• Nuclear Engineering and Technology
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• 제50권1호
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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.

• 한국분무공학회지
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• 제9권4호
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• pp.24-30
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• 2004

Feedwater flowing tube side of number 5 high pressure feedwatrr heaters was heated by extracting steam from high pressure turbine and draining water from moisture separators and number 6 high pressure feedwater heaters and supplied into steam generators. Because the extracting steam from the high pressure turbine is two phase fluid of high temperature, high pressure, and high speed and flows to inverse direction after impinging to impingement baffle. the shell wall of the number 5 high pressure feedwater heater may be affected by flow accelerated corrosion. On May 14, 1999, Point Beach Nuclear Plant (PBNP) with operating at full power experienced a steam leak from rupture of shell side of number 4B feedwater heater. Also, d domestic nuclear power plant experienced a severe wall thinning of shell side of number 5A and 5B feedwater heaters. This paper describes the fluid mixing analysis study using PHOENICS code in order to get at the root of the shell wall thinning of the feedwater heaters. The sections included in the fluid mixing analysis model are around the number 5h feedwater heater shell including the extracting pipeline. To identify the relation between the local velocities and wall thinning. the local velocities according to the analysis results were compared with the distribution of the shell wall thickness by ultrasonic test.

• KEPCO Journal on Electric Power and Energy
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• 제4권1호
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• pp.33-38
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• 2018

This study applied upgrades to the processes of a 10 MW wet amine $CO_2$ capture pilot plant and conducted performance evaluation. The 10 MW $CO_2$ Capture Pilot Plant is a facility that applies 1/50 of the combustion flue gas produced from a 500 MW coal-fired power plant, and is capable of capturing up to 200 tons of $CO_2$. This study aimed to quantitatively measure efficiency improvements of post-combustion $CO_2$ capture facilities resulting from process upgrades to propose reliable data for the first time in Korea. The key components of the process upgrades involve absorber intercooling, lean/rich amine exchanger efficiency improvements, reboiler steam TVR (Thermal Vapor Recompression), and lean amine MVR (Mechanical Vapor Recompression). The components were sequentially applied to test the energy reduction effect of each component. In addition, the performance evaluation was conducted with the absorber $CO_2$ removal efficiency maintained at the performance evaluation standard value proposed by the IEA-GHG ($CO_2$ removal rate: 90%). The absorbent used in the study was the highly efficient KoSol-5 that was developed by KEPCO (Korea Electric Power Corporation). From the performance evaluation results, it was found that the steam consumption (regeneration energy) for the regeneration of the absorbent decreased by $0.38GJ/tonCO_2$ after applying the process upgrades: from $2.93GJ/ton\;CO_2$ to $2.55GJ/tonCO_2$. This study confirmed the excellent performance of the post-combustion wet $CO_2$ capture process developed by KEPCO Research Institute (KEPRI) within KEPCO, and the process upgrades validated in this study are expected to substantially reduce $CO_2$ capture costs when applied in demonstration $CO_2$ capture plants.

• 비파괴검사학회지
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• 제5권2호
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• pp.58-70
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• 1986

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 추계학술발표대회 요약집
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• pp.211.1-211.1
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• 2003

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2002년도 유체기계 연구개발 발표회 논문집
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• pp.346-353
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• 2002

The steam generator feedwater flow rate in a nuclear power plant was estimated by means of artificial neural networks with the wavelet analysis for enhanced information extraction. The fouling of venturi meters, used for steam generator feedwater flow rate in pressurized water reactors, may result in unnecessary plant power derating. The backpropagation network was used to generate models of signals for a pressurized water reactor. Multiple-input single-output heteroassociative networks were used for evaluating the feedwater flow rate as a function of a set of related variables. The wavelet was used as a low pass filter eliminating the noise from the raw signals. The results have shown that possible fouling of venturi can be detected by neural networks, and the feedwater flow rate can be predicted as an alternative to existing methods. The research has also indicated that the decomposition of signals by wavelet transform is a powerful approach to signal analysis for denoising.

• 한국산학기술학회:학술대회논문집
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• 한국산학기술학회 2006년도 춘계학술발표논문집
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• pp.217-222
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• 2006

This paper describes a development of ultrasonic examination analysis software to analyze steam generator of nuclear power plant. The developed software includes classical analysis method such as A, B, C and D-scan images. This software provides the information of shape, depth, size and position of flaws. To do such, we obtain raw data from specimens and/or real pipeline of power plants and, modify the obtained ultrasonic 1-dimensional data according to prepared software design schedule. The developed analysis software is applied to specimens containing various flaws with known dimensions. The results of applications showed that the developed software provided accurate and enhanced images of flaws on various specimens.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.597-604
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• 2013

Containment Filtered Vent Systems (CFVSs) have been mainly equipped in nuclear power plants in Europe and Canada for the controlled depressurization of the containment atmosphere under severe accident conditions. This is to keep the containment integrity against overpressure during the course of a severe accident, in which the radioactive gas-steam mixture from the containment is discharged into a system designed to remove the radionuclides. In Korea, a CFVS was first introduced in the Wolsong unit-1 nuclear power plant as a mitigation measure to deal with the threat of over pressurization, following post-Fukushima action items. In this paper, the overall features of a CFVS installation such as risk assessments, an evaluation of the performance requirements, and a determination of the optimal operating strategies are analyzed for the Wolsong unit 1 nuclear power plant using a severe accident analysis computer code, ISAAC.

• 한국유체기계학회 논문집
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• 제16권6호
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• pp.26-31
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• 2013

Last blades of low-pressure turbine in nuclear power plant are the highly damaged part and always suffered from different types of loadings leading to various stress components, stresses due to centrifugal force and steam flow loading. Especially, centrifugal stress generated by turbine rotation is one of the main problems and more significant than other stresses as they have the greatest effect on total stress. Therefore, this study was performed to obtain the important information for estimation model development of the blade centrifugal stress level and distribution.

• 한국산학기술학회논문지
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• 제18권6호
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• pp.29-36
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• 2017

본 논문은 열병합 복합발전이란 하나의 프로세스에서 전기 또는 기계 동력과 열에너지의 두 형태를 생산하는 것이다. 가스터빈 열병합 발전 시스템의 각 구성부의 성능을 변수로 전체 시스템의 연료 소모와 각 구성부의 열과 전기의 성능을 표현하여야 한다. 전체시스템은 상부 시스템인 가스터빈 2대와 하부시스템인 열회수 증기발생기(HRSG) 2대, 증기터빈 1대, 지역난방열교환기 2대로 구성되어 있다. 가스터빈 열병합 복합발전시스템에서 가동시간 기준 10,000시간 후 성능시험을 각종 시험장치 설치 및 ASME PTC 46에 준한 성능시험으로 실시하였고, 발전소 전체의 종합출력과 효율에 대한 성능을 분석하였다. 이러한 성능시험 실시자료를 기초로 시험성능을 비교하여 성능변화 값을 확인하였다. 이 논문에서 가스터빈, 열회수 증기발생기, 증기터빈의 열역학적 시스템 해석을 통하여 이론적 결과 값을 산출하였다. 비교 대상은 전체 시스템의 생산열량과 대기로 배출되는 열량을 이론값과 실험값을 비교하였고, 전기출력 및 열 출력에 대한 효율을 이론값과 실제 값을 비교하였다. 가스터빈 열병합 복합발전소 성능 특성에 대한 시험결과를 열역학적 효율 특성과 비교하였으며, 0.3%의 오차를 보였다.