• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.39 no.4
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• pp.348-353
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• 2011

Characteristics of cryogenic single jet flow were investigated. Liquid nitrogen was injected into a high-pressure chamber and formed single jet. Ambient condition around jet was changed from subcritical to superctirical condition of nitrogen. Injector geometries also were changed. A shape of the jet and core diameter were measured by flow visualization technique, and core spreading angle was calculated. Flow instability was found at atmospheric pressure condition. As ambient pressure increased, core spreading angle was increased and maintained after certain pressure.

• Journal of Biomedical Engineering Research
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• v.37 no.5
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• pp.186-194
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• 2016

In this study, a co-axial flow induced microfluidic chip to fabricate pure collagen type I microfiber via the control of collagen type I and Na-alginate gelation process. The pure collagen type I microfiber was generated by selective degradation of Ca-alginate from 'Core-Shell' structured hydrogel microfiber. To make 'Core-Shell' structure, collagen type I solution was introduced into core channel and 1.5% Na-alginate solution was injected into side channel in microfluidic chip. To evaluatethe 'Core-Shell' structure, the red and green fluorescence substances were mixed into collagen type I and Na-alginate solution, respectively. The fluorescence substances were uniformly loaded into each fiber, and the different fluorescence images were dependent on their location. By immoblizing EpH4-Ras and C6 cells within collagen type I and Na-alginate solution, we sucessfully demonstrated the co-culture of EpH4-Ras and C6 cells with 'Core-Shell' like hydrogel microfiber for 5 days. Only to produce pure collagen type I hydrogel fiber, tri-sodium citrate solution was used to dissolve the shell-like Ca-alginate hydrogel fiber from 'Core-Shell' structured hydrogel microfiber, which is an excellent advantage when the fiber is employed in three-dimensional scaffold. This novel method could apply various application in tissue engineering and biomedical engineering.

• Transactions of the Korean Society of Mechanical Engineers
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• v.8 no.3
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• pp.274-287
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• 1984

밀폐용기 내의 자연대류에 의한 흐름현상에 고유하며 아직까지 해결되지 않고 있는 중요한 문제 는 그 내부흐름의 형태를 주어진 물리적인 조건으로부터 미리 예측할 수 없다는 것이다. 이는 이 문제의 해석적인 해결에 가장 큰 난점이 되고 있다. 본 논문에서는 multiple scales method를 이 용한 해석적인 모델을 개발하여 종횡비가 낮은 직각형 밀폐용기 내에서 운동력이 코어에 존재하 는 경우 그 내부 흐름형태를 scaling analysis를 통해 정성적으로 예측, 기존의 결과와 비교 검토 하였다.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1532-1539
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• 2019

This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

• IEIE Transactions on Smart Processing and Computing
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• v.4 no.2
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• pp.83-88
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• 2015

Core-A is 32-bit synthesizable processor core with a unique instruction set architecture (ISA). In this paper, the Core-A ISA is introduced with discussion of useful features and the development environment, including the software tool chain and hardware on-chip debugger. Core-A is described using Verilog-HDL and can be customized for a given application and synthesized for an application-specific integrated circuit or field-programmable gate array target. Also, the GNU Compiler Collection has been ported to support Core-A, and various predesigned platforms are well equipped with the established design flow to speed up the hardware/software co-design for a Core-A-based system.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2204-2220
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• 2020

An open-pool type research reactor is designed and operated considering the accessibility around the pool top area to enhance the reactor utilization. The reactor structure assembly is placed at the bottom of the pool and filled with water as a primary coolant for the core cooling and radiation shielding. Most radioactive materials are generated from the fuel assemblies in the reactor core and circulated with the primary coolant. If the primary coolant goes up to the pool surface, the radiation level increases around the working area near the top of the pool. Hence, the hot water layer is designed and formed at the upper part of the pool to suppress the rising of the primary coolant to the pool surface. The temperature gradient is established from the hot water layer to the primary coolant. As this temperature gradient suppresses the circulation of the primary coolant at the upper region of the pool, the radioactive primary coolant rising up directly to the pool surface is minimized. Water mixing between these layers is reduced because the hot water layer is formed above the primary coolant with a higher temperature. The radiation level above the pool surface area is maintained as low as reasonably achievable since the radioactive materials in the primary coolant are trapped under the hot water layer. The key to maintaining the stable hot water layer and keeping the radiation level low on the pool surface is to have a stable flow of the primary coolant. In the research reactor with a downward core flow, the primary coolant is dumped into the reactor pool and goes to the reactor core through the flow guide structure. Flow fields of the primary coolant at the lower region of the reactor pool are largely affected by the dumped primary coolant. Simple, circular, and duct type discharge headers are designed to control the flow fields and make the primary coolant flow stable in the reactor pool. In this research, flow fields of the primary coolant and hot water layer are numerically simulated in the reactor pool. The heat transfer rate, temperature, and velocity fields are taken into consideration to determine the formation of the stable hot water layer and primary coolant flow. The bulk Richardson number is used to evaluate the stability of the flow field. A duct type discharge header is finally chosen to dump the primary coolant into the reactor pool. The bulk Richardson number should be higher than 2.7 and the temperature of the hot water layer should be 1 ℃ higher than the temperature of the primary coolant to maintain the stability of the stratified thermal layer.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.1-10
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• 2006

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability.

• Journal of the Korean Electrochemical Society
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• v.15 no.2
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• pp.74-82
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• 2012

Proton exchange membrane fuel cell performance is changed by the complicated physical phenomenon. In this study, water discharge performance of proton exchange membrane fuel cell were performed numerically to compare serpentine channel flow fields of 5-pass 4-turn serpentine and 25 $cm^2$ reaction surface between with and without sub-channel at the rib. Through the supplement of sub channel flow field, it is shown from the results that water removal characteristic inside channel improves because the flow direction of under-rib convection is changed into the sub channel. Reacting gases supplied from entrance disperse into sub channel flow field and electrochemical reaction occurs uniformly over the reaction surface. The results obtained that total current density distributions become uniform because residence time of reacting gases traveling to sub-channel flow field is longer than to main channel.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.721-730
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• 2013

Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth). Two types of cores are being studied for the ASTRID project. The first is based on a 'large pin/small spacing wire' concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth. In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents. An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design. This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning: - Core performance, - Intrinsic behavior during unprotected transients, - Simulation of severe accident scenarios, - Qualification requirements. The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase.

• International Journal of Fluid Machinery and Systems
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• v.5 no.2
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• pp.49-59
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• 2012

The present study is devoted to the influence of a superposed radial inflow in a rotor-stator cavity with a peripheral opening. The flow regime is turbulent, the two boundary layers being separated by a core region. An original theoretical solution is obtained for the core region, explaining the reason why a weak radial inflow has no major influence near the periphery of the cavity but strongly affects the flow behavior near the axis. The validity of the theory is tested with the help of a new set of experimental data including the radial and tangential mean velocity components, as well as three components of the Reynolds stress tensor measured by hot-wire anemometry. The theoretical results are also in good agreement with numerical results obtained with the Fluent code and experimental data from the literature.