• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.421-432
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• 2002

As one of the advanced design features of the APR1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, test results of direct ECC bypass performed in the steam-water test facility tailed MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) are presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’with the l/4.93 length scale . From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region are obtained.

• Transactions of the Korean Society of Automotive Engineers
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• v.19 no.6
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• pp.61-67
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• 2011

The cooling module needs enough space (or distance) from hood to absorb the energy from any pedestrian collision. Downsized cooling module for pedestrian protection is important to reduce the severity of pedestrian injury. When a vehicle collision happens, the downsized cooling module is required to reduce the risk of injury to the upper legs of adults and the heads of children. In this study, the performance of cooling module to cool the engine was investigated under 25% height reduction. The heat dissipation and pressure drop characteristics have been experimentally studied with the variation of coolant flow rate, air inlet velocity and A/C operation ON/OFF for the downsized cooling module. The results indicated that the cooling performance was about 94% level compared to that of the conventional cooling module. Therefore, we checked that the cooling module had good performance, and expected that the cooling module could meet the same cooling performance as conventional cooling module through optimization of components efficiency.

• Journal of the korean Society of Automotive Engineers
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• v.2 no.1
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• pp.39-50
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• 1980

This paper presents the variations obtained in heat flow rate and engine performance of a four-stroke cycle Diesel engine when there were changes in the temperature of cooling water, compression ratio, injection timing of fuel, and other factors. Heat dissipation of engine cylinder was calculated by the heat transfer coefficient of Nusselt's empirical equation and the analysis of distribution of temperature in cylinder barrel was obtained by the finite element method of two-dimensional steady state heat conduction. In this experiment, the out side temperature of cylinder liner was measured by the data logger, and the temperature distribution of liner was computed by the analysis of triangular finite element model under the assumption due to surface heat flux of cylinder inner surface. The results obtained by this study are as follows. Under the given operating condition, the temperature distribution of cylinder liner by using finite element method shows that the mean temperature of barrel is in accordance with the experimental results of Eichelberg and temperature difference is lower than 4.23.deg. C. The heat dissipation of engine decrease in accordance with the decrease of piston mean velocity, compression ratio, and the increase of coolant temperature. Influence on the delay of injection timing of fuel brings about the decrease of heat rejection over the cylinder at constant test conditions.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.817-824
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• 2014

To ensure the safety of research reactors, the water level must be maintained above the required height. When a pipe ruptures, the siphon phenomenon causes continuous loss of coolant until the hydraulic head is removed. To protect the reactor core from this kind of accident, a siphon breaker has been suggested as a passive safety device. This study mainly focused on two variables: the size of the pipe rupture and the timing of air entrainment. In this study, the size of the pipe rupture was increased to the guillotine break case. There was a region in which a larger pipe rupture did not need a larger siphon breaker, and the water flow rate was related to the size of the pipe rupture and affected the residual water quantity. The timing of air entrainment was predicted to influence residual water level. However, the residual water level was not affected by the timing of air entrainment. The experimental cases, which showed the characteristic of partical sweep-out mode in the separation of siphon breaking phenomenon [2], showed almost same trend of physical properties.

• The Transactions of the Korean Institute of Electrical Engineers D
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• v.55 no.3
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• pp.131-137
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• 2006

The secondary system of nuclear power plants consists of sophisticated piping systems operating in very aggressive erosion and corrosion environments, which make a piping system vulnerable to the wear and degradation due to the several chemical components and high flow rate (~10 m/sec) of the coolant. To monitor the wear and degradation on a pipe, the vibration signals are measured from the pipe with an accelerometer For analyzing the vibration signal the time-frequency analysis (TFA) is used, which is known to be effective for the analysis of time-varying or transient signals. To reduce the inteferences (cross-terms) due to the bilinear structure of the time-frequency distribution, an adaptive cone-kernel distribution (ACKD) is proposed. The cone length of ACKD to determine the characteristics of distribution is optimally selected through an adaptive algorithm using the normalized Shannon's entropy And the ACKD's are compared with the results of other analyses based on the Fourier Transform (FT) and other TFA's. The ACKD shows a better signature for the wear/degradation within a pipe and provides the additional information in relation to the time that any analysis based on the conventional FT can not provide.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.12 no.4
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• pp.6-13
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• 1998

