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Song, Chul Hwa
Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)1995
Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)1995
AbstractAbstract
[en] An experimental and analytical work is performed to investigate the relation between the developing phenomena in bubble flow and the propagation phenomena of void waves. For this purpose, the structural developments in bubble flow and the propagation property of void waves are measured over a broad range of flow conditions including the bubble-to-slug flow regime transition (BSFRT) region. And a linear stability analysis is performed, based on the two-fluid model, to establish the analytical model on the wave propagation parameters, and the predictability of the model is validated by comparing analytical results with experimental observations. In the experimental work, an impedance void meter is developed to measure the void fraction, and a series of test are performed by varying the bubble size in order to investigate the bubble size effect on the bubble flow structures for various flow conditions. Statistical signal processing techniques are applied to void signals in order to objectively identify the changing modes of bubble flow structures and to estimate the wave propagation properties. The impedance void meter developed in this study showed very good temporal and spatial resolutions enough to identify the developing phenomena in bubble flow structures and to investigate the void wave propagations, and the void distribution effect could be minimized by electrically shielding the guard electrodes. It was also designed so that the inherent errors due to the phase shifts between channels be negligible. Various features occurred in the transitional process of bubble flow could be objectively identified by introducing some statistical parameters evaluated from void signals. Two distinct modes of structural development in bubble flow were observed in the transitional process, and they are found to be much influenced by the initial bubble size. And the mechanism to govern BSFRT could be characterized by two ways depending on the developing modes of bubble flow structures. The transitional process for large bubble case can be explained by the wake model, whereas, for small bubble case, by the bubble coalescence model. It is clarified that the diversity in the wave propagations is closely related to the developing modes of bubble flow structures. And the different features shown in previous works on the void wave propagations could be systematically explained by the present observations. It turns also out that the instability criterion of void waves correctly indicates the appearance of large structures of gas phase. Thus, it is proposed that the spatial attenuation factor, which is introduced to quantify the degree of wave damping, can be used to objectively identify the BSFRT boundary. In the analytical work, generalized form of the wave propagation properties, which are dependent on the wave number, is derived from the linear stability analysis, and from that, the wave damping phenomena could be analytically predicted. It is shown that the analytical model on the wave dispersion, including the wave damping, can predict qualitatively well the experimental observations. The spatial gain factor is proposed to use as an indicator to analytically quantify the degree of spatial damping, and it will be very useful for relating both analytical and experimental data because it is a measurable quantity by experiments. The higher derivatives in the governing equations introduce, in general, a dependency of the wave parameters and the stability condition on the wave number, but the wave number-dependency of the stability condition could be eliminated in the case of long wavelength limit. It is also shown that considering algebraic terms only as momentum source terms and assuming the equal phasic pressure lead to the ill-posed problem. From both the characteristics and linear stability analyses, the relation between the hyperbolicity condition and the stability condition is analytically clarified. The concept of the 'most unstable waves' is proposed to make the wave parameters independent of the wave number, and these parameters showed well the essential features of wave propagation phenomena. Therefore, this concept could enable one to draw valuable information even from naturally generated void waves. The analytical model on the wave parameters can predict qualitatively well the experimental observations on the wave propagation properties and BSFRT criteria, but they are very sensitive to the closure laws on the momentum interactions. Especially, the virtual mass coefficient affects very much the stability condition and the wave propagation parameters, as the bubble size does in experiments
Primary Subject
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Feb 1995; 245 p; Available from Korea Advanced Institute of Science and Technology, Daejeon (KR); 74 refs, 59 figs, 8 tabs; Thesis (Dr. Eng.)
