• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.54-60
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• 2022

ASME BPVC provides stress evaluation rules for Class 2 and 3 nuclear piping. However, such rules are difficult to be applied to reinforced wall-thinned pipes during service. To resolve this issue, a new method for stress evaluation of reinforced wall-thinned pipes is proposed in this work, based on the equivalent stiffness concept. By converting a reinforced wall-thinned pipe to an equivalent straight pipe having the same stiffness, stress evaluation can be proceeded using the current ASME BPVC rules. The proposed method is applied to pipes with 4 different normal pipe size and the effects of reinforcement and wall-thinning dimensions on evaluated stresses are discussed.

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.5
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• pp.422-429
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• 2013

It is of great importance to localize leakages in complex pipelines for assuring their safety. A sensor array that can detect where leakages occur enables us to monitor a wide area with a relatively low cost. Beamforming is a fast and efficient algorithm to estimate where sources are, but it is generally made use of in free field condition. In practice, however, many pipelines are placed in a closed space for the purpose of safety and maintenance. This leads us to take reflected waves into account to the beamforming for interior leakage localization. Beam power distribution of reflected waves in a closed space is formulated, and spatial average is introduced to suppress the effect of reflected waves. Computer simulations and experiments ensure how the proposed method is effective to localize leakage in a closed space for structural health monitoring.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.29 no.7 s.238
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• pp.947-954
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• 2005

The purpose of this study is to develop failure pressure evaluation models, which are applicable to straight pipes and elbows containing an internally wall thinning defect induced by flow-accelerated-corrosion (FAC). In this study, thus, three dimensional finite element (FE) analyses are performed to investigate the dependences of failure pressure of internally wall thinned pipe on the defect shape, the pipe geometry, and the defect location and bend radius of elbow. Also, the existing failure pressure assessment models for externally wall thinned pipes are examined. Based on these, the new models for assessing failure pressure of piping components with an internally wall thinning defect are proposed. Comparison of failure pressure, predicted by proposed models, with FE analysis result shows good agreement regardless of pipe type, defect shape, and defect location and bend radius.

• Proceedings of the Korean Information Science Society Conference
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• 2002.04b
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• pp.511-513
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• 2002

최근 가상현실과 Web 3D에 대한 관심이 고조되면서 관련 연구와 기술개발이 활발히 진행되고 있다. 본 논문에서는 Web 기반에서 3차원 그래픽을 표현하는 표준언어인 VRML을 이용하여, 원자력발전소에서 주기적으로 안정성 평가를 위해 수행하는 ISI(In-Service Inspection : 가동중검사) 업무에 적용하여 가상원전 3D ISI 시스템을 개발하고 그 구현방법을 제시한다. 검사대상 기기, 배관 및 각종 지지구조물에 대한 도면확보의 작업을 시작으로 하여, 3D 모델 구축, VR Data 작성, 그래픽 관리 시스템 개발 사례와 가상면실에 의해 구현된 Scene과 각종 DB글 연결하는 Interactive 3D Visualization Tool을 개발하여 기존의 2차원적 DB 운영을 3차원 가상공간에서 운영함으로써 보다 효과적이고 효율적인 DB 운용 방법에 대해서 기술한다.

• Proceedings of the KIEE Conference
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• 2002.07d
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• pp.2761-2763
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• 2002

원자력발전소에서는 건설완료 단계에서 기기, 배관 및 각종 지지구조물에 대하여 PSI(Pre-Service Inspection)를 실시하여 시공된 설비의 건전성을 점검하고, 운영중에는 주기적으로 시간의 경과에 따른 취약화 정도를 비과괴검사로 감시 및 평가하는 ISI (In-Service Inspection)를 실시하여 안전성 관련 설비의 건전성을 점검하고 있다. 이 과정에서 방대한 양의 검사데이터가 생산되며 원전 가동년수가 증가함에 따라 검사자원을 체계적으로 관리하고 자동화 할 수 있는 시스템 개발이 필요하게 되었다. 본 논문에서는 PSI/ISI 검사결과 정보관리를 비롯하여 검사 기초자료 관리 및 검사계획 수립, 파괴역학 분식 기능 등을 포함한 Web 기반의 Application 개발 사례를 소개한다. 또한 가상현실 기법을 적용하여 기존의 2차원적 DB 운영을 3차원 가상공간에서 운영할 수 있는 'Touch & Tell' 기법을 개발하고 그 구현방법을 제시한다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.31-39
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• 2015

