• Nuclear Engineering and Technology
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• 제40권1호
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• pp.69-76
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• 2008

Most pressurized water reactors use Venturi flow meters to measure the feedwater flow rate. However, fouling phenomena, which allow corrosion products to accumulate and increase the differential pressure across the Venturi flow meter, can result in an overestimation of the flow rate. In this study, a soft-sensing model based on fuzzy support vector regression was developed to enable accurate on-line prediction of the feedwater flow rate. The available data was divided into two groups by fuzzy c means clustering in order to reduce the training time. The data for training the soft-sensing model was selected from each data group with the aid of a subtractive clustering scheme because informative data increases the learning effect. The proposed soft-sensing model was confirmed with the real plant data of Yonggwang Nuclear Power Plant Unit 3. The root mean square error and relative maximum error of the model were quite small. Hence, this model can be used to validate and monitor existing hardware feedwater flow meters.

• 한국분무공학회지
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• 제13권4호
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• pp.173-179
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• 2008

The current machinery and tools of secondary channel of the nuclear power plants were produced in the carbon-steel and low-alloy steel. What produced with the carbon-steel occurs wall thinning effect from flow accelerated corrosion by the fluid flow at high temperature, high pressure. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle-installed. Wall thinning by flow accelerated corrosion occurs piping system, the heat exchanger, steam condenser and feedwater heaters etc,. Feedwater heaters of many nuclear power plants have recently experienced sever wall thinning damage, which will increase as operating time progress. This study describes the comparisons between the numerical results using the FLUENT code and experimental data of down scale model.

• 대한기계학회논문집B
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• 제20권2호
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• pp.680-688
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• 1996

In this paper, the unsteady state calculational model is proposed for the thermal stratification analysis in the feedwater line of the PWR plant. By defining dimensionless parameters in the two-dimensional polar coordinate system and applying SIMPLE algorithm, the temperature and flow profiles due to the thermal stratification are obtained. Base on the fact that the most significant condition occurs when the fluid temperature difference between the piping ends reaches as high as 166.deg. C, the present result shows that max. Dimensionless temperature difference of 0.6 (about l00.deg. C) obtained between hot and cold sections of pipe wall at dimensionless time 47.0.

• 한국유체기계학회 논문집
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• 제12권4호
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• pp.54-62
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• 2009

The feedwater heaters are Critical components in a nuclear power plant. As the operation years of heaters go by, the maintenance costs required for continuous operation increase. When the carbon steel components in nuclear make contact with running fluid, the wall thinning caused by FAC (flow accelerated corrosion) can be generated. Local wall thinning is inevitable at the area around wet steam entrance to be attacked due to the long term operation. Sometimes the shell with thinned wall is eventually ruptured. To identify the relationship between the local wall thinning and fluid behavior of the feedwater heater, the practical data of a plant, which were based on ultrasonic thickness measurement tests, were analyzed and CFD(Computed Fluid Dynamics) analyses were performed.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2631-2636
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• 2008

The most components and piping of the secondary side of domestic nuclear power plants were manufactured carbon-steel and low-alloy steel. Flow accelerated corrosion leads to wall thinning (metal loss) of carbon steel components and piping exposed to the flowing water or wet steam of high temperature, pressure, and velocity. The feedwater heaters of many nuclear power plants have recently experienced sever wall thinning damage, which increases as operating time progress. Several nuclear power plants in Korea have also experienced wall thinning damage in the shell wall around the impingement baffle. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the experimental results based on down-scaled experimental facility. The experiments were performed based on several types of impingement baffle plates which are installed in low pressure feedwater heater.