It is difficult to control for the level of the steam generator in the nuclear power plant because there is swell and shrink, and many disturbance such as, feed water rate, feedwater temperature, main steam flow rte, coolant temperature effect steam generator level. If the conventional PI controller use in this system, we cannot have a stability in the control of the lower power, the rejection function of disturbance, and the load following effectively. In this paper, e study the application of the of neural network based Kp, Ti for Pi controller to the level control of the steam generator of nuclear power plant through the simulation and experimental on the steam generator. We are satisfied with the resulting against the inturrupt of the disturbance, the change of setpoint through the simulation and the swell and shrink, the response of controller on the experimental steam generator.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.908-917
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• 2020

Corrosion of structural materials presents a critical challenge in the use of lead-bismuth eutectic (LBE) as a nuclear coolant in an accelerator-driven system. By forming a protective layer on the steel surfaces, corrosion of steels in LBE cooled reactors can be mitigated. The amount of oxygen concentration required to create a continuous and stable oxide layer on steel surfaces is related to the oxidation process. So far, there is no oxidation experiment in fuel assemblies (FA), let alone specific oxidation detail information. This information can be, however, obtained by numerical simulation. In the present study, a new coupling method is developed to implement a coupling between the oxygen mass transfer model and the commercial computational fluid dynamics (CFD) software ANSYS-CFX. The coupling approach is verified. Using the coupling tool, we study the oxidation process of the FA and investigate the effects of different inlet parameters, such as temperature, flow rate on the mass transfer process.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1125-1139
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• 2023

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS). There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an incontainment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance. In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.

• Journal of Energy Engineering
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• v.17 no.2
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• pp.67-76
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• 2008

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at KAERI (Korea Atomic Energy Research Institute). Recently several integral effect tests for the reflood period of a LBLOCA (Large Break LOss of Coolant Accident) of the APR1400 have been performed with the ATLAS. In the APR1400 a high flow condition is changed to a low flow condition due to an fluidic device during an operation of the SIT. As the self-controlled fluidic device was not installed in the ATLAS, a set of characterization tests was performed to simulate its injection capability from the SIT for the APR1400 simulation. In the ATLAS the required SIT flow rate in the high flow condition was acquired by installing orifices with an optimized flow area to throttle the SIT discharge line and the low flow condition was achieved by changing the opening of the flow control valve in the SIT injection line. The test results showed that the safety injection systems of the ATLAS could simulate the required high and low flow rates of the SIT for the APR1400 simulation efficiently.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.18 no.4
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• pp.15-20
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• 2017

Currently, residential air conditioners operate as a heat pump during winter. In this case, the outdoor heat exchanger acts as an evaporator obtaining heat from cold air. On the other hand, it acts as a condenser during summer transferring heat to hot air. The outdoor temperature changes significantly from high to low. Generally, the air-side j and f factors are obtained at a standard outdoor temperature. Therefore, the applicability of the j and f factors under different outdoor conditions needs to be checked. In this study, tests were conducted for a two-row louver finned heat exchanger changing the outdoor temperature to subzero. The effects of the tube-side brine flow rate were also checked. The results showed that air-side j and f factors were essentially constant and independent of the outdoor temperature, suggesting that an extension of j and f factors obtained under standard conditions to a low outdoor temperature is acceptable. All j and f factors agreed within 9% and 3%, respectively. Tests were also conducted by changing the coolant flow rate. Both the j and f factors did not change according to the flow rate, suggesting that the tube-side heat transfer correlation is acceptable.