• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.404-410
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• 1998

In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold leg temperature was very close to the technical specification limit of 298$^{\circ}C$ during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2009.11a
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• pp.119-122
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• 2009

The purpose of this study is to design of film-cooling ring for the small thrust rocket engine using green propellants(Hydrogen peroxide and kerosene). Cold flow test was carried out to measure the mass flow rate and atomizing characteristic. Required mass flow rate was obtained from thermal analysis of the engine, and measured flow rate 42.25g/s was in the range of permissible coolant flow rate. With the same mass flow rate, cooling ring with more hole and high velocity shows better spray pattern. The result of thermal analysis, cooling ring has enough cooling performance.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.7 no.1
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• pp.89-97
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• 1995

Experimental study was performed for free convection and ice formation around two horizontal circular tubes which were placed vertically. Temperature and velocity distributions were visualized with real time holographic interferometry technique and tracer method. When water was cooled, super cooled region was formed around cooling pipe. It was found that flow induced by free convection always directed downwards when the coolant temperature was low, while it directed upwards when the coolant temperature was comparably high though it directed downwards initially. Flow phenomena with free convection were investigated in detail with varying cooling rate and length between cooling pipes. And growing process of dense ice was also investigated. Dendritic ice is suddenly formed within a supercooled region, and a dense ice layer begins to develop from the cooling wall.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.217-218
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• 2005

• Transactions of the Korean Society of Automotive Engineers
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• v.20 no.5
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• pp.95-101
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• 2012

An electric water pump for engine cooling system has an advantage which particularly in the cold start, the use of the electric water pump saves fuel and leads to a corresponding reduction in emissions. The canned type electric water pump without mechanical sealing elements was selected to meet the requirements for operational reliability and life. However, the electric water pump for internal combustion engine generates much more heat loss than for hybrid electric vehicle since it is operated by the electric power of high current and low voltage. In this study, the fluid flow and thermal characteristics of the canned type electric water pump as an inverter integrated water pump has been investigated under the effects of heat generation. The analysis conditions such as outdoor air temperature of $125^{\circ}C$, water pump speed of 6000 rpm, coolant temperature of $106^{\circ}C$ and coolant flow rate of 120 L/min was used as a standard condition. Therefore, flow fields and temperature distribution inside the water pump were obtained. Also, we checked the feasibility of the canned type for the electric water pump in comparison with the mechanical seal type.

• Transactions of the Korean Society of Mechanical Engineers
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• v.10 no.1
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• pp.7-14
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• 1986

To predict the fuel clad temperature during the reflooding phase of a LOCA, one may need a knowledge of reflood heat tranfer mechanism in a rod bundle. For this purpose reflooding experiments have been carried out with an electrically heated 3*3 rod bundle. Using the method for the determination of local heat transfer coefficient from the measured wall temperature the parametric effects of coolant flow rate, initial wall temperature, coolant subcooling and heat generation rate on the propagation of rewetting front were investigated. Prediction of the wall temperature histories for these experiments was discussed using REFLUX code with modification of the rewetting temperature correlation. Through this modification, better agreement between experiment and prediction was obtained.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2010.11a
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• pp.601-604
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• 2010

An experimental study was carried out to investigate the effect of film cooling in a liquid rocket engine using Hydrogen peroxide/Kerosene as propellants. The heat fluxes were calculated by the measured wall temperatures on the axial direction of thrust chamber for mass flow rate of coolant and different type of film cooling rings. The flow rate of coolant was 0~20 percent of the total propellant.

• 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.37 no.4
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.