Pipe wall-thinning by flow-accelerated corrosion (FAC) and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs). Most NPPs have management programs to ensure pipe integrity due to wall-thinning that includes periodic measurements for pipe wall thicknesses using ultrasonic tests (UTs). Nevertheless, thinning evaluations are not easy because the amount of thickness reduction being measured is often quite small compared to the accuracy of the inspection technique. U.S. Electric Power Research Institute (EPRI) had proposed Total Point Method (TPM) as a thinning occurrence evaluation method, which is a very useful method for detecting locally thinned pipes or fittings. However, evaluation engineers have to discern manually the measurement data because there are no numerical algorithm for TPM. In this study, numerical algorithms were developed based on non-parametric and parametric statistical method.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.44-50
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• 2011

Turbulent mixing of hot and cold coolants is one of the possible causes of high cycle thermal fatigue in piping systems of nuclear power plants. A typical situation for such mixing appears in turbulent flow through a T-junction. Since the high cycle thermal fatigue caused by thermal striping was not considered in the piping fatigue design in several nuclear power plants, it is very important to evaluate the effect of thermal striping on the integrity of mixing tees. In the present work, before conducting detailed evaluation, three thermal striping evaluation methodology suggested by EPRI, JSME and NESC are analyzed. Then, a by-pass pipe connected to the shutdown cooling system heat exchanger is investigated by using these evaluation methodology. Consequently, the resulting thermal stresses and the fatigue life of the mixing tee are reviewed and compared to each other. Futhermore, the limitation of each methodology are also presented in this paper.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.40-47
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• 2011

KSNPs(Korea Standard Nuclear Power Plant) have been applied the break exclusion criteria to the high energy lines passing through containment penetration area to ensure that piping failures would not cause the loss of containment isolation function, and to reduce the resulting dynamic effects. Systems with the criteria are the Main Steam system, Feed Water system, Steam Generator Blowdown system, and Chemical & Volume Control system. In accordance with FSAR(Final Safety Analysis Report), a 100% volumetric examination by augmented in-service inspection of all pipe welds appled the break exclusion criteria is required for the break exclusion application piping. However, it is difficult to fully satisfy the requirements of inspection because 12", 8" and 6" weldolet weldments of Main Steam pipe line have complex structural shapes. To resolve the difficulty on the application of conventional UT(Ultrasonic Testing) technique, realistic mock-ups and UT calibration blocks were made. Simulations of conventional UT were performed utilizing CIVA, a commercial NDE(Nondestructive Examination) simulation software. Phased array UT experiments were performed through mock-up including artificial notch type flaws. A phased array UT technique is finally developed to improve the reliability of ultrasonic test at main steam line pipe to 12", 8" and 6" branch connection weld.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.54-60
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• 2010

Piping failures due to thermal fatigue have been widely reported in normally stagnant non-isolable reactor coolant branch lines. Since the thermal fatigue due to thermal stratification was not considered in the piping fatigue design in old NPPs, it is important to evaluate the effect of thermal stratification on the integrity of branch lines. In this study, geometrical screening criteria for Up-horizontal branch lines in MRP-132 were applied to SI(Safety Injection) lines of KSNP 2-loop and WH 3-loop. Some computational fluid dynamic(CFD) analyses on the Reactor Coolant System(RCS) branch lines were also performed to develop the regulatory guidelines for screening criteria. As a result of applying MRP-132 screening criteria, KSNP 2-loop and WH 3-loop SI lines are determined to need further detailed evaluation. Results of CFD analyses show that both valve isolation and amount of leakage through valve can be used as technical bases for the screening criteria on the thermal fatigue analysis.