• 한국전산유체공학회:학술대회논문집
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• 2009.04a
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• pp.232-238
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• 2009

The objective of this study is to investigate the suitable design for a domestic CWP pump, which is used in cooling-water intakes for the unit 3 and 4 of Yeonggwang nuclear power plant. All the simulations are performed, using CFD method with a commercial code STAR-CCM+ version 3.02. After modeling a present design of the pump, the flow around the rotating blade was calculated by using quasi-static method and sliding mesh method with the almost same condition as an actual state. Based on fundamental simulations with various depth of sea water, the reference pressure for the boundary condition of the present study was decided. To verify the reliability of the calculation results, the suction flow rate of the data was compared with that of the experimental data. As a result of this comparison, it is confirmed that two results are fairly consistent. For the improvement of the suction flow rate, computational analysis was done by changing a flow channel and blade shapes. It is shown that the suction flow rate of the new pump was improved.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3612-3624
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• 2021

At present, in the case of pump fast optimization, there is a problem of rapid, accurate and effective prediction of cavitation performance. In "A Cavitation Performance Prediction Method for Pumps PART1-Proposal and Feasibility" [1], a new cavitation performance prediction method is proposed, and the feasibility of this method is demonstrated in combination with experiments of a mixed flow pump. However, whether this method is applicable to vane pumps with different specific speeds and whether the prediction results of this method are accurate is still worthy of further study. Combined with the experimental results, the research evaluates the sensitivity and accuracy at different flow rates. For a certain operating condition, the method has better sensitivity to different flow rates. This is suitable for multi-parameter multi-objective optimization of pump impeller. For the test mixed flow pump, the method is more accurate when the area ratios are 13.718% and 13.826%. The cavitation vortex flow is obtained through high-speed camera, and the correlation between cavitation flow structure and cavitation performance is established to provide more scientific support for cavitation performance prediction. The method is not only suitable for cavitation performance prediction of the mixed flow pump, but also can be expanded to cavitation performance prediction of blade type hydraulic machinery, which will solve the problem of rapid prediction of hydraulic machinery cavitation performance.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.41 no.1
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• pp.57-62
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• 2017

The purpose of this study is to develop a method for diagnosing the degree of gradual degradation of a reciprocating pump caused by continuous use as a water supply pump in a nuclear power plant. Normally, the progress of such degradation is too slow to be noticed. Hence, it is difficult to determine the degree of degradation using the existing diagnostic methods. In this paper, we propose a new method by which the normal state and the degraded state of the pump can be differentiated, so that the degree of degradation can be identified. First, an emulator was developed using FPGA by providing the parameters of the pump under normal state, so that the emulator generates the information of the pump in the healthy state. Then, by comparing this information with the parameters received from various output sensors of the emulator during the current state, it is possible to identify and measure the degree of gradual degradation. This paper presents some of the results obtained during the development process, and results that show how the emulator operates, by comparing the data collected from an actual pump.

• 유체기계공업학회:학술대회논문집
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• 2003.12a
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• pp.161-166
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• 2003

It is the common features of the integral reactors that the main components of the RCS are installed within the reactor vessel, and so there are no any flow pipes connecting the coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the RCS of the integral reactors, and it also makes impossible measure the flow-rate of the reactor coolant. As a alternative method, the method by the measurement of coolant pump power has been introduced in this study. Up to now, we did not found out a precedent which the coolant pump power is used for the real-time flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the real-time flow-rate calculation method by the measurement of coolant pump power in an integral reactor. As a result of the study, we could theoretically reason that the capacity-head curve and capacity-shaft power curve around the rated capacity with the high specific-speeded axial flow pumps have each diagonally steep incline but show the similar shape. Also, we could confirm the above theoretical reasoning from the measured result of the pump motor inputs, So, it has been concluded that it is possible to calculate the real-time flow-rate by the measurement of pump motor inputs. In addition, the compensation for a above new method can be made by HBM being now used in the commercial nuclear power plants.

• Plant Journal
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• v.9 no.2
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• pp.36-41
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• 2013

For areas installed with one pump, the trend for expected sizes of pump room areas is observed once pump power and floor dimensions are provided. However, these pump rooms with auxiliary charging pumps, turbine driven auxiliary feedwater pumps, and pump rooms with a separate valve room have unique ways to determine the pump room area. No definite trends are identified for areas installed with two pumps using pump power and floor dimensions. The relationship between pump power and floor dimensions is also unable to be found.

• Structural Engineering and Mechanics
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• v.51 no.6
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• pp.1005-1016
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• 2014

Roles of reactor internals are to support nuclear fuel, provide insertion and withdrawal channels of nuclear fuel control rods, and carry out core cooling. In case of functional loss of the reactor internals, it may lead to severe accidents caused by damage of nuclear fuel assembly and deterioration of reactor vessel due to attack of fallen out parts. The present study is to examine fluid flows in reactor internals subjected to hydrodynamic loads. In this context, an integrated model was developed and applied to two kinds of numerical analyses; one is to analyze periodic loading effect caused by pump pulsation and the other is to analyze random loading effect employing different turbulent models. Acoustic pressure distributions and flow velocity as well as pressure and temperature fields were calculated and compared to establish appropriate analysis techniques.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.296-302
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• 1996

The Loss of a Main Feedwater Pump test at 100% Power for YGN 4 was simulated in order to verify and validate the KSNPA. The comparison of the test data with the KSNPA prediction results showed reasonable agreement in the trends of the major plant parameters. All plant control systems including NSSS and T/G control systems are properly actuated and stabilized the plant conditions to a new steady state conditions in the KSNPA. From the comparison results, the KSNPA showed its capability to simulate the LOMFP event for the Korean Standard Nuclear Power Plant.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.128.2-128
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.121.1-121.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.132.2-132
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• 2001