• Journal of the Earthquake Engineering Society of Korea
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• v.19 no.1
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• pp.29-36
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• 2015

Base isolation is considered as a seismic protective system in the design of next generation Nuclear Power Plants (NPPs). If seismic isolation devices are installed in nuclear power plants then the safety under a seismic load of the power plant may be improved. However, with respect to some equipment, seismic risk may increase because displacement may become greater than before the installation of a seismic isolation device. Therefore, it is estimated to be necessary to select equipment in which the seismic risk increases due to an increase in the displacement by the installation of a seismic isolation device, and to perform research on the seismic performance of each piece of equipment. In this study, modified NRC-BNL benchmark models were used for seismic analysis. The numerical models include representations of isolation devices. In order to validate the numerical piping system model and to define the failure mode, a quasi-static loading test was conducted on the piping components before the analysis procedures. The fragility analysis was performed by using the results of the inelastic seismic response analysis. Inelastic seismic response analysis was carried out by using the shell finite element model of a piping system considering internal pressure. The implicit method was used for the direct integration time history analysis. In addition, the collapse load point was used for the failure mode for the fragility analysis.

• Journal of the Korea institute for structural maintenance and inspection
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• v.28 no.3
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• pp.91-99
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• 2024

Piping systems are crucial facilities used in various industries, particularly in areas related to daily life and safety. Piping systems are fixed to the main structures of buildings and facilities but do not support external loads and serve as non-structural elements performing specific functions. Piping systems are affected by relative displacements owing to phase differences arising from different behaviors between two support points under seismic loads; this can cause damage owing to the displacement-dominant cyclic behavior. Fittings and joints in piping systems are representative elements that are vulnerable to seismic loads. To evaluate the seismic performance and limit states of fittings and joints in piping systems, a high-stroke actuator is required to simulate relative displacements. However, this is challenging because only few facilities can conduct these experiments. Therefore, element-level experiments are required to evaluate the seismic performance and limit states of piping systems connected by fittings and joints. This study proposed a method to evaluate the seismic performance of an elbow specimen that includes fittings and joints that are vulnerable to seismic loads in vertical piping systems. The elbow specimen was created by connecting straight pipes to both ends of a 90° pipe elbow using flexible groove joints. The seismic performance of the elbow specimen was evaluated using a cyclic loading protocol based on deformation angles. To determine the margin of the evaluated seismic performance, the limit states were assessed by applying cyclic loading with a constant amplitude.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.10
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• pp.1105-1110
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• 2011

Flow-accelerated corrosion (FAC) is a well-known phenomenon that may occur in piping and components. Most nuclear power plants have carbon-steel-pipe wall-thinning management programs in place to control FAC. However, various other erosion mechanisms may also occur in carbon-steel piping. The most common forms of erosion encountered (cavitation, flashing, Liquid Droplet Impingement Erosion (LDIE), and Solid Particle Erosion (SPE)), have caused wall thinning, leaks, and ruptures, and have resulted in unplanned shutdowns in utilities. In particular, the damage caused by LDIE is difficult to predict, and there has been no effort to protect piping from erosive damage. This paper presents an evaluation method for LDIE. It also includes the calculation results from prediction models, a review of the experimental results, and a comparison between the UT data in the damaged components and the results of the calculations and experiments.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.2
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• pp.14-20
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• 2011

The end of design lifetime for Wolsong unit 1 will be reached on 20th November in 2012. So the license renewal documents for the continuous operation of Wolsong unit 1 is under reviewing now. Major components of primary system such as pressure tubes, feeder pipes including delayed neutron monitoring system tubing are being replaced and many components of secondary system are also being repaired. In this paper, the assessment on the wear degradation of delayed neutron monitoring system tubing(on the other hand, DN tube was called) was performed for the ageing management of the same component. The wear defects of this component was one of causes that resulted in heavy water leakage accidents. Therefore design specifications of Wolsong uint 1 and heavy water leakage accidents of pressurized heavy water reactors were reviewed and causes of wear defect for DN tubes were analyzed. Wear propagation equations based on the heavy water leakage history were made and the proper repairing time was possible to be expected if the continued operation was considered. Finally design change items of DN tubes that were conducted for the long term operation of Wolsong unit 1 are introduced.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.1-7
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• 2013

