• Nuclear Engineering and Technology
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• 제53권1호
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• pp.322-336
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• 2021

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

• 한국유체기계학회 논문집
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• 제5권4호
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• pp.47-53
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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 back-propagation network was used to generate models of signals for a pressurized water reactor Multiple-input, single-output hetero-associative 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.

• 한국분무공학회지
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• 제12권1호
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• pp.18-29
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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 analysis results using the FLUENT code and the down scale experimental data in an effort to determine root causes of the shell wall thinning of the high pressure feedwater heaters. The numerical analysis and experimental data were also confirmed by actual wall thickness measured by an ultrasonic test.

• 대한기계학회논문집B
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• 제31권1호
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• pp.1-7
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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 analysis results using the FLUENT code and the down scale experimental data in an effort to determine root causes of the shell wall thinning of the high pressure feedwater heaters. The numerical analysis and experimental data were also confirmed by actual wall thickness measured by an ultrasonic test.

• 시스템엔지니어링학술지
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• 제14권2호
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• pp.58-66
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• 2018

This paper represents a systematic approach aimed at improving the performance of the proportional integral (PI) controller for the Advanced Power Reactor (APR) 1400 Feedwater Control System (FWCS). When the performance of the PI controller offers superior control and enhanced robustness, the steam generator (SG) level is properly controlled. This leads to the safe operation and increased the availability of the nuclear power plant. In this paper, a systems engineering approach is used in order to design a novel PI controller for the FWCS. In the reverse engineering stage, the existing FWCS configuration, especially the characteristics of the feedwater controller as well as the feedwater flow path to each SG from the FWCS, were reviewed and analysed. The overall block diagram of the FWCS and the SG was also developed in the reverse engineering process. In the re-engineering stage, the actual design of the feedwater PI controller was carried out using a genetic algorithm (GA). Lastly, in the validation and verification phase, the existing PI controller and the PI controller designed using GA method were simulated in Simulink/Matlab. From the simulation results, the GA-PI controller was found to exhibit greater stability than the current controller of the FWCS.

• 플랜트 저널
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• 제15권2호
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• pp.56-61
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• 2019

본 연구에서는 500 MW급 초임계압 석탄 화력발전소의 발전보일러용 고압 급수가열기에서 발생한 내부 튜브의 파손 사례 분석을 통해 운전 기록 모니터링에 의한 발전보일러용 고압 급수가열기 내부 튜브의 파손 예측 방안을 모색하고자 하였다. 이 연구를 통해 고압 급수가열기 내부 튜브 파손 시 쉘 측 수위 조절 밸브 개도와 보일러 급수펌프 흡입 유량의 변화로 내부 튜브 파손을 진단할 수 있는 예측 모형을 제안하였고, 제안된 예측 모형은 급수 계통의 불균형이 일어난 추가 사례를 통해 실증하였다. 이에 따라 본 연구와 유사한 특성의 발전보일러용 고압 급수가열기의 경우에도 쉘 측 수위 조절 밸브 개도와 보일러 급수펌프의 흡입 유량의 정상 운전 상태 값 대비 현재 운전 값 비교는 고압 급수가열기 내부 튜브의 파손에 대한 유력한 예측 진단 방안이 될 수 있다고 판단된다.

• 한국유체기계학회 논문집
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• 제2권2호
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• pp.19-26
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• 1999

Thermal output in a nuclear power plant is verified with calorimetric heat balance on the secondary plant. The calorimetry involves the precise measurement of the feedwater flow rate. However, the correct indication of feedwater flow rate obtained by a pressure-difference measurement across a venturi can be affected by instrument errors, fouling or a poorly developed velocity profile. This can result in an inaccurate mass flow rate and consequently an inaccurate estimate of power. The purpose of this study is to develop verification methods with accuracy better than $0.5\%$ for high precision flow measurement to be used for measuring feedwater flow rate. This chemical tracer method is a testing process that uses tracers which can be applied to quantify losses in electrical output due to the incorrect measurements of feedwater flow rate. And this system has good response to the variation of the flow rate. Accuracy of better than 0.5 percent can be expected for feedwater flow measurement, providing that the system can be stabilized during the test. This methodology is applicable to other flow systems well.

• 한국분무공학회지
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• 제15권2호
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• pp.83-93
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• 2010

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 inside feedwater heater installed downstream of the turbine extraction stream line. At that point, the extract steam from the turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows to reverse direction after impinging the impingement baffle, the shell wall of feedwater heaters may be affected by flow-accelerated corrosion. In this paper, to compare degree of shell wall thinning mitigation rate to squared type with mitigation rate of other type baffle plate, four different types of impingement baffle plate-squared, curved, mitigating type and multi-hole type-applied inside the shell. With these comparison data, this paper describes operation of experiments and numerical analysis which is composed similar condition with real feed water heater. And flow visualization is operated for verification of experiments and numerical analysis. In conclusion, this study shows that mitigating type and multi-hole type baffle plate are more effective than other baffle plate about prevention of pressure concentration and pressure value decrease.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2006년도 춘계학술대회논문집
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• pp.1290-1295
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• 2006

The 120 meters high pressure feedwater heater drain piping in nuclear power plant had been suffered by excessive vibration from the beginning of power generation. As time goes by, the piping vibration was beyond the allowable limit and an appropriate countermeasure was required to prevent the fatigue failure of the pipeline from the abnormal vibration. In this study, the vibrational characteristics of high pressure feedwater heater drain piping and the countermeasure for abnormal vibration were investigated. Among the several vibration reduction methods, the piping layout changed by making the smooth pipeline was applied to the high Pressure feedwater heater drain piping in nuclear Power plant. Applying the countermeasure, the vibration level was found to reduce over 54 percents and was satisfied under the allowable velocity at the full-power operation condition.

• 한국정밀공학회지
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• 제27권6호
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• pp.55-63
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• 2010

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 inside feedwater heater installed downstream of the turbine extraction stream line. At that point, the extract steam from the turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows to reverse direction after impinging the impingement baffle, the shell wall of feedwater heaters may be affected by flow-accelerated corrosion. In this paper, to compare degree of shell wall thinning mitigation rate to squared type with mitigation rate of other type baffle plate, three different types of impingement baffle plate-squared, curved and mitigating type-applied inside the shell. With these comparison data, this paper describes operation of experiments and numerical analysis which is composed similar condition with real feed water heater. And flow visualization is operated for verification of experiments and numerical analysis. In conclusion, this study shows that mitigating type baffle plate is more effective than other baffle plate about prevention of pressure concentration and pressure value decrease.