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

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

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2007년도 춘계학술대회B
• /
• pp.2977-2982
• /
• 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 which effect on disclosing of the shell wall thinning of the high pressure feedwater heaters by porous plate.

• 한국유체기계학회 논문집
• /
• 제6권1호
• /
• pp.57-65
• /
• 2003

The measurement uncertainties of an ultrasonic flowmeter were analyzed to evaluate its applicability to the measurement of the steam generator feedwater flow-rate in a nuclear power plant. The analyses of measurement uncertainties of a reactor power were also performed with the analyses of feedwater flow measurement uncertainties. Two ultrasonic flowmeters based on a cross-correlation technique and a transit time method were used in this study. The ultrasonic flowmeters were installed on a feedwater pipe line of a typical 1000 MWe Korea-standardized nuclear power plant to take the necessary data. The results have shown that the measurement uncertainties of the ultrasonic flowmeters are adequately smaller than those or a venturi meter. The research has also indicated that the measurement uncertainties of the reactor power based on the ultrasonic flowmeter uncertainties are sufficiently bounded by the uncertainty range usually assumed in nuclear safety analyses.

• 대한설비공학회:학술대회논문집
• /
• 대한설비공학회 2009년도 하계학술발표대회 논문집
• /
• pp.841-846
• /
• 2009

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 stream 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. This paper describes operation of experience and numerical analysis composed similar condition with real high pressure feedwater heater. This study applied several impingement baffle plates to feedwater heater same as previous study. In addition, it shows difference of pressure distribution and value between single phase and two phase based on experience and numerical analysis.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2008년도 추계학술대회B
• /
• pp.2664-2669
• /
• 2008

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damange, 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 stream 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 operation of experience and numerical analysis composed similar condition with real high pressure feedwater heater. This study applied squared, curved and new type impingement baffle plates to feedwater heater same as previous study. In addition, it shows difference of pressure distribution and value between single phase and two phase based on experience and numerical analysis.

• 설비공학논문집
• /
• 제21권9호
• /
• pp.483-495
• /
• 2009

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damange, 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 stream 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 operation of experience and numerical analysis composed similar condition with real high pressure feedwater heater. This study applied squared, curved and new type impingement baffle plates to feedwater heater same as previous study. In addition, it shows difference of pressure distribution and value between single phase and two phase based on experience and numerical analysis.

• 에너지공학
• /
• 제16권3호
• /
• pp.103-112
• /
• 2007

국내 1970년대에 설계 및 건설된 원자력발전소에 대해 침수분석을 수행한 결과 기기냉각수펌프 및 열교환기 건물, 주/보조건물, 중간건물 주증기 헤더 격실, 중간건물 주급수관 지역 및 하부층 등이 침수사고에 매우 취약하며 발전소 안전정지능력을 저해할 정도로 침수 영향이 심각한 것으로 판명되었다. 이들 지역에서의 침수원은 주급수관 파단이다. 현재 원자력발전소 내환경기기검증에서 주급수관 파단 방출량 계산은 수계산(Hand calculation)방법으로 Henry-Fauske 임계유량 모델 사용하고 있다. 이 방법은 배관파단 위치에서의 차압으로 계산되며, 실제 원자력발전소의 각종 제어로직에 의한 격리신호를 반영하지 못하므로 지나치게 보수적으로 파단 방출유량이 계산된다. 이러한 문제점을 개선하기 위해 원자력발전소 열수력계통 해석 전산코드인 RETRAN을 사용하여 원자력발전소 일/이차측 계통과 제어로직을 모사하고, 주급수관 파단 방출량 분석을 위한 입력가정과 해석방법을 개발하였다. 침수위 분석은 웨스팅하우스형 원자력발전소 격납건물 외부 하부격실에 대해 적용하였다. 전산코드 해석에서 각종 제어계통과 로직을 고려하였으며, 가장 제한적 사고조건을 계산하기 위해 노심출력, 파단형태, 면적, 위치 등의 조합으로 구성된 18개 사고 사례를 분석하였다. 그 결과 가장 제한적 사례 분석에서는 기존 수계산 분석에서보다 파단 방출유량이 크게 줄었고, 하부격실의 침수위도 상당히 낮아졌다.

• 한국압력기기공학회 논문집
• /
• 제9권1호
• /
• pp.15-19
• /
• 2013

The auxiliary feedwater is an important to remove the heat from the reactor core when the main feedwater system is unavailable. In most initiating events in Probabilistic Safety Assessment(PSA), the operaton of this system is required to mitigate the accidents. For one of domestic nuclear power plants, a design change of a turbine-driven auxiliary feedwater pump(TD-AFWP), pipe, and valves in the auxiliary system is implemented due to the aging related deterioration by long time operation. This change includes the replacement of the TD-AFWP, the relocation of some valves for improving the system availability, a new cross-tie line, and the installation of manual valves for maintenance. The design modification affects the PSA because the system is critical to mitigate the accidents. In this paper, the safety effect of the change of the auxiliary feedwater system is assessed with regard to the PSA view point. The results demonstrate that this change can supply the auxiliary feedwater from the TD-AFWP in the accident with the motor-driven auxiliary feedwater pump(MD-AFWP) unavailable due to test or maintenance. In addition, the change of MOV's normal position from "close" to "open" can deliver the water to steam generator in the loss of offsite power(LOOP) event. Therefore, it is confirmed that the design change of the auxiliary feedwater system reduces the total core damage frequency(CDF).

• Nuclear Engineering and Technology
• /
• 제53권3호
• /
• pp.1029-1040
• /
• 2021

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.