In around 2000, the U.S. NPPs have developed the various advanced engineering processes based on the INPO AP-913(Equipment Reliability Process Description) and showed the high performance in availability. With these benchmarking cases, the Korean NPPs have introduced the advanced engineering technology since 2005. The first step of the advanced engineering is to analyze and determine component importance for all components of a plant. This process is called Functional Importance Determination(FID). These results are basically utilized to determine the priority with limited resources in various areas. However, because the consistency of FID results is insufficient despite applying the same criteria in the existing operating NPPs, the degree of application is low. Therefore, this paper presents the improved methodology for FID interfacing system functions of Maintenance Rule Program and results of Single Point Vulnerability(SPV). This improved methodology is expected to contribute to enhance the reliability of FID data.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.1
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• pp.24-31
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• 2018

The mechanical design procedure is studied for the PCHE(printed circuit heat exchanger) with electrochemical etched flow channels. The effective heat transfer plates of PCHE are assembled by diffusion bonding to make a module. PCHE is widely used for industrial applications due to its compactness, cost efficiency, and serviceability at high pressure and/or temperature conditions. The limitations and technical barriers of PCHE are investigated for application to nuclear components. Rules for design and fabrication of PCHE are specified in ASME Section VIII but not in ASME Section III of nuclear components. Therefore, the calculation procedure of key dimensions of PCHE is defined based on ASME section VIII. The effective heat transfer region of PCHE is defined by several key dimensions such as the flow channel radius, edge width, wall thickness, and ridge width. The mechanical design procedure of key dimensions was incorporated into a program for easy use in the PCHE design. The effect of assumptions used in the key dimension calculation on stress values is numerically investigated. A comparative analysis is done by comparing finite element analysis results for the semi-circular flow channels with the formula based sizing calculation assuming rectangular cross sections.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.26-32
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• 2012

The systems of the nuclear power plant are designed based on the User Requirement Document, and Korea Hydro & Nuclear Power Co. (KHNP) implements preventive maintenance activities to keep the specific design function of the system consistently. To monitor the preventive maintenance effectiveness, KHNP has also developed maintenance effectiveness monitoring (MR) program based on NUMARC 93-01 since 2003, and has implemented the program in all operating plants. Recently, KHNP has upgraded MR programs by reflecting implementing experiences ; reestablishing the performance monitoring level, improving analysis for standby function and performance criteria for passive components, reestablishing the availability performance criteria and the performance criteria for the same type of components. These upgraded MR programs will contribute to enhance safety and improve equipment reliability through monitoring maintenance effectiveness.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.38-43
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• 2011

For the efficient and safe operation of nuclear power plant, evaluating quantitatively aging phenomenon of major components is necessary. Especially, typical aging parameters such as stresses and cumulative usage factors should be determined accurately to manage the lifetime of the plant facility. The 3-D finite element(FE) model is generated to calculate the aging parameters. Mechanical and thermal transfer functions called Green's functions are developed for the FE model with standard step input. The stress results estimated from transfer functions are verified by comparing with 3-D FE analyses results. Lastly, we suggest an effective fatigue evaluation methodology by using the transfer functions. The usefulness of the proposed fatigue evaluation methodology can be maximized by combining it with an on-line monitoring system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.25-30
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• 2013

The periodic inspection of piping and pressure vessels welds in nuclear power plant has to provide reliable result related to weld flaws, such as location, maximum amplitude response, ultrasonic length, height and finally the nature or flaw pattern. The founded flaw in ultrasonic inspection is accepted or rejected based on these data. Specially, the amplitude of flaw response is used as basic parameter for flaw sizing and it may cause some deviation in length sizing result. Currently the ultrasonic inspections in nuclear power plant components are performed by specific inspection procedure which describing inspection technique include inspection system, calibration methodology and flaw characterizing. To perform ultrasonic inspection during in-service inspection, reference gain should be established before starting ultrasonic inspection by the requirement of ASME code. This reference gain used as basic criteria to evaluate flaw sizing. Sometimes, a little difference in establishing reference gain between calibration and field condition can lead to deviation in flaw sizing. Due to this difference, the inspection result may cause flaw sizing error. Therefore, the objective of this study is to compare and evaluate the ultrasonic amplitude difference between air filled and water filled pipe in nuclear power plant. Additionally, the accuracy of flaw sizing is estimated by comparing both conditions.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.2
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• pp.66-70
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• 2016

Following the 2011 Fukushima Nuclear Power Plant accident, the IAEA has issued a revised version of the Nuclear Safety Standard for beyond design basis earthquake to consider the core meltdown accident. In Korea, relevant laws and regulations were also revised to consider beyond design basis earthquake to nuclear components. In this paper, CAV, an seismic damage factor that determines the restart of nuclear power plant after operating breakdown earthquake, is proposed for extension to the beyond design basis earthquake. For pipings not satisfying the beyond design basis earthquake condition, several evaluation methods are suggested, such as strain-based evaluation methods, simple nonlinear analysis method and cumulative damage evaluation method.