• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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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.

• Corrosion Science and Technology
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• 제14권1호
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• pp.1-11
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• 2015

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion (FAC), cavitation, flashing and/or liquid drop impingements, are a main concern in carbon steel piping systems of nuclear power plant in terms of safety and operability. Thinned pipe management program (TPMP) had been developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning in the secondary side piping system. This program also consists of several technical elements such as prediction of wear rate for each component, prioritization of components for inspection, thickness measurement, calculation of actual wear and wear rate for each component. Decision making is associated with replacement or continuous service for thinned pipe components. Establishment of long-term strategy based on diagnosis of plant condition regarding overall wall thinning is also essential part of the program. Prediction models of wall thinning caused by FAC had been established for 24 operating nuclear plants. Long term strategies to manage the thinned pipe component were prepared and applied to each unit, which was reflecting plant specific design, operation, and inspection history, so that the structural integrity of piping system can be maintained. An alternative integrity assessment criterion and a computer program for thinned piping items were developed for the first time in the world, which was directly applicable to the secondary piping system of nuclear power plant. The thinned pipe management program is applied to all domestic nuclear power plants as a standard procedure form so that it contributes to preventing an accident caused by FAC.

• Journal of Advanced Marine Engineering and Technology
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• 제37권7호
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• pp.784-791
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• 2013

Recent advanced IT made layout design fast and accurate by using algorithms. Layout design should be determined by considering the position of equipment with satisfying various space constraints and its component works with optimum performance. Especially, engine room layout design is performed with mother ship data, theoretical optimal solution, design requirements and several design constraints in initial design stage. Piping design is affected by position of equipment seriously. Piping design depends on experience of designer. And also piping designer should consider correlation of equipment and efficiency of space. In this study, space evaluation method has been used to evaluate efficiency of space. And also this study suggested object function for optimal piping route, Average Reservation Index(ARI), Estimated Piping Productivity(EPP) and with modified space evaluation method. In this study, optimum pipe routing system has been developed to reflect automated piping route with space efficiency and experience of piping designer. Engine room is applied to the design of the piping in order to confirm validity of the developed system.

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.33-41
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• 2012

The component cooling water heat exchangers are critical components in a nuclear power plant. As the operation years of the heat exchanger go by, the maintenance costs required for continuous operation also increase. Most heat exchangers have carbon steel shells, tube support plates and flow baffles. The titanium tube is susceptible to flow induced vibration. The damage on carbon steel tube support rod and titanium tube around cooling water entrance area is inevitable. Therefore, analysis of internal flow around the component cooling water entrance and tube channel is a good opportunity to seek for failure prevention practice and maintenance method. The numerical study was carried out by FLUENT code to find out the causes of tube failure and its location.

• 한국산업융합학회 논문집
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• 제23권4_2호
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• pp.627-635
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• 2020

In this study, the optimal support span determination of pipeline system was carried out in consideration of the effects of seismic loads. The theoretical support and structural analysis were used to determine the optimal support span of piping system according to pipe diameter using theoretical and structural deflection criteria. The reliability of the analysis results was secured by comparing the structural and theoretical results. In particular, the optimum support span of piping system was obtained by considering the effects of seismic load, and the optimal support span of pipe diameter and piping system tended to be proportional to each other. When considering the effects of earthquakes on different pipe diameters(300~2,500mm), the span length is reduced by up to 48% at the allowable stress criterion, and the pipe span length is reduced by up to 5.9% at the deflection criterion. It can be seen that the effect of the seismic load on the determination of the piping span length has a greater effect on the stress than the displacement.

