• Progress in Superconductivity and Cryogenics
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• v.10 no.4
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• pp.33-37
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• 2008

The main objective of this study is to identify the pressure drop characteristics of coolant flow passages of 154kV/1GVA High Temperature Superconducting (HTS) power cable, experimentally. The passages were consisted of two parts, the one is the circular path with spiral ribs in the core to cool the cable conductor layer and the other is annular path with spirally corrugated outer wall to cool the shield layer. Thus the experiments to acquire the pressure drop data were performed with two types of circular spiral tubes and eight types of the concentric annuli in various range of Reynolds number. The pressure drops in the core tubes and the annuli were much higher than those in the tubes with smooth surface. Therefore, modified correlations to present the experimental results in each flow passage were suggested.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2458-2473
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• 2024

The turbulent flow in reactor pressure vessel (RPV) of pressurized water reactor (PWR) is important for the flow rate distribution at core inlet. Thus, it is vital to study the turbulent flow phenomena in RPV. However, the complicated fluid channel consisted of inner structures of RPV will block or refract the laser sheet of particle image velocimetry (PIV). In this work, the matched index of refraction (MIR) of sodium iodide (NaI) solution and acrylic was applied to support optical path for flow field measurements by PIV in the 1/10th scaled-down RPV model. The experimental results show detailed velocity field at different locations inside the scaled-down RPV model. Some interesting phenomena are obtained, including the non-negligible counterflow at the corner of nozzle edge, the high downward flowing stream in downcomer, large vortices above vortex suppression plate in lower plenum. And the intensity of counterflow and the strength of vortices increase as inlet flow rate increasing. Finally, the case of asymmetry flow was also studied. The turbulent flow has different pattern compared with the case of symmetrical inlet flow rate, which may affect the uniformity of flow distribution at the core inlet.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.4
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• pp.425-429
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• 2011

In the unlikely of nuclear reactor meltdown, the leaked core melt or corium must be contained in a device called core-catcher so that the corium can be cooled and stabilized. The ex-vessel behavior of corium involves complex physical and chemical mechanisms of flow propagation, heat transfer, and reactions with sacrificial substrates. In this study, the detailed characteristics of corium flow and heat transfer were investigated by using a commercial CFD code for VULCANO VE-U7 test reported in the literature. The volume-of-fluid (VOF) model was used to predict the interfacial surface formation of corium and the surrounding air, and the discrete ordinate model was adopted to calculate radiation between corium and the surroundings. It was found that cooling via radiation through the top surface of corium had a dominant effect on the temperature and viscosity profiles at the front of the corium flow.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.598-603
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• 1996

When locked rotor event. occurs, instantaneously affected loop and core flow were quickly reduced, which resulted in an increase in coolant temperature and system pressure. Analysis method of this event was that constant core inlet temperature and system pressure as well as change in core flow calculated from COAST code were statically used as an input variable to HERMITE code, because of no tools to simulate NSSS behavior and 1-D core neutronics transient coincidently. With employing TASS code revised with 1-D neutronics model, this event was analyzed in point of DNBR. By doing so, analysis procedure could be simplified and unreasonable conservatism might be removed in DNBR calculation by consideration of pressure increase.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1417-1428
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• 2020

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MW_th. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MW_th.

• IEMEK Journal of Embedded Systems and Applications
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• v.17 no.6
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• pp.375-382
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• 2022

This paper proposes a multi-core cooperative computing structure considering the heterogeneous features of automotive embedded on-chip software. The automotive embedded software has the heterogeneous execution flow properties for various hardware drives. Software developed with a homogeneous execution flow without considering these properties will incur inefficient overhead due to core latency and load. The proposed method was evaluated on an target board on which a automotive MCU (micro-controller unit) with built-in multi-cores was mounted. We demonstrate an overhead reduction when software including common embedded system tasks, such as ADC sampling, DSP operations, and communication interfaces, are implemented in a heterogeneous execution flow. When we used the proposed method, embedded software was able to take advantage of idle states that occur between heterogeneous tasks to make efficient use of the resources on the board. As a result of the experiments, the power consumption of the board decreased by 42.11% compared to the baseline. Furthermore, the time required to process the same amount of sampling data was reduced by 27.09%. Experimental results validate the efficiency of the proposed multi-core cooperative heterogeneous embedded software execution technique.

