plant are directly related to the safety and operation of the plant. Various dampers have been applied to the piping system to reduce the damage caused by earthquakes. In order to reduce the vibration of the piping system, this study developed a steel coil damper (SCD) with a straightforward structure but excellent damping performance. SCD reduces the vibration of the objective structure by hysteretic damping. The new SCD damper can be applied to high-temperature environments since it consists of steel members. The paper introduces a design method for the elastoplastic coil spring, which is the critical element of SCD. The practical applicability of the design procedure was validated by comparing the nonlinear force-displacement curves calculated by design equations with the results obtained from nonlinear finite element analysis and repeated loading test. It was found that the designed SCD's have a damping ratio higher than 25%. In addition, this study performed a set of seismic tests using a shaking table with an existing piping system to verify the vibration control capacity on the piping system by SCD. Test results prove that the SCD can effectively control the displacement vibration of the piping system up to 80%.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1989년도 봄 학술발표회 논문집
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• pp.50-55
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• 1989

Piping systems are composed of pipes with various thickness, diameter and length. Accurate analysis of a piping system requires a complicated three dimensional finite element model and a computer system with large memory size, while simplified models result in system response prediction with deteriorated accuracy. An efficient analysis model for piping systems is proposed in this study. The proposed model is developed by introducing a joint model which accounts for the behavior of a pipe connection. Pipes are represented by beam elements and the effect of local deformation of pipe connections are replaced by joint element deformations. The proposed model which is as simple and efficient as a beam model can be used to obtain piping system response with accuracy close to that of a finite element model.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.807-812
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• 2001

This paper represents the wall thinning analysis results for secondary side piping of two types of domestic nuclear power plants based on the DB establishment and F AC analysis study for NPP secondary system piping. CHECWORKS code utilized in this study has been applied world widely to wall thinning analyses for secondary side piping and its reliability has also been proved. The predicted wear rates for several piping systems of a pressurized water reactor NPP are compared with those of a pressurized heavy water reactor NPP and with the measured wear rates. On the basis of comparison results of the predicted and measured wear rates, the analysis results can be effectively applied to the development of a standard thinned pipe management program targeted all domestic nuclear power plants.

• Journal of Mechanical Science and Technology
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• 제19권12호
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• pp.2245-2252
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• 2005

Thermal stratification in the branch piping of power plants can be generated by turbulent penetration or by valve leakage. In this study, a numerical analysis was performed to estimate the thermal stratification phenomenon by in-leakage in the SIS branch piping of nuclear power plant. Leakage rate, leakage area and leakage location were selected as evaluation factors to investigate the thermal stratification effect. As a result of the thermal stratification effect according to leakage rate, the maximum temperature difference between top and bottom of the horizontal piping was evaluated to be about 185K when the valve leakage rate was about 10 times as much as the allowed leakage rate. For leakage rate more than 10 times the allowed leakage rate, the temperature difference was rapidly decreased due to the increased mixing effect. In the result according to leakage area, the magnitude of temperature difference was shown in order of $3\%,\;1\%\;and\;5\%$ leakage area of the total disk area. In the thermal stratification effect, according to the leakage location, temperature difference when leakage occurred in the lower disk was considerably higher than that of when leakage occurred in the upper disk.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.617-621
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• 2003

The purpose of this paper is to analyze piping failure trend of safety pipings In domestic nuclear power plants. First, database for the piping failure was constructed with 105 data fields. The database includes plant population data, event data, and service history data. 7 kinds of piping failures in domestic NPPs were investigated. Among the 7 cases, detailed root causes were investigated for 3 cases. The first one is pipe wall thinning in main feedwater pipings of Westinghouse 3 loop type plants. The root cause of the wall thinning was flow accelerated corrosion near welding area. The next one is leak event in chemical and volume control system(CVCS) due to vibration. Some cracks occurred in socket welding area. The events showed that the integrity or socket weld is very vulnerable to vibration. The last one is also a leak event in primary sampling line in Korean standard reactor due to thermal fatigue. Although the structural integrity was not maintained by the events, there was no effect on nuclear safety in the above 3 piping failure eases.

