• KEPCO Journal on Electric Power and Energy
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• 제5권4호
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• pp.323-330
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• 2019

500 MW 표준 석탄화력발전소는 국내에서 가장 큰 용량의 규격화된 발전소로써 20년 넘게 국내 전력생산에 중추적인 역할을 수행하고 있다. 장기간 사용으로 인한 경년 열화와 더불어 최근 석탄화력발전소의 대기오염 문제가 대두되면서 석탄화력발전소 가동률 제한 정책에 따른 잦은 기동·정지에 의해 발전 설비의 고장 확률이 증가하고 있다. 그 중 증기 배관은 보일러에서 만들어진 고온·고압의 증기를 전력생산을 위해 터빈으로 이송시키는 중요한 역할을 하는 설비로 최근 국내 대용량 발전소 증기 배관의 고장 사례가 빈번하게 발생하고 있다. 이에 본 연구에서는 국내 500 MW 표준 석탄화력발전소 주증기배관 연결 용접부에 반복적으로 발생된 손상에 대해 손상 해석을 수행하였다. 동일 규격의 타 발전소에서 발생될 수 있는 고장의 사전 예방을 위해 균열부 금속 조직 분석과 배관 응력 해석을 통해 배관 지지 구조에 의한 고 응력에 의해 발생된 원인을 규명하고 고 응력부 응력 저감을 위한 지지 구조 개선 방안을 제시하였다.

• 한국압력기기공학회 논문집
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• 제13권2호
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• pp.31-37
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• 2017

The heat transfer function or thermal performance is the most important function of the steam generator component in nuclear power plants. The declining of thermal performance, fouling does not affect the electric power of the nuclear power plant within a certain fouling level, but it affects the output when goes beyond the governor valve wide open of the turbine. The VWO steam pressure can be predicted through the thermal performance evaluation of steam generators in the nuclear power plant. In consideration of the fouling characteristics of the steam generator, methods of the thermal performance evaluation and fouling cases are reviewed, and also the critical VWO value is estimated through the actual thermal performance evaluation. It is necessary to apply the VWO theory based on the thermal performance of the steam generators.

• KEPCO Journal on Electric Power and Energy
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• 제2권4호
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• pp.583-588
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• 2016

The high-temperature steam pipes of thermal power plants are subjected to severe conditions such as creep and fatigue due to the power plant frequently being started up and shut down. To prevent critical pipes from serious damage and possible failure, inspection methods such as computational analysis and online piping displacement monitoring have been developed. However, these methods are limited in that they cannot determine the life consumption rate of a critical pipe precisely. Therefore, we set out to develop a life assessment system, based on a three-dimensional piping displacement monitoring system, which is capable of evaluating the life consumption rate of a critical pipe. This system was installed at the "M" thermal power plant in Malaysia, and was shown to operate well in practice. The results of this study are expected to contribute to the increase safety of piping systems by minimizing stress and extending the actual life of critical piping.

• 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.

• 한국압력기기공학회 논문집
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• 제13권1호
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• pp.54-60
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• 2017

When a sudden rupture occurs in high energy lines, ejection of inner fluid with high temperature and pressure causes blast wave as well as thrust forces on the ruptured pipe itself. The present study is to examine pipe whip behaviors and blast wave phenomena under postulated pipe break conditions. In this context, typical numerical models were generated by taking a MSL (Main Steam Line) piping, a steam generator and containment building. Subsequently, numerical analyses were carried out by changing break locations; one is pipe whip analyses to assess displacements and stresses of the broken pipe due to the thrust force. The other is blast wave analyses to evaluate the broken pipe due to the blast wave by considering the pipe whip. As a result, the stress value of the steam generator increased by about 7~21% and von Mises stress of steam generator outlet nozzle exceeded the yield strength of the material. In the displacement results, rapid movement of pipe occurred at 0.1 sec due to the blast wave, and the maximum displacement increased by about 2~9%.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.183-190
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• 1996

Leak-before-break(LBB) approach has been shown to be both cost and risk effective by reducing maintenance cost and occupational exposure when applied to high energy piping in nuclear power plants. For Korean Next Generation Reactor(KNGR) development, LBB is considered for the Main Steam Line(MSL) piping inside containment. Unlike the reactor coolant piping leakages which can be detected by particulate and gaseous radiation monitoring, main steam line leak detection systems must be based on principles that do not involve radioactivity. Ceramics are widely used as humidity sensor materials which can be further developed for nuclear applications. In this paper, we describe the progress in the development of ceramic humidity sensors for use with the main steam lines of KNGR.