Record Type
Miscellaneous
Literature Type
Thesis/Dissertation
Country of publication
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Kim, Hwan Yeol; Song, Chul Hwa
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2003
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2003
AbstractAbstract
[en] The SCWR(Super-Critical Water cooled Reactor) is one of the six reactor candidates selected in the Gen-IV project which aims at the development of new reactors with enhanced economy and safety. The SCWR is considered to be a feasible concept of new nuclear power plant if the existing technologies developed in fossil fuel fired plant and LWR technologies together with additional research on several disciplines such as materials, water chemistry and safety. As KAERI takes part in the GIF(Generation IV Forum) for the Gen-IV project, domestic concerns about the SCWR have been recently increased. In order to establish a foundation for the development of SCWR, efforts should be concentrated on the conceptual design of systems and the associated key experiments as well. Heat transfer experiments, among others, under supercritical condition are required for the proper prediction of thermal hydraulic phenomena, which are essential for the thermal hydraulic designs of reactor core. Nevertheless, the experiments have not been performed in Korea yet. This report deals with fundamental surveys on the heat transfer experiments under supercritical conditions, which are required for the understanding of heat transfer characteristics for the thermal hydraulic designs of supercritical reactor core. Investigations on the physical properties of water and CO2 showed that the physical properties such as density, specific heat, viscosity and thermal conductivity are significantly changed near the pseudo-critical points. The state of the art on the heat transfer characteristics in relation with heat transfer deterioration and heat transfer coefficient is briefly described. In addition, previous experiments with supercritical water as well as supercritical CO2 and Freon used for an alternating fluid are presented
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Secondary Subject
Source
Jul 2003; 91 p; 75 refs, 63 figs, 5 tabs
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Report
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CONVECTION, ENERGY TRANSFER, GAS COOLED REACTORS, HALOGENATED ALIPHATIC HYDROCARBONS, HEAT TRANSFER, HEAVY WATER MODERATED REACTORS, MASS TRANSFER, ORGANIC COMPOUNDS, ORGANIC HALOGEN COMPOUNDS, PHYSICAL PROPERTIES, POWER REACTORS, PRESSURE TUBE REACTORS, REACTOR COMPONENTS, REACTORS, THERMAL REACTORS, THERMODYNAMIC PROPERTIES
Reference NumberReference Number
INIS VolumeINIS Volume
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Lee, Jae Seon; Choi, Suhn; Song, Chul Hwa
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)2008
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)2008
AbstractAbstract
[en] Recently newly-developed nuclear reactors with increased safety and enhanced performance by developed countries in the nuclear area are competing in the global nuclear market. Several reactors, for example AP600 and AP1000 by Westinghouse Electric Co. in USA, EPR by Areva in Europe, APWR by Mitsubishi Heavy Industry in Japan in the pressurized power reactor, are competing to preoccupy the nuclear market during Nuclear Renaissance. Dedicated control element drive mechanism with enhanced performance and increased safety are developed for these new reactors. And load follow capability is required, and it is estimated that load follow requirement make design requirement of a control element drive mechanism harsh. It is necessary to review the current technical state of a control element drive mechanism. This work is aimed to review the design characteristics of a past and current control element drive mechanism for a nuclear reactor and to check the direction and goal of CEDM design development recently
Primary Subject
Source
Oct 2008; 56 p; Available from KAERI; 5 refs, 18 figs, 1 tab; This record replaces 40068014
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Report
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Country of publication
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Chu, In Cheol; Choi, Nam Hyun; Song, Chul Hwa
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)2007
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)2007
AbstractAbstract
[en] A test apparatus was designed and constructed to simulate the multiple feeder pipes of CANDU reactor inlet/outlet headers, and inception criteria on the phase separation phenomena (gas entrainment and liquid entrainment) have been experimentally investigated using the test apparatus. The phase separation phenomena through a side brandch and a -36 degree inclined branch were intensively investigated, and working fluids were air and water at room temperature and near atmospheric pressure. Ultrasonic level measurement system was adopted in order to accurately measure the water level in the simulated reactor header. The present experimental results on the phase separation phenomena through the side and -36 degree inclined branches agreed relatively well with the KfK's empirical correlations for side branch. It was found that the interaction of the flow fields between the neighboring branches was not significant
Primary Subject
Source
Mar 2007; 48 p; Also available from KAERI; 6 refs, 14 figs, 5 tabs
Record Type
Report
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Song, Chul Hwa; Baek, W. P.; Yoon, B. J.