• Nuclear Engineering and Technology
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• 제45권1호
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• pp.99-106
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• 2013

Thermal fatigue is a significant long-term degradation mechanism in nuclear power plants. In particular, as operating plants become older and life time extension activities are initiated, operators and regulators need screening criteria to exclude risks of thermal fatigue and methods to determine significant fatigue relevance. In general, the common thermal fatigue issues are well understood and controlled by plant instrumentation at fatigue susceptible locations. However, incidents indicate that certain piping system Tee connections are susceptible to turbulent temperature mixing effects that cannot be adequately monitored by common thermocouple instrumentations. Therefore, in this study thermal fatigue evaluation of piping system Tee-connections is performed using the fluid-structure interaction (FSI) analysis. From the thermal hydraulic analysis, the temperature distributions are determined and their results are applied to the structural model of the piping system to determine the thermal stress. Using the rain-flow method the fatigue analysis is performed to generate fatigue usage factors. The procedure for improved load thermal fatigue assessment using FSI analysis shown in this study will supply valuable information for establishing a methodology on thermal fatigue.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.43-49
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• 2010

The piping failure probability of the reactor coolant system in Kori unit 1 was evaluated considering stress corrosion cracking. The P-PIE program (Probabilistic Piping Integrity Evaluation Program) developed in this study was used in the analysis. The effect of some variables such as oxygen concentration during start up and steady state operation, and operating temperature, which are related with stress corrosion cracking, on the piping failure probabilities was investigated. The effects of leak detection capability, the size of big leak, piping loops, and reactor types on the piping failure probability were also investigated. The results show that (1) LOCA (loss of coolant accident) probability of Kori unit 1 is extremely low, (2) leak probability is sensitive to oxygen concentration during steady state operation and operating temperature, while not sensitive to the oxygen concentration during start up, and (3) the piping thickness and operating temperature play important roles in the leak probabilities of the cold leg in 4 reactor types having same inner diameter.

• 대한기계학회논문집
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• 제15권6호
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• pp.1861-1871
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• 1991

본 연구에서는 아직까지 연구가 미진한 내용 즉, 유속과 압력이 시간과 위치 의 함수인 유동특성과 파이프의 운동이 상호 연계되어 영향을 주는 일반적인 경우의 운동방정식을 유도하였고 단순지지된 직선 파이프를 모델로 설정하여 동적 안정성 (dynamic stability)과 진동응답을 수치적으로 고찰하였다.

• 한국기계가공학회지
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• 제13권1호
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• pp.23-29
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• 2014

The purpose of this paper is to assess the structural integrity of piping penetrations for nuclear power plants. A piping qualification analysis describes loads due to deadweight, pressure difference acts normal to the plate, thermal transients, and earthquakes, among other events, on piping penetrations that have been modeled as an anchor. Amodel was analyzed using a commercial finite element program. Apiping penetration analysis model was constructed with an assembly of pipe, head fittings and sleeves. Normally, the design load, thus obtained, will consist of three moments and three forces, referred to a Cartesian coordinate system. When comparing the stress analysis results from each required cutting position, the general membrane stress intensities and local membrane plus bending stress intensities during a structural evaluation cannot exceed the allowable amount of stress for the design loads. Therefore, the piping penetration design satisfies the code requirements.

• 한국주거학회논문집
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• 제19권5호
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• pp.111-120
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• 2008

The aim of this study is to present the design methods about manifold location being installed and size and to draw out the proper piping size as comparing the fluctuation of discharge with manifold size and residence size through the piping network analysis, when using the same faucet in accordance. The findings are summarized as follows, 1) an appropriate header main body pipe diameter was deemed to be $32{\sim}50\;mm$. 2) the research presented design measures for the application of appropriate water supply inlet pipe diameters according to residential buildings with various sizes. 3) the header direct branch piping method is ideal for small and medium-sized residential complexes, and the header branching and semi header methods are deemed to be more favorable for large residential complexes. 4) this study offered design measures for appropriate header system main body pipe diameters, water supply inlet pipe diameters, header system piping methods, application methods for functional auxiliary equipment units, and header system installation spaces and location.