• Journal of Ocean Engineering and Technology
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• v.16 no.5
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• pp.66-72
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• 2002

Fracture behaviors of pipes with local wall thinning are very important for the integrity of nuclear power plant. However, effects of local wall thinning on strength and fracture behaviors of piping system were not well studied. Acoustic emission(AE) has been widely used in various fields because of its extreme sensitivity, dynamic detection ability and location of growing defects. In this study, we investigated failure modes of locally wall thinned pipes and AE signals by bending test. From test results, we could be divided four types of failure modes of ovalization, crack initiation after ovalization, local buckling and crack initiation after local buckling. And fracture behaviors such as elastic region, yielding region, plastic deformation region and crack progress region could be evaluated by AE counts, accumulative counts and time-frequency analysis during bending test. The result of the frequency range is expected to be basic data that can inspect plants in real-time.

• Proceedings of the KIEE Conference
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• 2006.07b
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• pp.901-902
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• 2006

In high-speed PM machines, rotor losses form a larger proportion of the total losses than usual in conventional low speed machines. In order to maintain the mechanical integrity of a high-speed PM rotor intended for high-speed operation, the rotor assembly is often retained within a sleeve or can. The sleeve is exposed to field produced by the stator from either the slotting or the mmf harmonics that are not synchronous with the rotor. These non-synchronous fields cause the significant rotor losses. An optimum design of high-speed PM machines requires the accurate prediction for these rotor losses. On the basis of analytical field analysis and 2D finite element analysis (FEA), this paper deals with the rotor losses.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.283-288
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• 1996

Core support barrel (CSB) assembly is one of the most important reactor internals structures supporting and protecting the nuclear core during normal operation and faulted events. For Yonggwang 3 and 4 (YGN 3&4), the adequacy of the analytical response prediction of reactor internals for flow induced vibration was demonstrated through the comprehensive vibration assessment program (CVAP) performed during hot functional test. Besides, the vibration characteristics of the CSB of operating nuclear power plant can be examined via the excore neutron noise monitoring signal. In this paper data from YGN 4 analyses, CVAP, and neutron noise monitoring system are compared and evaluated. In general, the results are comparable each other and conservative enough to ensure sufficient design margin and structural integrity. Further investigations on the modelling and analyses procedure are recommended to utilize the experimental results to the maximum extent. And collection of the neutron noise data is desired to serve as a baseline information.

• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.239-250
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• 2008

If there is a water flow with a range of temperature inside a pipe, the wanner water tends to float on top of the cooler water because it is lighter, resulting in the upper portion of the pipe being hotter than the lower portion. Under these conditions, such thermal stratification can play an important role in the aging of nuclear power plant piping because of the stress caused by the temperature difference and the cyclic temperature changes. This stress can limit the lifetime of the piping, even leading to penetrating cracks. Investigated in this study is the effect of thermal stratification on the structural integrity of the pressurizer surge line, which is reported to be one of the pipes most severely affected. Finite element models of the surge line are developed using several element types available in a general purpose structural analysis program and stress analyses are performed to determine the response characteristics for the various types of top-to-bottom temperature differentials due to thermal stratification. Fatigue analyses are also performed and an allowable environmental correction factor is suggested.

• Journal of the Korean Society of Safety
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• v.18 no.3
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• pp.39-45
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• 2003

The failure mode of a pipe due to local wall thinning is increasingly more attention in the nuclear power plant industry. To assess the integrity of locally wall thinned pipe, it is necessary to perform many simulations under various conditions. Because the modeling for locally wall thinned elbow is more complicated than that of straight pipe the efficient modeling method for finite element analysis is necessary. In this study, the more simple efficient modeling method of three-dimensional finite element analysis for locally wall thinned elbow has been suggested and verified. And using the method, the failure mode of local wall thinned elbows that have different thinning lengths and circumferential angles is evaluated. From the results, we concluded that the collapse load of elbows has been decreased by the increase of wall thinning shape factors such as thinning lengths and circumferential angles.