• Structural Engineering and Mechanics
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• 제72권1호
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• pp.31-41
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• 2019

Structural integrity assessment of piping components is of paramount important for remaining life prediction, residual strength evaluation and for in-service inspection planning. For accurate prediction of these, a reliable fracture parameter is essential. One of the fracture parameters is stress intensity factor (SIF), which is generally preferred for high strength materials, can be evaluated by using linear elastic fracture mechanics principles. To employ available analytical and numerical procedures for fracture analysis of piping components, it takes considerable amount of time and effort. In view of this, an alternative approach to analytical and finite element analysis, a model based on relevance vector machine (RVM) is developed to predict SIF of part through crack of a piping component under fatigue loading. RVM is based on probabilistic approach and regression and it is established based on Bayesian formulation of a linear model with an appropriate prior that results in a sparse representation. Model for SIF prediction is developed by using MATLAB software wherein 70% of the data has been used for the development of RVM model and rest of the data is used for validation. The predicted SIF is found to be in good agreement with the corresponding analytical solution, and can be used for damage tolerant analysis of structural components.

• Smart Structures and Systems
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• 제14권6호
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• pp.1005-1030
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• 2014

The need for assessing dynamic response of typical industrial piping systems subjected to seismic loading motivated the authors to apply model reduction techniques to experimental dynamic substructuring. Initially, a better insight into the dynamic response of the emulated system was provided by means of the principal component analysis. The clear understanding of reduction basis requirements paved the way for the implementation of a number of model reduction techniques aimed at extending the applicability range of the hybrid testing technique beyond its traditional scope. Therefore, several hybrid simulations were performed on a typical full-scale industrial piping system endowed with a number of critical components, like elbows, Tee joints and bolted flange joints, ranging from operational to collapse limit states. Then, the favourable performance of the L-Stable Real-Time compatible time integrator and an effective delay compensation method were also checked throughout the testing campaign. Finally, several aspects of the piping performance were commented and conclusions drawn.

• Nuclear Engineering and Technology
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• 제49권2호
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• pp.349-359
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• 2017

During the past 25 years, in the context of probabilistic safety assessment, efforts have been directed towards establishment of comprehensive pipe failure event databases as a foundation for exploratory research to better understand how to effectively organize a piping reliability analysis task. The focused pipe failure database development efforts have progressed well with the development of piping reliability analysis frameworks that utilize the full body of service experience data, fracture mechanics analysis insights, expert elicitation results that are rolled into an integrated and risk-informed approach to the estimation of piping reliability parameters with full recognition of the embedded uncertainties. The discussion in this paper builds on a major collection of operating experience data (more than 11,000 pipe failure records) and the associated lessons learned from data analysis and data applications spanning three decades. The piping reliability analysis lessons learned have been obtained from the derivation of pipe leak and rupture frequencies for corrosion resistant piping in a raw water environment, loss-of-coolant-accident frequencies given degradation mitigation, high-energy pipe break analysis, moderate-energy pipe break analysis, and numerous plant-specific applications of a statistical piping reliability model framework. Conclusions are presented regarding the feasibility of determining and incorporating aging effects into probabilistic safety assessment models.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.9-15
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• 2010

This paper describes the Visual Tool for automatic ultrasonic examination that is under developing as a part of the project for development of automatic ultrasonic wave acquisition and analysis program. This tool that is supported by various image processing techniques will be adopted to detect the flaws in the component and piping welds in NPP. Visual Tool will enhance the integrity of nuclear power plant. The object of this paper is to address the Visual Tool which is developing for automatic ultrasonic inspection of welds in NPP.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.74-78
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• 2015

Nuclear power plant steam generator that is one of the main component has several thousands of thin tubes. And the steam generator tube is subject to damage because of the severe operation conditions such as the high temperature and pressure. Therefore periodic inspections are conducted to ensure the integrity of steam generator component. Hanul unit 3 also has been inspected in accordance with in-service inspection program and is scheduled to be replaced for exceeding the plugging rate which was recommended by manufacturer. During the steam generator replacement activity, we found several clustered porosity on inner surface of main feed water pipe. Additionally crack-like indications were found at weld interface between base material and weld of main feed water pipe. This paper describes the field experience and visual testing results for inner surface of main feed water pipes. The destructive test result had shown that these indications were porosities which were caused by manufacturing process not by operation service.