• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.1 no.1
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• pp.32-45
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• 1989

The airflow characteristics of clean tunnel type clean room were experimentally investigated with the change of operating speed and exit type. Distributions of air velocity and pressure were measured in clean room which is located lower than HEPA filters, and the pressure distribution was also measured in upper plenum which is located above the HEPA filters, to identify the performance of clean room. Through the analysis it was turned out that air velocity characteristics in clean room were significantly affected by the upper plenum flow conditions such as pressure distribuion. This results will be useful in the actual clean room design to enhance the performance.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.239-239
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.543-544
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.1075-1076
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.576-577
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• 2005

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Proceedings of the KSME Conference
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• 2000.04b
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• pp.795-800
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• 2000

A CCFL(Counter Current Flow Limit) test have been performed in narrow annular gaps with large diameter, because it has been confirmed that the CCFL phenomena affected the critical power in hemispherical narrow gap geometries from the SONATA(Simulation Of Naturally Arrested Thermal Attack)-IV (In-Vessel)/VISU(Visualization)-II experiments. The objectives of the CCFL experiments are to investigate the small gap sizes(1, 2mm) effect on CCFL under the large diameter condition and to confirm the findings of the VISU-II study that global dryout in hemispherical narrow gaps was induced by the CCFL. The test section was made of acrylic resin to allow visual observation on the two-phase flow behaviors inside annular gaps. It was observed from visualization that a part of water supplied was accumulated in the upper plenum and a significant increase in the differential pressure across the gap was occurred, which was the definition of the CCFL occurrence in this experimental study. From the experimental results in annular gap with large diameter it can be known that an increase in the differential pressure was not big at small air flow-rates. When the CCFL was occurred, the differential pressure across gaps was increased significantly and a water accumulated in the upper plenum. The occurrence of CCFL was correlated using the Wallis parameter.

• Nuclear Engineering and Technology
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• v.39 no.1
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• pp.43-50
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• 2007

A partial flow blockage in an assembly of a liquid metal reactor could result in a cooling deficiency of the core. To develop a partial blockage detection system, we have studied the changes of the temperature fluctuation characteristics in the upper plenum according to changes of the t10w blockage conditions in an assembly. We analyzed the temperature fluctuation in the upper plenum with the Large Eddy Simulation (LES) turbulence model in the CFX code and evaluated its statistical parameters. Based on the results of the statistical analyses, we developed a neural network model for detecting a partial flow blockage in an assembly. The neural network model can retrieve the size and the location of a flow blockage in an assembly from a change of the root mean square, the standard deviation, and the skewness in the temperature fluctuation data. The neural network model was found to be a possible alternative by which to identify a flow blockage in an assembly of a liquid metal reactor through learning and validating various flow blockage conditions.

• Transactions of the Korean Society of Mechanical Engineers
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• v.17 no.8
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• pp.2110-2121
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• 1993

A numerical investigation has been carried out for the flow characteristics after exchange of some filters from the original layer to the new low pressure loss layer with equal filtering efficiency. The solution domain includes upper plenum, filter layer, clean space, access panels, and lower plenum. The concept of the distributed pressure resistance was applied to describe the momentum loss in filter layer and access panels. The evolution of the flow field is simulated using the low Reynolds number k-.epsilon. over bar turbulent model and SIMPLE algorithm based on the finite volume method. As a result, after the exchange of filter layer the power requirement can be reduced by 8-9 percent. The results also demonstrate that the perpendicularity of the flow near access panels may become worse at new filter layer than origianl filter layer. But the situation can be recovered by adjusting the jopening ratio of access panels.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1821-1829
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• 2023

The temperature distribution of the reactor pool under natural circulation induced by the RVACS operation was experimentally studied. According to the Bo' based similarity law, which could reproduce the temperature distribution of the working fluid under natural circulation, SINCRO-2D facility was designed based on the PGSFR. It was reduced to 1 : 25 in length scale, having water as a simulant of the sodium, which is the original working fluid. In general, temperature was stratified, however, effect of the natural circulation flow could be observed by the entrainment of the stratified temperature. Relative cooling contribution of the upper plenum (narrow gap) and lower plenum was approximately 0.2 and 0.8, respectively. In the range of decay heat from 0.2% to 1.0%, only the magnitude of the temperature was changed, while the normalized temperature maintained. Boundary temperature distribution change made a global temperature offset of the pool, without a significant local change. Therefore, the decay heat and cooling boundary condition had no significant effect on temperature distribution characteristics of the pool within the given range of the decay heat and boundary temperature distribution.