• 한국초전도ㆍ저온공학회논문지
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• 제2권1호
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• pp.7-13
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• 2000

A rotating cryogenic system was designed similar to the cooling system for the rotor of a superconducting generator. The experimental rotor has an inner vessel which simulates the winding space of an actual superconducting rotor, and a torque tube of comparable design. This paper describes the evaluation of the total heat leak into the inner vessel that leads to the study of the heat transfer characteristic of the rotating cryogenic system. To examine the insulation performance of the experimental rotor. temperature was measured at each part of the system at various rotaing speeds from 0 rpm to 600 rpm. Total heat leak into the inner vessel was calculated by measuring the boil-off rate of liquid helium. Conduction heat leak to the inner vessel was obtained by the vent tube, and radiation heat leak was calculated by subtracting the conduction heat lent from the total heat leak. There seemed to be no rotaional dependency of total heat leak at least up 600 rpm.

• 대한조선학회논문집
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• 제41권6호
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• pp.134-139
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• 2004

It is known that over 60% of engine power is dissipated into circumstance, cooling water and cooling oil without any conversion into useful work. Following the first law of thermodynamics, it is possible that heat loss to cooling water can be converted into mechanical work through crankshaft. But in case that the engine is operating without any cooling effect, the serious problem unsolved so far is the engine durability. In this study, LHR(Low Heat Rejection) engine was introduced as one of the promising engine and several useful qualitative and quantitative data were drawn.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.144-155
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• 2023

The core catcher is used as a passive safety system in new generation nuclear power plants to create a space in the containment for the placing and cooling of the molten corium under various severe accidents. This research investigates the role of the core catcher in the VVER-1000 reactor containment system in mitigating the effects of core meltdown under various severe accidents within the context of the Ex-vessel Melt Retention (EVMR) strategy. Hence, a comparison study of three severe accidents is conducted, including Station Black-Out (SBO), SBO combined with the Large Break Loss of Coolant Accident (LB-LOCA), and SBO combined with the Small Break Loss of Coolant Accident (SB-LOCA). Numerical comparative simulations are performed for the aforementioned scenario with and without the EX-vessel core-catcher. The results showed that considering the EX-Vessel core catcher reduces the amount of hydrogen by about 18.2 percent in the case of SBO + LB-LOCA, and hydrogen production decreases by 12.4 percent in the case of SBO + SB-LOCA. Furthermore, in the presence of an EX-Vessel core-catcher, the production of gases such as CO and CO2 for the SBO accident is negligible. It was revealed that the greatest decrease in pressure and temperature of the containment is related to the SBO accident.

• 에너지공학
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• 제22권1호
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• pp.8-16
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• 2013

A core catcher is considered as a promising engineered system to stabilize the molten corium in the containment during a postulated severe accident in a nuclear power plant. Conceptually, the core catcher consists of a carbon steel body, sacrificial material, protection material, and engineered cooling channel. The cooling capacity of the engineered cooling channel should be guaranteed to remove the decay heat of the molten corium. The flow in ex-vessel core catcher is a combined problem of a two-phase flow in the engineered cooling channel and a single-phase natural circulation in the whole core catcher system. In this study, the analysis of the test facility for the core catcher using the CUPID code, which is a three-dimensional thermal-hydraulic code for the simulation of two-phase flows, was carried out to evaluate its cooling capacity.

• 설비공학논문집
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• 제19권12호
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• pp.850-856
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• 2007

In a supercooled or capsule type ice storage system, aqueous solution (or water) may have trouble with non-uniform dissolution though the system contributes to the simplicity of system and ecological improvement. The non-uniform dissolution increases the instability of the system because it may cause an ice blockage in pipe or cooling part. In order to observe the supercooled state, a cooling experiment was performed with pressurization to an ethylene glycol(EG) 3 mass% solution in stationary state. Also, the effect of the pressurization from 101 to 505 kPa to the dissolution of supercooled aqueous solution was measured with the dissolution time of the supercooled aqueous solution at a fixed cooling rate of brine. At results, the dissolution of supercooled point decreased as the pressure of the aqueous solution in the vessel increased. Moreover, the dissolution point increased as the heat flux for cooling increased.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.9-18
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• 2024

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a severe accident management (SAM) strategy that has been adopted and used in many nuclear reactors such as AP1000, APR1400, and light water reactor etc. Some reactor accidents have raised concerns about nuclear reactors among residents, leading to a decrease in residents' acceptability and many studies on SAM are being conducted. Experiments on IVR-ERVC are almost impossible due to its specificity, so fluid characteristics are analyzed through BALI experiments with similar condition. In this study, computational fluid dynamics (CFD) via Reynolds-averaged Navier-Stokes (RANS) and large eddy simulation (LES) for BALI experiments were performed. Steady-state CFD analysis was performed on three turbulence models, and SST k-ω model was in good agreement with the experimental measurement temperature within the maximum error range of 1.9%. LES CFD analysis was performed based on the RANS analysis results and it was confirmed that the temperature and wall heat flux for depth was consistent within an error range of 1.0% with BALI experiment. The LES CFD analysis results were compared with those of the Lagrangian-based solver. LES matched the temperature distribution better than SOPHIA, but SOPHIA calculated the position of boundary between stratified layer and convective layer more accurately. On the other hand, Lagrangian-based solver predicted several small eddy behaviors of the convective layer and LES predicted large vortex behavior. The vibration characteristics near the cooling part of the BALI experimental device were confirmed through Fast Fourier Transform (FFT) investigation. It was found that the power spectral density for pressure at least 10 times higher near the side cooling than near the top cooling.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.383-387
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• 2004

The Fuel Test Loop (FTL) consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). The test condition in IPS such as pressure, temperature and quality of the main cooling water, can be controlled by the OPS. The FTL has been developed to be able to irradiate three pins to the core irradiation hole (IR1 hole) by considering for its utility and user's irradiation requirement. The IPS vessel assembly (IVA) consists of IPS head, outer pressure vessel, inner pressure vessel, inner assembly and test fuel carrier. The IVA is approximately 5.6 m long and fits within a 74 mm in diameter envelope over the full height of the chimney. Above the top of the chimney, the head of the IPS is enlarged to allow the closure flanges and pipe work connections. IVA was designed to test the CANDU and PWR nuclear fuel pin together. Specially, wished to minimize interference by nuclear fuel change in design and synthesize these items and shape design for IVA.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집B
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• pp.247-254
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• 2000

The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, $10^{-6}{\sim}10^{-7}Pa$, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least $250^{\circ}C,\;350^{\circ}C$ respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• Journal of Advanced Marine Engineering and Technology
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• 제33권4호
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• pp.443-450
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• 2009

This study has dealt with the ice making characteristics in a rectangular vessel cooled from below direction with experiment and numerical analysis. The experiment and numerical analysis were carried out under the following conditions which are the cooling wall temperatures of -5[$^{\circ}C$], -10[$^{\circ}C$], and -15[$^{\circ}C$]. The temperature profile of non-frozen layer was calculated by numerical analysis as the form of non-dimensional temperature. From this study, it is cleared that the existence of natural convection is clearly known. And also the non-dimensional freezing amount was derived from experimental result. This correlation equation will give a useful information to the designers of ice making system.