• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.1633-1638
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• 2003

In case that in-leakage through the valve disk occurs, a numerical study is performed to estimate on thermal stratification phenomenon in the Safety Injection piping connected with the Reactor Coolant System piping of Nuclear Power Plant. As the leakage flow rate increases, the temperature difference between top and bottom of horizontal piping has the inflection point. In the connection point of valve and piping, the maximum temperature difference between top and bottom was 185K and occurred in the condition of 10 times of standard leakage flow rate. In the connection point of elbow and horizontal piping, the maximum temperature difference was 145K and occurred in the condition of 15 times of standard leakage flow rate. In the vertical piping of Safety Injection piping, the near of connection point between elbow and vertical piping showed the outstanding thermal stratification phenomenon in comparison with another region because of turbulent penetration from Reactor Coolant System piping. In order to prevent damage of piping due to the thermal stratification when in-leakage through the valve disk occurs, the connection points between valve and piping, and the connection points between elbow and piping need to be inspected continually.

• International Journal of Quality Innovation
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• 제7권2호
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• pp.97-124
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• 2006

The purpose of this research is to create an expert risk-based piping system inspection model. The proposed system includes a risk-based piping inspection system and a piping inspection guideline system. The research procedure consists of three parts: the risk-based inspection model, the risk-based piping inspection model, and the piping inspection guideline system model. In this research procedure, a field plant visit is conducted to collect the related domestic information (Taiwan) and foreign standards and regulations for creating a strategic risk-based piping inspection and analysis system in accordance with the piping damage characteristics in the petrochemical industry. In accordance with various piping damage models and damage positions, petrochemical plants provide the optimal piping inspection planning tool for efficient piping risk prediction for enhancing plant operation safety.

• 기계저널
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• 제19권4호
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• pp.326-331
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• 1979

압력배관에 대한 미국의 국가규준으로는 다음과 같은 것이 있다. Section 1. Power Piping ANSI B31.1 Section 2. Fuol Gas Piping ANSI B31.2 Section 3. Petroleum Refinery Piping ANSI B31.3 Section 4. Liquid Potroleum Transportation Piping ANSI B31.4 5. Section Refrigeration Piping ANSI 31.5 Section 6. Chemical Plant Piping ANSI B31.6 Section 7. Nuclear Power Piping ANSI B31.7 Section 8. Gas Transmission and Distribution Piping Systems ANSI B31.8 이중에서 Power Piping ANSI B31.1은 1977년도에 공진청에서 제정한 "압력배관 기술 기준 (1) "의 기본이 되고 있다. 금반의 냉동배관 기술기준 제정에 있어서도 이것이 압력배관의 범주내에 포함되는 것이기 때문에 기준의 통일성을 기하기 위하여서는 압력 배관기술기준(1)에 준하여 ANSI B31.5 Refrigeration Piping을 기본으로 하여야 할 것으로 고려하였다. 현재 각국의 압력 배관에 대한 기술기준은 그 형식은 여하간에 기본적으로는 ANSI B31. 시리즈에 따르고 있고 또 이 규준이 국제적으로 인정 널리 시행되고 있으므로 본 냉동배관 기술기준도 ANSI B31.5에 따라 제정하는 것이 타당성이 있는 것으로 고려하였다.

• 한국압력기기공학회 논문집
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• 제13권1호
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• pp.16-39
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• 2017

This paper reviewed and summarized the experimental studies conducted during last three decades to evaluate the structural integrity and to establish the acceptance criteria for piping system of nuclear power plants (NPPs) under seismic loading condition. These experimental studies contain the results of large-scale piping system tests under excessive seismic loading as well as standard specimen tests, simplified piping specimen tests, and piping components tests under simplified dynamic and cyclic loading. These would be useful as a basis for establishing integrity assessment procedure and acceptance criteria for piping systems of NPPs under beyond design basis earthquake (BDBE) conditions, and also could be used in planing the scope and direction of further related researches.

• 설비공학논문집
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• 제18권7호
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• pp.603-611
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• 2006

A variety of schemes were sought for a mitigation of thermal stratification phenomenon in the branch piping of domestic nuclear power plant. Several mechanisms of thermal stratification occurrence were introduced in this paper. A number of factors were selected to find out the schemes for thermal stratification mitigation and the computational analysis were performed. The length of vertical branch piping, the diameter, the radius of curvature of the elbow, the direction of connection between main piping and branch piping, the slope of branch piping, the leakage flow rate, the installation of additional valve, the change of the 1st valve position and another branch piping connected with branch piping were mentioned as factors in this paper.

• 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 $50^{\circ}C$.

• 한국기계기술학회지
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• 제13권4호
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• pp.9-14
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• 2011

The flow rate analysis for sanitary fixtures has been studied to determine the water supply piping system and size. The study has been carried out to analyze for a various water supply pressure and piping size theoretically. Also, the study has been carried out to analyze for a various water supply piping system experimentally. The water supply pressure is varied from 0.01MPa to 0.07MPa, and the piping size is varied from 6mm to 15mm. The water supply piping systems are one-to-one, all-loop-type, and bathroom-loop-type water supply piping system. The results indicate that the piping size is able to supply water fully in case of smaller than 15mm if the water supply pressure keep an necessary minimum pressure. And the gap of flow rate is very little for the various water supply piping systems.

• 한국압력기기공학회 논문집
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• 제14권2호
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• pp.41-48
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• 2018

In this study, the dynamic absorber and the damper are applied to improve the seismic performance of the piping system, and their quantitative effects on the piping system performance are examined. For this purpose, the response performances of piping system applied with the dynamic absorber/damper are compared with those of the original piping system. Firstly, the frequency response analyses of the piping system with the presence or the absence of dynamic absorber/damper are performed and these results are compared. It has been shown that the maximum acceleration response per the frequency of the piping system is considerably reduced by installing the dynamic absorber and the damper. Secondly, the seismic responses of the piping systems with and without dynamic absorber/damper are compared. As a result of the numerical analyses, it is confirmed that key responses are reduced by 17%-63% due to the installation of the dynamic absorber and damper. Finally, as a result of the seismic performance evaluation, it is confirmed that the HCLPF (High Confidence of Low Probability of Failure) seismic performances are increased by 1.22 to 2.70 times with respect to the failure modes with an aid of the dynamic absorber and damper.

• 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.

• 대한기계학회논문집A
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• 제23권7호
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• pp.1182-1195
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• 1999

The purpose of this study is to investigate experimentally the effects of both seismic loading and crack length on the fracture behavior of piping system with a circumferential crack in nuclear power plants. The experiments were performed using both large scale piping system facility and 4 points bending test machine under PWR operating conditions. The difference in the load carrying capacities between cracked piping and non-cracked piping was also investigated using the results from experiments and numerical calculations. The results obtained from the experiments and estimation are as follows : (1) The safety margin under seismic loading is larger than those under quasi static loading or simple cyclic loading. (2) There was no significant effect of crack length on tincture behavior of piping system with both a surface crack and a through-wall crack. (3) The load carrying capacity in cracked piping was reduced by factors of 7 to 46 compared to non-cracked piping.