• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.4
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• pp.405-413
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• 2015

In the US, the number of cases of subterranean water contamination from tritium leaking through a damaged buried nuclear power plant pipe continues to increase, and the degradation of the buried metal piping is emerging as a major issue. A pipe blocked from corrosion and/or degradation can lead to loss of cooling capacity in safety-related piping resulting in critical issues related to the safety and integrity of nuclear power plant operation. The ASME Boiler and Pressure Vessel Codes Committee (BPVC) has recently approved Code Case N-755 that describes the requirements for the use of polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. This paper contains tensile and slow crack growth (SCG) test results for high-density polyethylene (HDPE) pipe welds under the environmental conditions of a nuclear power plant. Based on these tests, the fracture surface of the PENT specimen was analyzed, and the fracture mechanisms of each fracture area were determined. Finally, by using 3D finite element analysis, limit loads of HDPE related to premature failure were verified.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.50-56
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• 2010

Fatigue has been known as a major degradation mechanism of ASME class 1 components in nuclear power plants. Fatigue damage could be accelerated by combined interaction of several loads and environmental factors. However, the environmental effect is not explicitly addressed in the ASME S-N curve which is based on air at room temperature. Therefore many studies have been performed to understand the environmental effects on fatigue behavior of materials used in nuclear power plants. As a part of efforts, we performed low cycle fatigue tests under various environmental conditions and analyzed the environmental effects on the fatigue life of SA508 Gr. 1a low alloy steel by comparing with higuchi's model. Test results show that the fatigue life depends on water temperature, dissolved oxygen and strain rate. But strain rate over 0.4%/s has little effect on the fatigue life. To find the cause of different fatigue life with ANL's and higuchi's model, another test performed with different heat numbered and heat treated materials of SA508 Gr. 1a. On a metallurgical point of view, the material with bainite microstructure shows much longer fatigue life than that with ferrite/pearlite microstructure. And the characteristics of crack propagation as different microstructure seem to be the main cause of different fatigue life.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.1-18
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• 2017

Metal fatigue is an important aging mechanism that material characteristics can be deteriorated when even a small load is applied repeatedly. An accurate fatigue evaluation is very important for component structural integrity and reliability. In the design stage of a nuclear power plant, the fatigue evaluations of the Class 1 components have to be performed. However, operating experience shows that the design evaluation can be very conservative due to conservatism in the transient severity and number of occurrence. Therefore, the fatigue monitoring system has been considered as a practical mean to ensure safe operation of the nuclear power plants. The fatigue monitoring system can quantify accumulated fatigue damage up to date for various plant conditions. The purpose of this paper is to describe the fatigue monitoring procedure and to introduce the fatigue monitoring program developed by the authors. The feasibility of the fatigue monitoring program is demonstrated by comparing with the actual operating data and finite element analysis results.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.37-42
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• 2010

The stay cylinder weld at the steam generator of Korean Standard Nuclear Power Plants is safety class I component and is subjected to be inspected by the volumetric examination such as ultrasonic method. As accessibility of this area is limited due to the narrow space and high radiation, the existing manual inspection method involves various difficulties. Moreover operators may be exposed to internal contamination by contaminated dust during the surface buffing process to improve the inspection reliability of this area. Recently the new automatic inspection system for stay cylinder welds has been developed. The inspection system basically consists of a driving assembly, data acquisition device and signal processing units. The driving assembly is classified by 1) the scanner for inspecting and buffing the weld, 2) pillars for guiding the scanner and 3) the base frame for loading and supporting pillars. The scanner has 4 sensor modules to inspect in 4 refracted angles and 4 incident directions. These components can be inserted into the skirt of the stay cylinder through the manway hole and assembled easily by one-touch in the skirt. Data acquisition device and signal processing units developed in previous works are also newly upgraded for better processing of data analysis and evaluation. The system has been successfully demonstrated not only in the mock-up but also in the field. In this paper, newly developed inspection system for the stay cylinder weld of the steam generator is introduced and their field applications are discussed.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.4
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• pp.378-383
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• 2004

The Korean Performance Demonstration(KPD) System for the ultrasonic testing personnel, equipments and procedures applicable to the Class 1 and 2 piping examination for nuclear power plant in Korea has been established. A round robin test was conducted in order to compare the examination results by the method of Performance Demonstration(PD) with the traditional dB-drop method. The round robin test shows that the reliability of the PD method is better than that of the dB-drop method. As a result, adoption of the PD method to the in-service inspection of the nuclear power plants will improve the reliability of the ultrasonic test results.

• Journal of the Korean Society of Marine Environment & Safety
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• v.28 no.1
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• pp.184-192
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• 2022

Valve internal leakage is caused by damage to the internal parts of the valve, resulting in accidents and shutdowns of the piping system. This study investigated the possibility of a real-time leak detection method using the acoustic emission (AE) signal generated from the piping system during the internal leakage of a butterfly valve. Datasets of raw time-domain AE signals were collected and postprocessed for each operation mode of the valve in a systematic manner to develop a data-driven model for the detection and classification of internal leakage, by applying machine learning algorithms. The aim of this study was to determine whether it is possible to treat leak detection as a classification problem by applying two classification algorithms: support vector machine (SVM) and convolutional neural network (CNN). The results showed different performances for the algorithms and datasets used. The SVM-based binary classification models, based on feature extraction of data, achieved an overall accuracy of 83% to 90%, while in the case of a multiple classification model, the accuracy was reduced to 66%. By contrast, the CNN-based classification model achieved an accuracy of 99.85%, which is superior to those of any other models based on the SVM algorithm. The results revealed that the SVM classification model requires effective feature extraction of the AE signals to improve the accuracy of multi-class classification. Moreover, the CNN-based classification can be a promising approach to detect both leakage and valve opening as long as the performance of the processor does not degrade.