• 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.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.597-610
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• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.7 s.250
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• pp.707-714
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• 2006

A numerical analysis study has performed in terms of fluid dynamics to identify the wall thinning generated in the main feedwater isolation valve body of a nuclear power plant. To review the relations between flow characteristics and the wall thinning induced by flow accelerated corrosion (FAC), numerical analysis using FLUENT code and ultrasonic tests (UT) were performed. The local velocities according to the analysis results were compared with the distribution of the measured wall thickness by ultrasonic tests. The comparison results show that the local velocity in the x-direction had no correlation with the wall thinning but the local velocity in the y-direction and turbulence intensity had a great influence on that. These results provide a good match to those of the previous studies - locations colliding vertically against components undergo severe wall thinning. These results may be utilized to the design modification and the wall thinning management for main feedwater isolation valves for preventing the wall thinning degradation.

• Tribology and Lubricants
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• v.26 no.4
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• pp.230-239
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• 2010

Mechanical face seal installed in boiler feedwater pump prevents leakage of working fluid using thin fluid film between stator and rotor. If the leakage of working fluid exceeds the allowable volume, serious malfunction of boiler feedwater pump will be happen. The thinner fluid film exists between stator and rotor, the less working fluid leaks out. However, if the thickness of fluid film is not enough, the wear of seal face will be increased. And it causes the decrease in life of mechanical face seal. Therefore appropriate design is necessary to maximize the performance and life of mechanical face seal. In this study, numerical analysis using finite volume method was conducted to investigate the static characteristics of wavy mechanical face seals which have 4 different wavy surface profiles on rotor. As a result, opening force, leakage volume of working fluid and friction torque were obtained. For the same minimum film thickness, the static characteristics of mechanical face seal were affected by the wavy surface profile which can change the thickness of working fluid film and pressure distribution.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.32-40
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• 1994

A digital control system for the steam generator oater level control is developed using the optimal control technique. To describe the more realistic situation, a feedwater valve actuator of the first order lag is included in the overall control system. The optimal gains are obtained by the LQ method which imposes the constraints on the feedwater valve motion as well as on the deviation between the input demand signal and the output feedwater. Developed also is a Kalman observer on account of the flow measurement uncertainty at low power. And a digital controller on the feedback loop is designed which makes the system maintain the same stability margins for all power ranges. The simulation results show that the optimal digital system has good control characteristics despite the adverse dynamics of the steam generator at low power.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.61 no.3
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• pp.442-450
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• 2012

Almost domestic power plants are being operated by foreign distributed control system. Many korean power plants are being operated over their lifetime so they need to be retrofitted. So we are developing the distributed control system to solve this problem by our own technique. The simulator was already made to verify the reliability of the algorithms. The unit loop function tests of all algorithms were finished in the actual distributed control system for installation of power plant and their results were satisfactory. The unit loop function tests are for each unit equipment algorithm. So the total operation tests will be made with all algorithms together in the actual distributed control system to be applied to power plant. When the verification through all tests is finished, algorithms with hardware will be scheduled to be installed and operated in the actual power plant. This research result will contribute to the safe operation of the deteriorated power plant and korean electric power supply as well as domestic technical progress. This entire processes and results for the development are written for the example of boiler feedwater master algorithm out of all algorithms in this paper.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.1-11
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• 2015

Feedwater Heaters are important components in a nuclear power plant. As the age of heater increases, the maintenance cost required for continuous operation also increases. Most heaters have the carbon steel shells, tube support plates and flow baffles. The carbon steel is susceptible to flow-accelerated corrosion. This is especially true if the flow has a two-phase mixture of steam and condensate. The wall thinning around the wet steam entrance area of the shell is inevitable during some long term operation. The structural integrity of the feedwater heater shell affects the safe operation of the nuclear power plant. Therefore, it is needed for the thinned shell to be repaired. The maintenance method for preventing failure of the shell should be determined by investigating various factors including the stress distribution of thinned area. The stress analysis of the shell including the steam entrance region is studied in this paper. The results of thinned shell is compared with that of intact shell.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.730-742
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• 1995

The steam generator feedwater and level control system is designed by two steps of the feedwater control design and the feedback loop controller design. The feedwater sen system is designed by the optimal LQR/LQG approach and then is modified by the LTR method to recover the robustness. The plant characteristics are subject to change with the power variation and these dynamic properties are considered in the design of the feedback controller. All the designs are made in the continuous domain and are digitalized by applying the proper sampling period. The system is simulated for the two cases of power increase and decrease. From the results of simulation, it is found that the controller constants would rather be invariable during the power increase, while for the case of power decrease they should be changed with the power variation to keep the system stability.

• Journal of the Korea Society for Simulation
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• v.14 no.2
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• pp.35-43
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• 2005

Feedwater heaters of many nuclear power plants have recently experienced wall thinning damage, which will increase as operating time progresses. As it is judged that the wall thinning damages have generated due to local fluid behavior around the impingement baffle installed in downstream of the high pressure turbine extraction steam line to avoid colliding directly with the tubes, numerical analyses using PHOENICS code were performed for two models with original clogged impingement baffle and modified multi-hole impingement baffle. To identify the relation between wall thinning and fluid behavior, the local velocity components in x-, y-, and z-directions based on the numerical analysis for the model with the clogged impingement baffle were compared with the wall thickness data by ultrasonic test. From the comparison of the numerical analysis results and the wall thickness data, the local velocity component only in the y-direction, and not in the x- and z-direction, was analogous to the wall thinning configuration. From the result of the numerical analysis for the modified impingement baffle to mitigate the shell wall thinning, it was identified that the shell wall thinning may be controlled by the reduction of the local velocity in the y-direction.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2017-2022
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• 2004

There are multistage preheaters in the power generation plan to improve the thermal efficiency of the plant and to prevent the components from the thermal shock. The energy source of these heaters comes from the extracted two phase fluid of working system. These two-phase fluid can cause the so-called Flow Accelerated Corrosion(FAC) in the extracting piping and the bubble plate of the heater for example, in case of point Beach Nuclear Power Plant and in the Wolsung Nuclear Power Plant. The FAC is due to the mass transport of the thin oxide layer by the convection. FAC is dependent on many parameters such as the operation temperature, void fraction, the fluid velocity and pH of fluid and so on. Therefore, in this paper velocity was calculated by FLUENT code in order to find out the root cause of the wall thinning of the feedwater heaters. It also includeed in the fluid mixing analysis model are around the number 5A feedwater heater shell including the extraction pipeline. To identify the relation between the local velocities and wall thinning, the local velocities according to the analysis results were compared with distribution of the shell wall thicknes by ultrasonic test.