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)2010
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)2010
AbstractAbstract
[en] The improvement of prediction models is needed to enhance the safety analysis capability through the fine measurements of local phenomena. To improve the two-phase interfacial area transport model, the various experiments were carried out used SUBO and DOBO. 2x2 and 6x6 rod bundle test facilities were used for the experiment on the droplet behavior. The experiments on the droplet behavior inside a heated rod bundle were focused on the break-up of droplets induced by a spacer grid in a rod bundle geometry. The experiments used GIRLS and JICO and CFD analysis were carried out to comprehend the local condensation of steam jet, turbulent jet induced by condensation and the thermal mixing in a pool. An experimental database of the CHF (Critical Heat Flux) and PDO (Post-dryout) had been constructed. The mechanism of the heat transfer enhancement by surface modifications in nano-fluid was investigated in boiling mode and rapid quenching mode. The special measurement techniques were developed. They are Double -sensor optical void probe, Optic Rod, PIV technique and UBIM system
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Source
Apr 2010; 527 p; Also available from KAERI; 229 refs, 298 figs, 53 tabs
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Report
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Song, Chul Hwa; Park, C. K.; Lee, S. J. and others
Korea Atomic Energy Research Institute, Taejon, (Korea, Republic of)2001
Korea Atomic Energy Research Institute, Taejon, (Korea, Republic of)2001
AbstractAbstract
[en] This report presents the results of scaling analysis on the integral test loop to simulate Korean PWR plants, which includes the scaling methodology, scaling priority and scaling distortion Based on the priority of the key test matrix, the scaling priority is drawn to apply to the design of the integral test loop. Scaling analysis on the test loop is done based on the design basis and performed by the so-called ''Modified Volume Scaling'' methodology. In the first stage of scaling analysis(global scaling), the specifications and capacity of the components in the major systems are determined by the volume scaling methodology. In the second stage of scaling, local phenomena scaling is performed on major thermal hydraulic phenomena in each component and reproduction of local thermal hydraulic phenomena is checked. According to the local phenomena scaling results, specifications and capacity of the components in a system are modified. Finally, scaling distortion, which might occur in a certain system, is analyzed. Scaling results in this report might be modified in the stage of detailed design of the integral test loop, which will be carried out in near future. More comprehensive analysis on scaling distortion shall be carried out based on manufacturing specifications and performance verification test results of the integral test loop
Primary Subject
Source
Mar 2001; 280 p; 68 figs, 85 tabs
Record Type
Report
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INIS VolumeINIS Volume
INIS IssueINIS Issue
Chu, In Cheol; Song, Chul Hwa
Proceedings of the tenth international topical meeting on nuclear reactor thermal hydraulics2003
Proceedings of the tenth international topical meeting on nuclear reactor thermal hydraulics2003
AbstractAbstract
[en] In the present study a new measurement technique has been developed, which uses an ultrasonic transmission signal in order to identify the vertical two phase flow pattern. The ultrasonic measurement system developed in the present study not only provides the information required for the identification of vertical two phase flow patterns but also makes the real time identification possible. Various vertical two phase flow patterns such as bubbly, slug, churn, annular flow etc. have been accurately identified with the present ultrasonic measurement system under atmospheric condition. In addition, the present test apparatus can practically simulate the ultrasonic propagation characteristics under of high temperature and high pressure systems. Therefore, it is highly expected that the present ultrasonic flow pattern identification technique could be almost equally applicable to the vertical two phase flow systems under high temperature and high pressure conditions
Primary Subject
Source
Korea Nuclear Society, Taejon (Korea, Republic of); American Nuclear Society, La Grange Park (United States); [1 CD-ROM]; 2003; [12 p.]; NURETH-10; Seoul (Korea, Republic of); 5-11 Oct 2003; Available from the Korea Nuclear Sociey, Taejon (Korea, Republic of); 8 refs, 15 figs
Record Type
Miscellaneous
Literature Type
Conference
Country of publication
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Related RecordRelated Record
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Choi, Nam Hyun; Chu, In Chul; Min, Kyeong Ho; Song, Chul Hwa
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2005
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2005
AbstractAbstract
