• Proceedings of the KIEE Conference
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• 2007.11b
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• pp.51-53
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• 2007

The nuclear power plant uses the steam which occurs from reactor and T/G the drive. By the T/G in consequence of the fact that the electricity which is produced the power and supplies in transmission system. But, recently the transmission and generation system are placed under deregulation situation from domestic and foreign, the maintenance control is difficult with the accident or the breakdown which relates is increasing. Hereupon, considering for effect to the reactor core against trip element which it does apply a probability concept from the NRC of the United States and it study and the recognition for the importance of the switchyard which is a power equipment which will be revaluated. Hereupon using the American example, the reliability establishment method which is suitable in domestic and it searches it does.

• Journal of Power System Engineering
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• v.10 no.3
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• pp.32-37
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• 2006

In this paper, we present the flow and heat transfer characteristics with the array of impinging jet nozzles by using the numerical computation and experiment. Numerical solutions were obtained for dimensionless gap H=6, dimensionless outlet length L=10 and Reynolds number Re=1500 by using the commercial CFD code, CFX-5. Experimental and numerical results were agreed well with each other. It was found that the impinging jet with circular array nozzles generated the uniform heat transfer area and the maximum heat transfer is higher than rectangular array nozzles for certain parameter sets. It is one of the most important utilities providing steam to turbine in order to supply mechanical energy in thermal power plant. It is composed of thousands of tubes for high efficient heat transfer.

• Journal of the Korean Society for Precision Engineering
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• v.19 no.8
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• pp.71-76
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• 2002

In this study low cycle fatigue (LCF) behavior of 12Cr steel at high temperature are described. Secondly, comparisons between predicted lives and experimental lives are made for the several sample life prediction models. Two minute hold period in either tension or compression reduce the number of cycles to failure by about a factor of two. Twenty minute hold periods in compression lead to shorter lives than 2 minute hold periods in compression. Experiments showed that life predictions from classical phenomenological models have limitations. More LCF experiments should be pursued to gain understanding of the physical damage mechanisms and to allow the development of physically-based models which can enhance the accuracy of the predictions of components. From a design point-of-view, life prediction has been judged acceptable for these particular loading conditions but extrapolations to thermo-mechanical fatigue loading, for example, require more sophisticated models including physical damage mechanisms.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2003.06a
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• pp.1458-1462
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• 2003

High temperature and high pressure heavy water flows through the pipes in atomic power plants. The curved parts of pipes are critical parts in that they change the direction of steam flow, and these parts are especially affected by severe wear. Therefore, most pipes in atomic power plants are tested by non-destructive examination by workers who use ultrasonic sensors to measure the wall thickness of pipes. But not only are these pipes located in a very dangerous environment, but the space is also very limited. For the safety of workers, it is necessary to design a device that uses a mobile robot that can inspect curved pipes. This paper presents the design and construction of a small device that can generate the necessary contact forces between ultrasonic sensors and pipe walls in a limited space. And a mobile robot is used in place ortho worker for successful non-destructive examination.

• Journal of Power System Engineering
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• v.14 no.6
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• pp.29-34
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• 2010

As the steam temperature of fossil power plant boiler is increasing, the use of 9Cr-1Mo high alloy material is prevalent and it is needed to investigate the characteristics of low cycle fatigue for high alloy and austenite stainless steel that has used up to recently. As a result of test, in 9Cr-1Mo high alloy steel, the relation of strain and fatigue life is non-linear and the crack mode of low cycle fatigue is brittle but in the austenite stainless steel, that of strain and fatigue life is linear and the crack mode of low cycle fatigue is ductile. Comparing the fatigue life between high alloy and austenite stainless steel, there is no consistent characteristics as to strains. But the fatigue life of 9Cr-1Mo steel is longer by 25% than that of STS304 stainless steel in the relatively low, 0.3% strain. In the other strain, the fatigue life of two materials is similar.

