• 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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• 2007.05b
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• pp.2626-2631
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• 2007

Air-water two phase natural circulation flow in the T-HERMES (Thermo-Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow)-1D experiment has been evaluated to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5 results have shown that an increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not effective on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases. The water level is not effective on the water circulation mass flow rate. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it is not effective on the local pressure.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.31 no.5
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• pp.482-490
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• 2007

When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests and the simple analysis were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The calculated circulation flow rate was similar to experimental ones within about ${\pm}$15% error bounds and depended on the form loss due to the inlet/outlet area.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2017.05a
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• pp.1021-1025
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• 2017

Numerical heat/flow analysis was performed on a liquid rocket engine head with the cooling water manifold to ensure the durability of a ground test facility for heat exchanger. Through these studies, the shapes of the injector and the flow path were determined and applied to the head of the engine under development. Firing tests were conducted to verify the designed coolant manifold and no thermal damage was found on the engine-head-face. Comparing the combustion test results with the numerical analysis, the outlet temperature of coolant showed a difference of about $15^{\circ}C$. This trend is reasonable considering existence of LOX manifold, thermal barrier coating, and the actual location of flame.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2847-2858
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• 2021

When a nuclear reactor with rectangular fuel assemblies runs for a long time, impurities and debris may be taken into coolant channels, which may cause flow blockage, and the blocked fuel assemblies might be destroyed. Therefore, the purpose of this study is to perform a thermal-hydraulic analysis of a rectangular fuel assembly by STAR-CCM+, under the condition of one subchannel with 80% blockage ratio. A rectangular fuel assembly of the International Atomic Energy Agency (IAEA) 10 MW material test reactor (MTR) is chosen. In view of the gasket material taken into the coolant channel is close to the single side of the coolant channel, in the flow blockage accident of the Oak Ridge Research Reactor (ORRR), a new blockage category called single side blockage is attempted. The blockage positions include inlet, middle and outlet, and the blockage is set as a cuboid. It is found by simulations that the blockage redistributes the mass flow rate, and large vortices appear locally. The peak temperature of the cladding is maximum, when the blockage is located at the single side of the coolant channel inlet, and no boiling occurs in all blockage cases. Moreover, as the height of the blockage increases, the damage caused by the blockage increases slightly.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.17 no.10
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• pp.725-731
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• 2016

This study numerically investigates the thermal performance of a 2.0-kW butane-based combustion heating system for an electric vehicle under cold conditions. The system is used for cabin space heating and coolant-based battery thermal management. ANSYS CFX 17 software was used for parametric analysis. The mass flow rates of cold air and coolant were varied, and their effects were compared. The numerical results were validated with theoretical studies, which showed an error of 0.15%. As the outside air mass flow rates were increased to 0.005, 0.01, and 0.015 kg/s, the cabin supply air temperature decreased continuously while the coolant outlet temperature increased. When the coolant mass flow rates were increased to 0.005, 0.01 and 0.015 kg/s, the air temperature increased while the coolant outlet temperatures decreased. The optimal mass flow rates are discussed in a consideration of the requirements for high cabin heating capacity and efficient battery thermal management.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.318-327
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• 1998

The deuterium uptake behavior of Zr-2.5Nb pressure tubes in Wolsong Unit 1 was analyzed in terms of longitudinal location, operation time, and coolant temperature. The results were compared with those obtained from Canadian CANDU reactors. The amount of deuterium uptake was higher at the outlet part than at the inlet part and was also higher when subjected to a longer operation time and a higher coolant temperature. The hydrogen uptake of Zr-2.5Nb in a hydrogen gas atmosphere was dependent on the microstructure of the alloy. The aged Zr-2.5Nb consisting of $\alpha$-Zr and $\beta$-Nb phases showed higher hydrogen uptake than that consisting of $\alpha$-Zr and $\beta$-Zr phases. The hydrogen in the alloy decreased the rate of oxidation. This could be explained in terms of the cathodic controlled reaction of Zr-2.5Nb oxidation.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.435-440
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• 1996

Large Loss-of-Coolant Accidents analyzed in Final Safety Analysis Reports are usually covered by Reactor Inlet Header. Reactor Outlet Header and Primary Pump Suction breaks as representative cases. In this study we analyze the total (guillotine) break of an Emergency Core Cooling System (ECCS) pipe located at the ECCS injection point into the Primary Heat Transport System (PHTS). It was expected that thermal-hydraulic behaviors in the PHT and ECC systems are different from those of a Reactor Inlet Header break, having an equivalent break size. The main purpose of this study is to get insights on the differences occurred between the two cases and to assess these differences from the phenomenon behavior point of view. It was also investigated whether the ECCS line break analysis results could be covered by header break analysis results. The study reveals that as the intact loop has almost the same behavior in both analyzed cases. broken loop behavior is different mostly regarding sheath temperature in the critical core pass and pressure decrease in the broken Reactor Inlet Header. Differences are also met in the ECCS behavior and in event sequences timings.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.11 no.6
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• pp.201-207
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• 2012

This study was carried out to improve the design of heat exchanger for small marine diesel engine. As air pollutants emitted from small marine diesel engine become international problem, IMO(International Marine Organization) tried to establish severe regulations for NOx reduction. The formation of NOx is affected by cooling system, for instance, such as intercooler, heat exchanger, exhaust manifold, and therefore cooling systems are one of essential parts for design of small marine diesel engine. In this study, heat exchanger for small marine diesel engine was modeled and simulated using CATIA V5R19 and ANSYS FLUENT V.13. Thermal flow simulation for heat exchanger was performed to find the optimal design. As the results, maximum velocity of engine coolant in shell inside was 9.1m/s and it was confirmed that outlet temperature and temperature drop for engine coolant could be calculated by simulating proportional relations of temperature between engine coolant and sea water.

• Journal of the Korean Society of Fisheries and Ocean Technology
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• v.58 no.4
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• pp.359-366
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• 2022

In this study, the load fluctuation of the main engine is considered to be a disturbance for the jacket coolant temperature control system of the low-speed two-stroke main diesel engine on the ships. A nonlinear PID temperature control system with satisfactory disturbance rejection performance was designed by rapidly transmitting the load change value to the controller for following the reference set value. The feed-forwarded load fluctuation is considered the set points of the dual loop control system to be changed. Real-coded genetic algorithms were used as an optimization tool to tune the gains for the nonlinear PID controller. ITAE was used as an evaluation function for optimization. For the evaluation function, the engine jacket coolant outlet temperature was considered. As a result of simulating the proposed cascade nonlinear PID control system, it was confirmed that the disturbance caused by the load fluctuation was eliminated with satisfactory performance and that the changed set value was followed.