[en] In the case of ultrasonic measurement technique, the measurement concept was derived for the measurement of liquid film thickness and vertical/horizontal two-phase flow patterns, then ultrasonic measurement system was constructed for each application. The feasibility of the ultrasonic liquid film thickness measurement technique was confirmed throughout the test to measure the thickness and profile of the liquid film established on the 1/5 scale downcomer wall. Various two-phase flow patterns in vertical and horizontal channels were measured by photographic method and ultrasonic method, separately. Then the acceptability of the present ultrasonic measurement technique was evaluated by comparing the measurement results of each method. The ultrasonic measurement technique not only has the advantage of non-intrusive technique with easy installation and utilization but also can provide the measurement quality which is equivalent to other special measurement technique
Primary Subject
Source
Dec 2005; 35 p; Also available from Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); 2 refs, 21 figs, 2 tabs
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Report
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Euh, Dong Jin; Yun, Byeong Jo; Song, Chul Hwa; Chung, Moon Ki
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2004
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2004
AbstractAbstract
[en] The signals from a local void sensor to get the information of phasic interface in the two-phase flow have different signal levels according to each phase at the sensor tip. For an ideal case, there should be a discontinuity in void signals at the time when the phasic interface contacts the sensor and the signal should have a discrete form corresponding to each phase. Due to the characteristics of signal processing unit and the fabrication of sensor as well as those of physics near the sensor tips, however, the signals show some interval of transient curved line near the instant that the sensor contacts the interface. To get the information of various physical variables from the measured signal, signal processing for the identification of each phase should precedes. In other words, one should convert the signals to discrete form as in the ideal case. This report describes the signal processing logic and software program to convert the continuous raw signal into the discrete form illustrating definite phase. Prior to the signal processing, one should define the phasic interface definitely. In actual flow conditions, the level varies abruptly at the instant that a sensor contacts the interface. The transient is shown at the various level states of signals. This study defines the instant of the interface as the beginning point of transient. The developed program is used by inserting it into the data acquisition software. In the latter part of this report, the example illustrating this program is presented as a part of data processing software. The signal characteristics can be different by depending on the sensor, working fluid and the experimental conditions. Therefore, it is necessary to check whether the program works well under the actual flow conditions, and if required, the set parameters in the program can be tuned
Primary Subject
Source
Jan 2004; 55 p; Also available from Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); 1 ref, 16 figs
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Report
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AbstractAbstract
[en] Highlights: • OECD/KAERI international CFD benchmark exercise was operated by KAERI. • The purpose is to validate relevant CFD codes based on the MATiS-H experiments. • Blind calculation results were synthesized in terms of mean velocity and RMS. • Quality of control volume rather than the number of it was emphasized. • Major findings were followed OECD/NEA CSNI report. - Abstract: The second international CFD benchmark exercise on turbulent mixing in a rod bundle has been launched by OECD/NEA, to validate relevant CFD (Computational Fluid Dynamics) codes and develop problem-specific Best Practice Guidelines (BPG) based on the KAERI (Korea Atomic Energy Research Institute) MATiS-H experiments on the turbulent mixing in a 5 × 5 rod array having two different types of vaned spacer grids: split and swirl types. For this 2nd international benchmark exercise (IBE-2), the MATiS-H testing provided a unique set of experimental data such as axial and lateral velocity components, turbulent intensity, and vorticity information. Blind CFD calculation results were submitted by twenty-five (25) participants to KAERI, who is the host organization of the IBE-2, and then analyzed and synthesized by comparing them with the MATiS-H data. Based on the synthesis of the results from both the experiments and blind CFD calculations for the IBE-2, and also by comparing with the IBE-1 benchmark exercise on the mixing in a T-junction, useful information for simulating this kind of complicated physical problem in a rod bundle was obtained. And some additional Best Practice Guidelines (BPG) are newly proposed. A summary of the synthesis results obtained in the IBE-2 is presented in this paper
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Source
S0029-5493(14)00140-X; Available from http://dx.doi.org/10.1016/j.nucengdes.2014.03.008; Copyright (c) 2014 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.; Country of input: International Atomic Energy Agency (IAEA)
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Journal Article
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