• Proceedings of the KIEE Conference
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• 1999.07g
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• pp.2926-2928
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• 1999

원자력발전소 증기발생기 전열관 건전성 유지를 위하여 매 주기마다 실시하고 있는 비파괴검사의 일종인 와전류검사(ECT, Eddy Current Testing)의 주요 공정은 크게 3가지로 분류할 수 있다. 첫 번째는 전열관 상태검사를 위한 신호데이터 취득공정이고, 두 번째는 취득된 신효를 판독하여 전열관의 건전성 여부를 진단하는 평가공정, 세 번째는 평가공정에서 발생하는 데이터를 토대로 전열관 이력 및 상태를 유지관리하는 공정으로 구분할 수 있다. 본 논문에서는 위의 세 번째 공정결과 생성되는 전열관 이력 및 상태자료를 데이터베이스화하여 유지 관리하고, 데이터베이스화된 내용을 바탕으로 전열관 상태 변화추이를 파악하는 기능, 현재까지 비 체계화된 모든 전열관의 이력자료를 다양한 보고서 형태로 출력할 수 있는 기능 둥을 제공하기 위한 "인트라넷 증기발생기 전열관 이력관리시스템"의 설계 및 구현과정 을 정 리 하였다.

• Journal of Welding and Joining
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• v.13 no.2
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• pp.12-24
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• 1995

Recently, more researches have been actively performed for the assessment of material degradation and residual-life of elevated temperature plant components, as some of domestic fossil power plants become older than 30 years. In this paper, results of on_site residual life assessment are reported for main steam pipes of Youngwol power station #2 which have operated since 1965. For critical weld locations such as butt welds branch welds, Y_sections and a T-section, replication technique and hardness measurement technique were employed for life_assessment. When cracks were detected by conventional NDT tests, crack growth life was calculated using a computer code. On the other hand, for matrix of pipes, residual life was quantitatively estimated by an analytic method and material degradation was estimated qualitatively using diameter measurement data and grain-boundary etching method. Also, direction in further improvement of on-site life assessment techniques are proposed.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.97-108
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• 2010

Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.3
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• pp.343-350
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• 2012

A planar solid oxide fuel cell (PSOFC) is studied in its application in a high-temperature stationary power plant. Even though PSOFCs with external reformers are designed for application from the distributed power source to the central power plant, such PSOFCs may sacrifice more system efficiency than internally reformed SOFCs. In this study, modeling of the PSOFC with an external reformer was developed to analyze the feasibility of thermal energy utilization for the external reformer. The PSOFC system model includes the stack, reformer, burner, heat exchanger, blower, pump, PID controller, 3-way valve, reactor, mixer, and steam separator. The model was developed under the Matlab/Simulink environment with Thermolib$^{(R)}$ modules. The model was used to study the system performance according to its configuration. Three configurations of the SOFC system were selected for the comparison of the system performance. The system configuration considered the cathode recirculation, thermal sources for the external reformer, heat-up of operating gases, and condensate anode off-gas for the enhancement of the fuel concentration. The simulation results show that the magnitude of the electric efficiency of the PSOFC system for Case 2 is 12.13% higher than that for Case 1 (reference case), and the thermal efficiency of the PSOFC system for Case 3 is 76.12%, which is the highest of all the cases investigated.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.3
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• pp.181-187
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• 2016

A thermal power plant has a heat exchanger tube to collect and convert the heat generated from the high temperature and pressure steam to energy, but the tubes are arranged in a complex manner. In the event that a leakage occurs in any of these tubes, the high-pressure steam leaks out and may cause the neighboring tubes to rupture. This leakage can finally stop power generation, and hence there is a dire need to establish a suitable technology capable of detecting tube leaks at an early stage even before it occurs. As shown in this paper, by applying acoustic emission (AE) technology in existing boiler tube leak detection equipment (BTLD), we developed a system that detects these leakages early enough and generates an alarm at an early stage to necessitate action; the developed system works better that the existing system used to detect fine leakages. We verified the usability of the system in a 560MW-class thermal power plant boiler by conducting leak tests by simulating leakages from a variety of hole sizes (ⵁ2, ⵁ5, ⵁ10 mm). Results show that while the existing fine leakage detection system does not detect fine leakages of ⵁ2 mm and ⵁ5 mm, the newly developed system could detect leakages early enough and generate an alarm at an early stage, and it is possible to increase the signal to more than 18 dB.