• Journal of the Korea Institute of Information and Communication Engineering
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• v.3 no.2
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• pp.373-382
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• 1999

Inheritance and concurrency are the primary feature of object oriented languages, and are especially important for code re-use. They provide maximum computational power and modeling power through concurrency of objects. But, concurrent objects and inheritance have conflicting characteristics, thereby simultaneously use of them causes the problem, so called inheritance anomaly, which requires code redefinition of inherited methods to maintain integrity of objects. In this paper, to solve the inheritance anomaly problems we introduce concept of state abstraction, in which internal states of encapsulated objects are made available from a part of object's external interface. And we design inheritance interface mechanisms which methods are inherited efficiently. In our scheme, we can solve the typical inheritance anomaly problems.

• Journal of Information Technology Services
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• v.16 no.2
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• pp.33-44
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• 2017

The financial industry recently introduces several issues for utilizing the blockchain technology as the core infrastructure of future finance. Blockchain, first introduced as the underlying technology of Crypto-currencies, Bitcoin is a technology that can ensure the integrity and reliability of data by verifying, recording, and storing data jointly in the network without a central administration organization or a manager. This blockchain has its potential power as a technology for issuing digital currencies, providing transparency, and securing record management, that is expected to be useful in the financial sector. At the same time, considering the characteristics of financial transactions which emphasize privacy, questions are raised about whether a blockchain structure in which information is distributed and shared among participants can be successful. How will we support to implement the potential of the blockchain in order to change the paradigm of the financial industry? How can we manage the side effects of blockchain effectively? Such a policy discussion is necessary. This study introduces the meaning of the blockchain technology, various utilization attempts, and possible problems facing technology from the viewpoint of financial industry, and suggests a policy direction for utilizing this technology as a catalyst to the progress of the financial industry or as a new technology power.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.88-96
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• 2018

Concerns over reliability assessments of the main components in nuclear power plants (NPPs) related to aging and continuous operation have increased. The conventional reliability assessment for main components uses experimental correlations under general conditions. Most NPPs have been operating in Korea for a long time, and it is predictable that NPPs operating for the same number of years would show varying extent of aging and degradation. The conventional reliability assessment does not adequately reflect the characteristics of an individual plant. Therefore, the reliability of individual components and an individual plant was estimated according to operating data and conditions. It is essential to reflect aging as a characteristic of individual NPPs, and this is performed through prognostics. To handle this difficulty, in this paper, the general path model/Bayes, a data-based prognostic method, was used to update the reliability estimated from the generic database. As a case study, the authors consider the aging for steam generator tubes in NPPs and demonstrate the suggested methodology with data obtained from the probabilistic algorithm for the steam generator tube assessment program.

• Nuclear Engineering and Technology
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• v.37 no.2
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• pp.201-206
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• 2005

Fretting, which is a special type of wear, is defined as small amplitude relative motion along the contacting interface between two materials. The structural integrity of steam generators in nuclear power plants is very much dependent upon the fretting wear characteristics of Inconel 690 U-tubes. In this study, a finite element model that can simulate fretting wear on the secondary side of the steam generator was developed and used for a quantitative investigation of the fretting wear phenomenon. Finite element modeling of elastic contact wear problems was performed to demonstrate the feasibility of applying the finite element method to fretting wear problems. The elastic beam problem, with existing solutions, is treated as a numerical example. By introducing a control parameter s, which scaled up the wear constant and scaled down the cycle numbers, the algorithm was shown to greatly reduce the time required for the analysis. The work rate model was adopted in the wear model. In the three-dimensional finite element analysis, a quarterly symmetric model was used to simulate cross tubes contacting at right angles. The wear constant of Inconel 690 in the work rate model was taken as $K=26.7{\times}10^{-15}\;Pa^{-1}$ from experimental data obtained using a fretting wear test rig with a piezoelectric actuator. The analyses revealed donut-shaped wear along the contacting boundary, which is a typical feature of fretting wear.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.