• 한국압력기기공학회 논문집
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• 제7권4호
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• pp.51-56
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• 2011

Steam generator is one of the main equipments that affect safety and long term operation in nuclear power plants. Fluid flows inside and outside of the steam generator tubes and induces vibration. To prevent the vibration the tubes are supported by AVB (anti vibration bar). When the steam generator tube contact to AVB, it is damaged by the accumulation of wear and corrosion. Therefore studies are required to determine the effects of the gap between the steam generator tube and AVB. In order to obtain the stress and the displacement distributions of the steam generator tube, three dimensional finite element analyses were performed by using the commercial program ANSYS. Using the calculated the stress and the displacement distributions, the static residual strength of the steam generator tube can be evaluated. The results show that the stress and displacement of the steam generator tube increase significantly compared with the results from a zero-gap model.

• Corrosion Science and Technology
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• 제12권5호
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• pp.227-232
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• 2013

A number of components installed in the secondary system of nuclear power plants are exposed to aging mechanisms such as FAC (Flow-Accelerated Corrosion), Cavitation, Flashing, and LDIE (Liquid Droplet Impingement Erosion). Those aging mechanisms can lead to thinning of the components. In April 2013, one (1) inch small bore piping branched from the main steam line experienced leakage resulting from wall thinning in a 1,000 MWe Korean PWR nuclear power plant. During the normal operation, extracted steam from the main steam line goes to condenser through the small bore piping. The leak occurred in the downstream of an orifice. A control valve with vertical flow path was placed on in front of the orifice. This paper deals with UT (Ultrasonic Test) thickness data, SEM images, and numerical simulation results in order to analyze the extent of damage and the cause of leakage in the small bore piping. As a result, it is concluded that the main cause of the small bore pipe wall thinning is liquid droplet impingement erosion. Moreover, it is observed that the leak occurred at the reattachment point of the vortex flow in the downstream side of the orifice.

• 한국소음진동공학회논문집
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• 제25권9호
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• pp.599-605
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• 2015

U.S. Nuclear Regulatory Commission(NRC), Regulatory Guide(RG) 1.20(Rev.3, 2007년)은 원자로 냉각재계통, 주증기, 주급수 및 복수시스템의 주요 배관 및 기기에 대하여 압력변동 및 진동에 의한 잠재적 유해효과에 대한 평가를 요구한다. 그러나 증기발생기와 연결된 주증기, 주급수 및 복수시스템의 주요 배관 전체에 대하여 상세 해석하는 것은 매우 복잡하여 한계가 있다. 이 논문은 APR1400 원전의 종합진동평가(comprehensive vibration assessment program, CVAP)를 수행하기 위하여 증기발생기에 연결된 2차측 주요 배관의 음향공진과 펌프유발진동을 위한 간이평가 방법에 관한 것이다. 이 논문에서는 이러한 배관시스템의 잠재적 진동 원인이 무엇인지, 음향공진과 펌프유발진동의 가능성을 예방하기 위한 간이평가 방법은 무엇인지를 고찰하고자 한다. 이 논문은 APR1400 원전 증기발생기와 연결된 주증기 및 주급수 배관의 유동유발진동 간이평가를 위해 사용될 것으로 판단된다.

• 한국압력기기공학회 논문집
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• 제7권2호
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• pp.26-33
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• 2011

To evaluate steam oxidation behaviours of Alloy 617 and Haynes 230, oxidation test were performed at $900^{\circ}C$ in steam and $steam+20\;vol.-%\;H_2$ environments. Oxidation rate in steam condition was similar to that in air for Alloy 617, while it was slightly lower for Haynes 230. When hydrogen was added to steam, oxidation rate was enhanced. Isolated $MnTiO_3$ particle were formed on $Cr_2O_3$ oxide layer and sub layer $Cr_2O_3$ were formed in steam and $steam+20\;vol.-%\;H_2$ for Alloy 617. On the other hands, $MnCr_2O_3$ layer were formed on top of $Cr_2O_3$ oxide layer for Haynes 230. The extensive sub layer $Cr_2O_3$ formation was resulted from the oxygen inward diffusion in such environments. When hydrogen was added, the oxide morphology was changed from polygonal to platelet because of the accelerated diffusion of cations under the oxide layer. In addition, decarburized zone was extended as hydrogen participated into the reactions causing carbide dissolution.