• 한국수자원학회논문집
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• 제49권5호
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• pp.431-438
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• 2016

역삼투 해수 담수화 공법의 주요 단점으로 여겨지고 있는 에너지 소비량을 줄이기 위한 방편으로 해안선을 통하여 바다로 유출되는 담수 지하수를 활용하여 낮은 염도의 원수를 확보하는 방안을 제시하였다. 저 염도 지하 염수는 담수화 비용 뿐 아니라 표류 해수 취수의 알려진 단점을 극복하는 데도 유리하다. 지하 염수의 염도는 해안선을 통하여 바다로 유출되는 담수 지하수량의 영향을 크게 받는다. 본 연구에서는 담수화 시설에 필요한 수량을 최저 염도로 충족시킬 수 있는 지하 염수 관정의 위치 및 양수량 분포를 산정할 수 있는 최적 전산설계모델을 개발하였다. 해안 지역에서 지하 염수를 개발하면 대수층으로 해수가 추가 침투하여 다른 사용자의 지하수 관정을 오염시킬 수 있다. 본 모델은 지하 염수 관정의 최적 설계 시에 기존 지하수 관정을 해수 침투로부터 보호할 수 있도록 구성되었다.

• 설비공학논문집
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• 제19권12호
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• pp.821-830
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• 2007

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which will increase as operating time progresses. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle-installed downstream of the high pressure turbine extraction steam line- inside number 5A and 5B feedwater heaters. At that point, the extracted steam from the high pressure turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows in reverse direction after impinging the impingement baffle, the shell wall of the number 5 high pressure feedwater heater may be affected by flow-accelerated corrosion. This paper describes the comparisons between the numerical results using the FLUENT code and the down scale experimental data on effect of geometry of the impingement baffle plate on the shell wall thinning. Additionally, a new type impingement baffle plate was installed above the impingement baffle plate in the feedwater heater and then the numerical and experimental study were performed in the same progress.

• Nuclear Engineering and Technology
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• 제36권1호
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• pp.112-120
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• 2004

The purpose of this paper is to analyze the piping failure frequency for the main feedwater system in domestic nuclear power plants(NPPs) for the application to an in-service inspection(ISI), leak before break(LBB) concept, aging management program(AMP), and probabilistic safety analysis(PSA). First, a database was developed for piping failure events in domestic NPPs, and 23 domestic piping failure events were collected. Among the 23 events, 12 locations of wall thinning due to flow accelerated corrosion(FAC) were identified in the main feedwater system in 4 domestic WH 3-loop NPPs. Two types of the piping failure frequency such as the damage frequency and rupture frequency were considered in this study. The damage frequency was calculated from both the plant population data and damage(s) including crack, wall thinning, leak, and/or rupture, while the rupture frequency was estimated by using both the well-known Jeffreys method and a new method considering the degradation due to FAC. The results showed that the damage frequencies based on the number of the base metal piping susceptible to FAC ranged from $1.26{\times}10^{-3}/cr.yr\;to\;3.91{\times}10^{-3}/cr.yr$ for the main feedwater system of domestic WH 3-loop NPPs. The rupture frequencies obtained from the Jeffreys method for the main feedwater system were $1.01{\times}10^{-2}/cr.yr\;and\;4.54{\times}10^{-3}/cr.yr$ for the domestic WH 3-loop NPPs and all the other domestic PWR NPPs respectively, while those from the new method considering the degradation were higher than those from the Jeffreys method by about an order of one.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2005년도 창립60주년 기념 추계 학술대회
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• pp.79.2-79.2
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• 2005

• Nuclear Engineering and Technology
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• 제31권2호
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• pp.192-201
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• 1999

This paper proposes an improved signal processing strategy for accurate feedwater flowrate estimation in nuclear power plants. It is generally known that ∼2% thermal power errors occur due to fouling Phenomena in feedwater flowmeters. In the strategy Proposed, the noises included in feedwater flowrate signal are classified into rapidly varying noises and gradually varying noises according to the characteristics in a frequency domain. The estimation precision is enhanced by introducing a low pass filter with the wavelet analysis against rapidly varying noises, and an autoassociative neural network which takes charge of the correction of only gradually varying noises. The modified multivariate stratification sampling using the concept of time stratification and MAXIMIN criteria is developed to overcome the shortcoming of a general random sampling. In addition the multi-stage robust training method is developed to increase the quality and reliability of training signals. Some validations using the simulated data from a micro-simulator were carried out. In the validation tests, the proposed methodology removed both rapidly varying noises and gradually varying noises respectively in each de-noising step, and 5.54% root mean square errors of initial noisy signals were decreased to 0.674% after de-noising. These results indicate that it is possible to estimate the reactor thermal power more elaborately by adopting this strategy.