• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.432-437
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• 2000

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The the radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively.

• 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.

• 한국압력기기공학회 논문집
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• 제13권2호
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• pp.44-50
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• 2017

The shaft seal of the reactor coolant pump is installed on the upper side of the rotating shaft of the pump to seal the reactor coolant from flowing out between the rotating shaft and the non-rotating parts. In this study, the loading conditions for the normal operation and faulted conditions are identified and structural integrity evaluation is performed using the finite element stress analysis for the sealing apparatus of the APR 1400 reactor coolant pump. It is confirmed that the stress analysis results satisfy the design criteria at all loading conditions.

• 한국압력기기공학회 논문집
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• 제14권1호
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• pp.58-65
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• 2018

The purpose of this study is to develop a Half Nozzle Repair(HNR) process to prevent the leakage from welds on small bore piping in Reactor Coolant System. The Codes & Standards of tempered bead and design requirements of J-Groove welds are reviewed. Automatic machine GTAW welding and machining equipments are developed to perform HNR process. Single pass welding and overlay welding equipments are conducted in order to obtain the optimal temper bead welding process parameters with Alloy 52M filler wire. Coarse grain heat affected zone(CGHAZ) is formed by rapid cooling rate in heat affected zone after welding. Accordingly, a proper temper bead technique is required to reduce CGHAZ in 1-Layer of welds by 2- and 3-Layers. Mock-up tests show that the developed HNR process is possible to meet ASME Code & Standard requirements without any defect.

• 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.

• 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 $50^{\circ}C$.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.1833-1838
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• 2003

In the nuclear power plant, emergency core coolant system(ECCS) is furnished at reactor coolant system(RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can be occurred due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack(TFC) accident. In the present study, when the turbulence penetration occurs in the branch piping, the maximum temperature differences of fluid at the pipe cross-sections of the T-branch with thermal stratification are examine

• 한국압력기기공학회 논문집
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• 제19권1호
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• pp.44-51
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• 2023

The RCP is one of the main components in nuclear power plants and plays an important role in circulating coolant to the RCS system. Currently, nuclear plants are monitored using various monitoring systems. However, since they operate independently according to their functional purpose, it is not able to analyze vibration and operation/performance information comprehensively, and thus failure diagnosis accuracy is limited. In addition, these systems do not provide some important information (such as fault type, parts and cause) necessary for emergency actions, but provide only alarm information. To improve these technical problems, this study proposes a diagnosis technique (M/L, Rule-based model, Data-driven model, Narrow band model) and methodology for comprehensive analysis.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 추계학술대회논문집
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• pp.74-79
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• 2002

The piping verification tests were performed in order to verify the structural integrity during initial operation of the reactor coolant systems and the primary heat transportation systems of nuclear power plants by KEPRI in Korea. The tests were conducted at full operating temperature and pressure. The objective is to evaluate the possibility of excessive load generating on piping, piping supports, and reactor structures etc. in the steady normal operation and expected pump transient conditions. As a result, the measured vibrations have been shown acceptable level according to ASME/ANSI OMa-Standard, Part 3.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 추계학술대회논문초록집
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• pp.315.1-315
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• 2002

The piping verification tests were performed in order to verify the structural integrity during initial operation of the reactor coolant systems and the primary heat transportation systems of nuclear power plants by KEPRI in Korea. The tests were conducted at full operating temperature and pressure. The objective is to evaluate the possibility of excessive load generating on piping, piping supports, and reactor structures etc. in the steady normal operation and expected pump transient conditions. (omitted)