• Nuclear Engineering and Technology
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• v.40 no.7
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• pp.607-614
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• 2008

Since pipes with wall-thinning defects can collapse at fluid pressure that are lower than expected, the collapse moment of wall-thinned pipes should be determined accurately for the safety of nuclear power plants. Wall-thinning defects, which are mostly found in pipe bends and elbows, are mainly caused by flow-accelerated corrosion. This lowers the failure pressure, load-carrying capacity, deformation ability, and fatigue resistance of pipe bends and elbows. This paper offers a support vector regression (SVR) model further enhanced with a fuzzy algorithm for calculation of the collapse moment and for evaluating the integrity of wall-thinned piping systems. The fuzzy support vector regression (FSVR) model is applied to numerical data obtained from finite element analyses of piping systems with wall-thinning defects. In this paper, three FSVR models are developed, respectively, for three data sets divided into extrados, intrados, and crown defects corresponding to three different defect locations. It is known that FSVR models are sufficiently accurate for an integrity evaluation of piping systems from laser or ultrasonic measurements of wall-thinning defects.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.40-45
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• 2019

Unlike the welded pipes in the primary system of light water nuclear power plants being periodically inspected with in-Service inspection program, relatively small pipes with the outer diameter less than 2 inch have not been regularly inspected to date. However, after several failure reports on the occurrence of critical crack-like defects in small pipes, inspection for the small pipes has been more demanded because it could cause the provisional outage of nuclear power plants. Nevertheless, there's no particular method to examine the small pipes having access limitations for inspection due to various reasons; inaccessible area, excessive radiation exposure, hazardous surrounding, and etc. This study is to develop a reliable inspection technique using torsional and flexural modes of guided wave to detect defects that could occur in inaccessible area. The attribute of guided wave that can travel a long distance enables to inspect even isolated range of the pipe from accessible location. This paper presents a case study of the evaluation test on 3/4" small-bore pipes with guide wave method. The test result demonstrates the crack signal behavior and assures possibility to detect the crack signal in a flexural mode, which is clearly distinguishable from the symmetric structure signal in a torsional mode.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.69-76
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• 2018

The paper presents preliminary investigation results for the effect of the baseline correction in the acceleration excitation method on finite element seismic analysis results (such as accumulated equivalent plastic strain, equivalent plastic strain considering cyclic plasticity, von Mises effective stress, etc) of nuclear safety Class I components. For investigation, finite element elastic-plastic time-history seismic analysis is performed for a surge line including a pressurizer lower head, a pressurizer surge nozzle, a surge piping, and a hot leg surge nozzle using the Chaboche hardening model. Analysis is performed for various seismic loading methods such as acceleration excitation methods with and without the baseline correction, and a displacement excitation method. Comparing finite element analysis results, the effect of the baseline correction is investigated. As a result of the investigation, it is identified that finite element analysis results using the three methods do not show significant difference.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.2
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• pp.101-108
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• 2021

To date, a number of stress intensity factor (SIF) solutions have been proposed for the cylindrical structure with circumferential through-wall cracks. However, each solution has a different format as well as applicable range. It is also known that there is a significant difference in predicted SIF values depending on the shape of the structure and the size of the crack. In this study, the applicability of various SIF solutions was analyzed by comparing the finite element analysis results for the case where a tensile load is applied to the cylindrical structure with circumferential through-wall crack. It is found that the calculated SIF gradually decreases and converges to a certain value with increasing length-to-radius ratio. Therefore, an appropriate length-to-radius ratio should be set in consideration of the dimensions of the actual cylindrical structure. For piping with sufficiently long cylinder, the ASME solution is found to be the most appropriate, and for a short cylinder, the API solution should be applied. On the other hand, the WEC solution requires careful attention to its application.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.93-101
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• 2023

In this study, a deep learning algorithm was used to diagnose electric potential signals obtained through CIPS and DCVG, used indirect inspection methods to confirm the soundness of buried pipes. The deep learning algorithm consisted of CNN(Convolutional Neural Network) model for diagnosing the electric potential signal and Grad CAM(Gradient-weighted Class Activation Mapping) for showing the flaw prediction point. The CNN model for diagnosing electric potential signals classifies input data as normal/abnormal according to the presence or absence of flaw in the buried pipe, and for abnormal data, Grad CAM generates a heat map that visualizes the flaw prediction part of the buried pipe. The CIPS/DCVG signal and piping layout obtained from the 3D finite element model were used as input data for learning the CNN. The trained CNN classified the normal/abnormal data with 93% accuracy, and the Grad-CAM predicted flaws point with an average error of 2m. As a result, it confirmed that the electric potential signal of buried pipe can be diagnosed using a CNN-based deep learning algorithm.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.18 no.3
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• pp.600-608
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• 2017

The presence of piping in a levee body allows water seepage to occur by producing a large cavity or water tunnel within it, ultimately resulting in the failure of the river levee and differential settlement. In order to properly cope with river levee failure due to piping and establish a proper remediation method for this problem, it is necessary to analyze the failure mechanism of the river levee due to piping. Therefore, this study analyzed the shape and mechanism of river levee failure due to piping through small-scale and large-scale models and evaluated the seepage pressure distribution characteristics in the hydraulic well, which has been suggested as a remediation method for piping. According to the results of this study, as the safety factor for the piping in the river levee decreased, the river levee failure shape was more clearly shown through the small-scale model test. In the large-scale model test, the type of local damage to the levee due to the piping was identified and the evaluation showed that the hydraulic well had the largest effect on the inhibition of piping below the center of the well. A follow-up study is needed to confirm the reliability of the results. However, it is thought that this study can be utilized as the baseline data for research into the piping-induced river levee failure mechanism and for the preparation of a remediation method.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.50-57
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• 2019

This paper investigates deformation characteristics of gas pipeline using the in-plane bending experiment and finite element analysis of a pipe bend. The effect of the bending angle and internal pressure on the deformation characteristics is analyzed. The pipe bend used in this study is API 5L X65 (out diameter: 20 inch) material with the thickness of 11.9 mm. The maximum load, displacement at maximum load, angle and local strain of 90° pipe bend are obtained from the in-plane bending experiment. Comparison between FE results and experimental data shows overall good agreements. In addition, the deformation characteristics of 22.5° and 45° pipe bend are calculated using the finite element analysis. As a result, the effect of the bend angle on the deformation characteristics is discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.