• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.16-39
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• 2017

This paper reviewed and summarized the experimental studies conducted during last three decades to evaluate the structural integrity and to establish the acceptance criteria for piping system of nuclear power plants (NPPs) under seismic loading condition. These experimental studies contain the results of large-scale piping system tests under excessive seismic loading as well as standard specimen tests, simplified piping specimen tests, and piping components tests under simplified dynamic and cyclic loading. These would be useful as a basis for establishing integrity assessment procedure and acceptance criteria for piping systems of NPPs under beyond design basis earthquake (BDBE) conditions, and also could be used in planing the scope and direction of further related researches.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.67-76
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• 2000

Significant wall thinning due to flow accelerated corrosion(FAC）was recently reported in main feedwater pipes in 3 Korean pressurized water reactor(PWR) plants. The main feedwater pipes in one plant were repaired using overlay weld method at the outside of pipe, while those in 2 other plants were replaced with new pipes. In this study, the effect of the wall thinning in the main feedwater pipes on piping integrity was evaluated using finite element method. Especially, the effects of both the overlay weld repair and the stress concentration in notch-type thinned area on the piping integrity were investigated. The results are as follows : (1) The piping load carrying capacity may significantly decrease due to FAC. In special, the load carrying capacity of the main feedwater pipe was reduced by about 40% during about 140 months operation in Korean PWR plants. (2) By performing overlay weld repair at the outside of pipe, the piping load carrying capacity can increase and the stress concentration level in the thinned area can be reduced.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.74-83
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• 2020

A number of so-called 'Type IV' cracking was reported to occur at the welded joints of the P91 steel or P92 steel reheat steam piping systems in Korean supercritical thermal power plants. The reheat steam piping systems are subjected to severe thermal and pressure loading conditions of coolant higher than 570℃ and 4MPa, respectively. In this study, piping analyses and design evaluations were conducted for the piping system of a specific thermal plant in Korea and suggestions were made how structural integrity could be improved so that type IV cracks at the welded joints could be prevented. Integrity evaluations were conducted as per ASME B31.1 code with implicit consideration of creep effects which was used in original design of the piping system and as per nuclear-grade RCC-MRx code with explicit consideration of creep effects. Comparisons were made between the evaluation results from the two design rules. Another approach with modification or reduction of the redundant supports in the piping systems was investigated as a tool to mitigate thermal stresses which should essentially contribute to prevention of Type IV cracking without major modification of the existing piping systems. In addition, a post weld heat treatment method and repair weld method which could improve integrity of the welded joint of P91 steel were investigated.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.264-267
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• 2004

Wall thinning of carbon steel pipe components due to Flow-Accelerated Corrosion (FAC) is one of the most serious threats to the integrity of steam cycle piping systems in Nuclear Power Plants (NPP). Since the mid-1990s, secondary side piping systems in Korean NPPs have experienced wall thinning, leakages and ruptures caused by FAC. Korea Electric power Research Institute (KEPRI) and Korea Hydro & Nuclear Power Co., LTD. (KHNP) have conducted a study to develop the methodology for systematic pipe management and established the Korean Thinned Pipe Management Program (TPMP). To effectively maintain the integrity of piping system, FAC engineer should understand the criterions of the structural integrity evaluation and the safety margin assessment for the thinned pipe component. This paper describes the technical items of TPMP, and shows the example of the integrity evaluation and safety margin assessment for three thinned pipe component of a NPP.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.617-621
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• 2003

The purpose of this paper is to analyze piping failure trend of safety pipings In domestic nuclear power plants. First, database for the piping failure was constructed with 105 data fields. The database includes plant population data, event data, and service history data. 7 kinds of piping failures in domestic NPPs were investigated. Among the 7 cases, detailed root causes were investigated for 3 cases. The first one is pipe wall thinning in main feedwater pipings of Westinghouse 3 loop type plants. The root cause of the wall thinning was flow accelerated corrosion near welding area. The next one is leak event in chemical and volume control system(CVCS) due to vibration. Some cracks occurred in socket welding area. The events showed that the integrity or socket weld is very vulnerable to vibration. The last one is also a leak event in primary sampling line in Korean standard reactor due to thermal fatigue. Although the structural integrity was not maintained by the events, there was no effect on nuclear safety in the above 3 piping failure eases.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.7
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• pp.1031-1038
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• 2001

In order to verify the analytical methods predicting failure behavior of cracked piping, full-scale pipe tests are crucial in nuclear power plant piping. For this reason, series of international test programs have been conducted. However, full-scale pipe tests require expensive testing equipment and long period of testing time. The objective of this paper is to develop a test system which can economically simulate the full-scale pipe test regarding the integrity evaluation. This system provides the failure behavior of cracked pipe by testing a wide-plate specimen. The system provides the failure behavior of cracked pipe by testing a wide-plate specimen. The system was developed for the integrity evaluation of nuclear piping based on the methodology of hardware-in-the-loop (HiL) simulation. Using this simulator, the piping integrity can be evaluated based on the elastic-plastic behavior of full-scale pipe, and the high cost full-scale pipe test may be replaced with this economical system.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.20 no.2
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• pp.575-584
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• 1996

The objective of this paper is to develop an expert system for nuclear piping integrity. This paper describes the selection methodology of integrity evalution method and the inference of material properties. To select the integrity evaluation method, the weight factor for respective material properties was obtained by the sensitivity analysis of the effect of material properties on integrity evaluation method. Subsequently the possession ratio for respective integrity evaluation method was computed, and the most appropriate integrity evaluation method for given input information is selected. In the material properties inference, stress-strain curves and J-R curves were predicted from tensile properties such as yield strength and tensile strength.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.51-60
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• 2015

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion, cavitation, flashing and/or liquid droplet impingement, is a main concern in secondary steam cycle piping system of nuclear power plants in terms of safety and operability. Thinned pipe management program (TPMP) has being developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning. In this paper, newest technologies, standards and regulations related to the integrity assessment, repair and replacement of thinned pipe component are reviewed. And technical improvement items in TPMP to secure the reliability and effectiveness are also presented.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.322-327
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• 2001

The internal leaks of RCS pressure boundary valves may cause thermal fatigue crack because of the TASCS in RCS branch line. After experienced unisolable piping failures in several PWR plants, many studies have peformed to understand these phenomena and various methods were applied to ensure the structural integrity of piping. In this paper, the cause of unisolable piping failures and the alternatives to prevent recurrence of failure were reviewed. Also, the severity of piping failure including susceptibility of valve leaks was evaluated for the Westinghouse 2-loop plant. The length of turbulent penetration on RHR inlet piping was measured and, thermal fluid analysis and fatigue analysis was performed for this piping. As a means of ensuring the structural integrity, temperature monitoring and specialized UT and other alternatives were compared for the further application.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.40 no.9
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• pp.775-782
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• 2016

In this study, elevated temperature design and integrity evaluation have been conducted using two different piping design codes for the high-temperature piping systems of sodium integral effect test loop for safety simulation and assessment(STELLA-2) being developed by KAERI(Korea Atomic Energy Research Institute). The design code of ASME B31.1 for power piping and French nuclear grade piping design guideline, RCC-MRx RD-3600 were applied, and conservatism of those codes was quantified based on the piping integrity evaluation results. The piping system of Model DHRS, Model IHTS and PSLS are to be installed in STELLA-2. The integrity evaluation results for the three piping systems according to the two design codes showed that integrity of the piping system was confirmed. As a code comparison result, ASME B31.1 was shown to be more conservative for sustained loads while RD-3600 was more conservative for thermal loads compared to B31.1.