• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2141-2145
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• 2004

As part of study on thermal hydraulic behavior in the reactor cavity under external vessel cooling in the APR (Advanced Power Reactor) 1400, one dimensional two phase flow of steady state in the reactor cavity have been analyzed to investigate a coolant circulation mass flow rate in the annulus region between the reactor vessel and the insulation material using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that a two phase natural circulation flow of 300 - 600 kg/s is generated in the annulus region between the reactor vessel and the insulation material when the external vessel cooling has been applied in the APR 1400. An increase in the heat flux of the inner vessel leads to an increase of the coolant mass flow rate. An increase in the coolant outlet area leads to an increase in the coolant circulation mass flow rate, but the coolant inlet area does not effective on the coolant circulation mass flow rate. The change of the lower coolant outlet to a lower position affects the coolant circulation mass flow rate, but the variation trend is not consistent.

• 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.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.23 no.7
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• pp.831-840
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• 1999

In this work, an aerothermodynamic calculation model for cooled axial flow turbine blades with trailing edge ejection is suggested and a mean line performance analysis of a turbine stage with nozzle cooling is carried out. A unique model regarding the interaction between coolant and main gas is proposed, while existing correlations are adopted to predict viscous loss and blade outflow angle. The interactions considered are the heat transfer from main gas to coolant and the temperature and pressure losses by the mixing of two streams due to the trailing edge coolant ejection. For a stator blade without ejection, trailing edge loss calculated by the trailing edge analysis is compared with that calculated by loss correlation. The effect of heat transfer effectiveness of coolant passage on the mixing loss is analyzed. For a model turbine stage with nozzle cooling, parametric analyses are carried out to investigate the effect of main design variables(coolant mass flow ratio, temperature and ejection area) on the stage performance.

• Transactions of the Korean Society of Automotive Engineers
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• v.3 no.5
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• pp.17-23
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• 1995

A numerical analysis on engine coolant is made by the use of FVM based general purpose 3 dimensional Navier-Stokes solver, TURB-3D. Numerical solutions are verified by comparison with the experimental data of Lotus model. The results show a good qualitative as well as quantitative comparison. Coolant flow rate control is attempted through adjusting the cross section area of passage base on the results of an original coolant passage. It is concluded from the results that the flow rate control is possible as attempted, and thus can be used in the real engine design.

• Journal of the Semiconductor & Display Technology
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• v.21 no.1
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• pp.33-38
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• 2022

This study investigates the outlet temperature of coolant heater based on heat and flow volume conditions. Through computer simulation, the coolant temperature at the outlet was analyzed to investigate the heat and flow volume conditions of the coolant heater, and the optimal conditions were derived. Results show that heat and flow volume conditions, it was confirmed that heat condition is 0.424 W/mm³, and flow volume condition is 500 l/h, demonstrates optimal conditions. The results of this study can be utilized to efficiently control the coolant temperature through various heat and flow volume conditions.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.559-564
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• 2001

Advanced gas turbine engines employ turbine entry temperatures so high that cooling of the turbine blades is essential. The coolant flow introduces losses which need to be minimized, and therefore it is important that the minimum amount of coolant is used. This work presents the result of the one-dimensional analysis and the effect of the boundary conditions on coolant flow rate in gas turbine blades.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.13 no.9
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• pp.888-894
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• 2001

Advanced gas turbine engines employ turbine entry temperatures so high that cooling of the turbine blades is essential. The coolant flow introduces losses which need to be minimized, and therefore it is important that the minimum amount of coolant should be used. This work presents the result of the one-dimensional analysis and the effect of the boundary conditions on coolant flow rate in gas turbine blades.

• 유체기계공업학회:학술대회논문집
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• 1999.12a
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• pp.245-251
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• 1999

The structural analysis of a reactor coolant pump(RCP) of a nuclear power plant is very important for the safety assessment of the plant. Accurate boundary conditions for the heat transfer coefficient are required for reliable thermal stress analysis of the pump casing, especially in transient operations of the pump since the coolant properties are largely dependent on operational conditions. In the present study, a 3D mixed flow type coolant pump was modeled from the RCP drawings and analyzed in the steady state and number of transient flow conditions by using a commercial code STAR-CD. From the result of the computation, it is seem that the average heat transfer coefficients for the cases considered are found to be the suggested values of the manufacturer, Westinghouse Energy System. The unevenness in local heat transfer coefficients, however, is found to be considerable so that the use of average heat transfer coefficients in all boundaries might not give reliable thermal stresses.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1532-1539
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• 2019

This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.256-266
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• 1985

The loss of coolant flow accident is analyzed for the pressurized water reactor of Korea Nuclear Unit-1. The loss of coolant flow accident is classified into three types in accordance with its severity; partial loss of coolant flow, complete loss of coolant flow and pump locked rotor accident. Analysis has been carried out in three stages; system transient and average core analysis, DNBR calculation and hot spot analysis. The purpose of developing KTRAN is to simulate the transient fast. For the DNBR calculation, the thermal hydraulic codes, SCAN and COBRA IV-1, are adopted. And for the hot spot analysis, the fuel thermal transient code LTRAN is employed. This code system should be fast responding to the transient analysis. In case the transient occurs, severity comes within a couple of seconds. So response should be fast to accomodate the following sequence of the accident. Unfortunately this purpose could not be achieved by KTRAN. However, the calculated results are well comparable with FSAR results in range. Thereby, the effectiveness of KTRAN code analysis in this type of accident is proven.