• The Magazine of the Society of Air-Conditioning and Refrigerating Engineers of Korea
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• v.14 no.1
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• pp.21-30
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• 1985

One of the problems with the Boiling Water Reactor involves the flow and thermal mixings in the suppression water pool high pressure steam discharge into the pool in case of emergency core relief. Varioos heat sensitive devices and pumps for the reactor core cooling are installed in the middle of the suppression pool. Especially the pumps utilize pool water in order to cool the reactor core in emergency cases. In this case, the water temperature for the reactor cool ins should be below a certain temperature specified by the reactor design. In the present investigation, in other to determine the optimum locations of these pumping devices, numerical solutions have been obtained for the model to determine the f low mixing characteristics. Experimental investigations have also been carried out for the flow mixing and for the thermal mixing in the pool during the discharge. Considering that the discharge steam through the Quenching Device becomes hot water immediately in the water pool, the steam- equivalent hot water has been utilized. Examining these characteristices, it becomes possible to deform me the best locations for RCIC, LPCI , HPCI pumps in the suppression water pool for the emermency reactor core cooling.

• The Bulletin of The Korean Astronomical Society
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• v.45 no.1
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• pp.30.2-31
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• 2020

We present the properties of circumnuclear gas associated with the AGN located in the center of Abell 1644-South. A1644-S is the main cluster in a merging system, which is also known for gas sloshing in its core as seen in X-ray. The X-ray emission of A1644-S shows a rapidly declining profile, indicating the presence of cooling gas flow. This flow of cool gas may fuel the supermassive black hole embedded in the brightest cluster galaxy, leading to the activation of the central AGN. Indeed, we find a parsec-scale bipolar jet feature in the center of A1644-S in our recent KaVA observation, which implies that its central AGN is likely to have been (re)powered quite recently. In order to verify the hypothesis that cooling gas flow in the cluster core can (re)activate the central AGN, we probe the cold gas properties of the central 1 kpc region of A1644-S using the archival VLA and ALMA data. Based on the spatially resolved morphology and kinematics of HI and CO gas, we challenge to identify inflow/outflow gas streams and clumps. We study the role of circumnuclear cool gas in fueling the centrally located cluster AGN in the cool-core environment. We also discuss how the feedback due to the (re)powered AGN affects the surrounding medium.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.949-958
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• 2024

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.502-516
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• 2002

Wolsong NPP are CANDU6 type reactors and there are 4 CANDU6 type reactors in commercial operation. CANDU type reactors require on-power refuelling by two remote controlled F/Ms (Fuelling Machines). Most of channels, fuel bundles are float by channel coolant flow and move toward downstream, however in about 30% of channels the coolant flow are not sufficient enough to carry fuel bundles to downstream. For those channels a special device, FARE (Flow Assist Ram Extension) device, is used to create additional force to push fuel bundles. It has been showing that during fuelling operation of some channels the channel coolant flow rate is reduced below specified limit (80% of normal), and consequently trip alarm signal turns on. This phenomenon occurs on selected channels that are instrumented for the channel flow and required to use the FARE device for refuelling. Hence it is believed that the FARE device causes the problem. It is also suspected that other channels that do not use the FARE device for refuelling might also go into channel flow low state. The analysis revealed that the channel How low occurs as the FARE device is introduced into the core and disappears as the FARE device is removed from the core. This paper presented the FARE device behavior, detailed fuelling operation sequence with the FARE device and effect on channel flow low phenomena. The FARE device components design changes are also suggested, such as increasing the number or now holes in the tube and flow slots in the ring orifice.