• Journal of the Korean Society of Visualization
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• v.19 no.3
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• pp.136-142
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• 2021

Submerged and semi-submerged vehicles expel cooling water through an outlet. In this process, induced noise and vibration by the flow around the outlet have been reported, and it may cause problems directly related to survivability of the navy vessels. The coolant outlet has a net-type structure and circular columns are mostly used. In this study, flow measurements using PIV and LDV were performed for different type outlets; conventional (flat plate with round bar) and improved (flat and flat plate) configurations. Experiments were conducted at a cavitation tunnel where pressure and steady flow rate conditions are ensured for sufficient time to measure the flow. The average velocity field of the outlets were measured and compared through LDV measurements, and instantaneous vorticities were evaluated through PIV measurements. The results show that the improved type of the outlet is advantageous in terms of flow stability compared to the conventional type of the outlet.

• 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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• 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.7
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• pp.2207-2220
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• 2021

Lead alloy is used as coolant in Lead-based cooled Fast Reactor (LFR). The natural characteristics of lead alloy are combined with the simple structural design of LFR. This constitutes the inherent safety characteristics of LFR. The main work of this paper is to take the main coolant pump (MCP) in the lead-cooled fast reactor (LFR) as the research object, and to study the flow pattern distribution of the internal flow field under the reverse rotation pump condition, the reverse rotation positive-flow braking condition and the reverse rotation negative-flow braking condition. In this paper, the double-outlet volute type and the space guide vane are selected as the potential designs of the CLEAR-I MCP. In this paper, the CFD method is used to study the potential reverse accident of the MCP. It is found that the highest flow velocity in the impeller appears at the impeller outlet, and the Q-H curves of the two design programs basically coincide. The space guide vane type MCP has better hydraulic performance under the reverse rotation positive-flow condition, the Q-H curves of the two designs gradually separate with increasing flow rate, and the maximum flow velocity inside the space guide vane type MCP is obviously lower than that of the double-outlet volute type. For the reverse rotation test of MCP, only the condition of the forward rotating pump of the main coolant pump is tested and verified. For the simulation of the MCP in LBE medium, it proved that the turbulence model and basic settings selected in the simulation are reliable.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3086-3091
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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 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 rectangular channel area of the APR1400 reactor vessel. As the water inlet area increased, the natural circulation mass flow rate asymptotically increased, that is, it converged at a specific value. And the circulation mass flow rate also increased as the outlet area, injected air flow rate, and outlet height increased. But the circulation mass flow rate was not changed along with the external water level variation if the water level was higher than the outlet height.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2183-2195
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• 2020

A mathematical model for the flowrate and rotation speed of RCP during idling was established. The numerical calculation method and dimensionless method were used to analyze the flow, head, torque and pressure and speed changes under idle conditions. Regularity, using the Q criterion vortex identification judgment method combined with surface flow spectrum morphology analysis to diagnose the vortex dynamic characteristics on RCP blade. On impeller blade, there is two oscillations in the pressure ratio on pressure surface in blade outlet region. The velocity on the suction surface is two times more oscillating than the inlet of blade, and there is an intersection with the velocity ratio curve on pressure surface. On blade of guide vane, the pressure ratio increases along the inlet to outlet direction, and the speed ratio decreases with the increase of idle time. There is a vortex that rotates counterclockwise on the suction surface, and the streamline on the suction surface of blade is subjected to the entrainment and blocking action of the vortex creates a large reverse flow in the main flow region. There are two vortices at the outlet of guide vane suction side and the vortices are in opposite directions.

• Journal of the Korean Society of Propulsion Engineers
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• v.24 no.5
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• pp.1-12
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• 2020

In this paper, the heat transfer characteristics of Jet A-1, which is used as a coolant and fuel in a regeneratively cooled thrust chamber, were experimentally studied. By varying the applied current for heating the cooling channel, the simulated specimen diameter, the specimen outlet pressure and the coolant flow rate, the wall temperatures of the specimen and the Jet A-1 temperatures at the specimen inlet/outlet were measured. It was found that the specimen diameter and the flow rate were important factors for the characteristics of heat transfer and the outlet pressure did not affect the performance of heat transfer. The results of the heat transfer experiments were compared with the previous Nusselt number empirical equations and novel Nusselt number empirical equations were finally derived.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1842-1852
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• 2019

The AP1000 reactor coolant pump is a vertical shielded-mixed flow pump, is the most important coolant power supply and energy exchange equipment in nuclear reactor primary circuit system, whose steadystate and transient performance affect the safety of the whole nuclear island. Moreover, safety demonstration of reactor coolant pump is the most important step to judge whether it can be practiced, among which software simulation is the first step of theoretical verification. This paper mainly introduces the fluid-solid coupling simulation method applied to reactor coolant pump, studying the feasibility of simulation results based on workbench fluid-solid coupling technology. The study found that: for the unsteady calculations of the pure liquid media, the average head of the reactor coolant pump with bidirectional fluid-solid coupling decreases to a certain extent. And the coupling result is closer to the real experimental value. The large stress and deformation of rotor under different flow conditions are mainly distributed on impeller and idler, and the stress concentration mainly occurs at the junction of front cover plate and blade outlet. Among the factors that affect the dynamic stress change of rotor, the pressure load takes a dominant position.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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• pp.407-408
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• 2002

This paper presents a split cooling system for a new inline 4-cylinder automotive engine. The split cooling system circulates coolant to the cylinder head and cylinder block separately. The coolant flow in the cylinder block is controlled by a $2^{nd}$ Thermostat installed at the outlet of cylinder block. The $2^{nd}$ thermostat closes when the coolant temperature is low. And this makes the coolant flow in cylinder block nearly stagnant, thereby reducing the coolant-side heat transfer coefficient and raising cylinder bore temperature. The $2^{nd}$ thermostat starts to open when the coolant temperature reaches a specified temperature. The test results on engine dynamometer show improved fuel economy and lower exhaust emission which result from the decrease in friction works and cooling loss. Also, several vehicle tests, with application of the new engine have been performed. Fuel economy improvement of 0.5{\sim}2.0%$ yields from different test modes and details are discussed in this paper.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.546-555
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• 2019

Since the first nuclear reactor first critical, nuclear systems has gone through four generations of history, and the fourth generation nuclear system will be truly realized in the near future. The notions of SVBR and lead-bismuth eutectic alloy coolant put forward by Russia were well received by the international nuclear science community. Lead-bismuth eutectic alloy with the ability of the better neutron economy, the low melting point, the high boiling point, the chemical inertness to water and air and other features, which was considered the most promising coolant for the 4th generation nuclear reactors. This study mainly focuses on the structural design optimization of the 4th-generation reactor coolant pump, including analysis of external characteristics, inner flow, and transient characteristic. It was found that: the reactor coolant pump with a central symmetrical dual-outlet volute structure has better radial-direction balance, the pump without guide vane has better hydraulic performance, and the pump with guide vanes has worse torsional vibration and pressure pulsation. This study serves as experience accumulation and technical support for the development of the 4th generation nuclear energy system.