• Nuclear Engineering and Technology
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• v.53 no.1
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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.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.20 no.2
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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.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2977-2982
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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 which effect on disclosing of the shell wall thinning of the high pressure feedwater heaters by porous plate.

• The KSFM Journal of Fluid Machinery
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• v.6 no.1 s.18
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• pp.57-65
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• 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.

• Proceedings of the SAREK Conference
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• 2009.06a
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• pp.841-846
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• 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.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2664-2669
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• 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.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.21 no.9
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• pp.483-495
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• 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.

• Journal of Energy Engineering
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• v.16 no.3
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• pp.103-112
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• 2007

Flood issue for nuclear power plants designed and built in 1970 is extremely severe for main steam header compartment and main feedwater line region of intermediate building and lower floor. A calculation for flood level at the main feedwater line isolation compartment is now performing by hand calculation. But, this methodology is quite conservative assumption. The goal of this study was to develop method to analyze flowrate using the RETRAN-3D computer code, and the developed method was applied to flood level analysis following main feedwater line break. As a result of analysis, flood level was low remarkably.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.15-19
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• 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
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• v.53 no.3
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• pp.1029-1040
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• 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.