• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.534-545
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• 2015

In the very high temperature reactor (VHTR), which is a next generation nuclear reactor system, ceramics are used as a fuel coating material and graphite is used as a core structural material. Even if a depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change only slowly. This is because the thermal capacity of the core is so high. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel. The objectives of this study are to investigate the heat transfer characteristics of natural convection of a one-side heated vertical channel with inserted porous materials of high porosity and also to develop the passive cooling system for the VHTR. An experiment was carried out using a one-side heated vertical rectangular channel. To obtain the heat transfer and fluid flow characteristics of the vertical channel with inserted porous material, we have also carried out a numerical analysis using a commercial Computational Fluid Dynamics (CFD) code. This paper describes the thermal performances of the one-side heated vertical rectangular channel with an inserted copper wire of high porosity.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2375-2387
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• 2024

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.3
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• pp.191-198
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• 2014

In this paper, structural integrity evaluation of reactor pressure vessel bottom head without penetration nozzles in core melting accident has been performed. Considering the analysis results of thermal load, weight of molten core debris and internal pressure, thermal load is the most significant factor in reactor vessel bottom head. The failure probability was evaluated according to the established failure criteria and the evaluation showed that the equivalent plastic strain results are lower than critical strain failure criteria. Thermal-structural coupled analyses show that the existence of elastic zone with a lower stress than yield strength is in the middle of bottom head thickness. As a result of analysis, the elastic zone became narrow and moved to the internal wall as the internal pressure increases, and it is evaluated that the structural integrity of reactor vessel is maintained under core melting accident.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.21 no.9
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• pp.610-616
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• 2020

This study conducted joint research with the Navy logistics command ship technology research institute to resolve the occurrence of naval vessel's high-temperature warning and equipment shutdown caused by VRTU overheating during summer operation and the dispatch of troops to equatorial regions. The cooling effect was checked according to the installation of a thermoelectric device cooling system, and heat flow and heat transfer characteristics inside VRTU was analyzed using Computational Fluid Dynamics. In addition, the temperature distribution inside the engine room was assessed through interpretation, and the optimal installation location to prevent VRTU overheating was identified. As a result, the average volume temperature inside the VRTU decreased by approximately 10 ℃ with the installation of the cooling system, and the fan installed in the cooling system made the heat circulation smooth, enhancing the cooling effect. The inside of the engine room showed a high-temperature distribution at the top of the engine room, and the end of the HVAC duct diffuser showed the lowest temperature distribution.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.995-1004
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• 2009

The Advanced Power Reactor 1400 (APR1400) is an evolutionary advanced light water reactor (ALWR) based on the Optimized Power Reactor 1000 (OPR1000), which is in operation in Korea. The APR1400 incorporates a variety of engineering improvements and operational experience to enhance safety, economics, and reliability. The advanced design features and improvements of the APR1400 design include a pilot operated safety relief valve (POSRV), a four-train safety injection system with direct vessel injection (DVI), a fluidic device (FD) in the safety injection tank, an in-containment refueling water storage tank (IRWST), an external reactor vessel cooling system, and an integrated head assembly (IHA). Development of the APR1400 started in 1992 and continued for ten years. The APR1400 design received design certification from the Korean nuclear regulatory body in May of2002. Currently, two construction projects for the APR1400 are in progress in Korea.

• 한국연소학회:학술대회논문집
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• 2002.06a
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• pp.95-98
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• 2002

A numerical investigation on the flame propagation and extinction in a micro combustor is described. Previous measurements of $H_2-air$ flame propagation in a submilimeter scale combustor exhibited significance of wall effects on burning velocity and extinction. The heat transfer to wall becomes important not only in the cooling of burnt gases but also during the flame ropagation, which has be by and large ignored in macro scale combustor calculations. In order to take the heat loss into account the combustion calculation, we developed a numerical code with a heat transfer model that was determined empirically from measured data. PISO algorithm was used for differencing of conservation equations. $H_2-air$ reaction was modeled with 10 species - 16 steps. Comparison with measured data showed good agreement in flame propagation speed. Also the pressure decrease after flame extinction was accurately predicted by the model. A further study is desirable for a better quenching model that can predict the quenching location.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.10
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• pp.1043-1047
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• 2012

The European nuclear design requirements, which should be satisfied by nuclear reactors in Europe, usually recommend a so-called core catcher, which is a molten core ex-vessel cooling facility, to manage a severe accident at a nuclear reactor. Two different types of core catcher concepts are compared to determine their abilities to manage severe accidents and cool core melts. The study reveals that direct cooling is better for cooling capacity and is convenient to construct, while indirect cooing is better for the management of a severe accident.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.99-108
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• 2006

SNU-RCCS is a water pool type RCCS (Reactor Cavity Cooling System) developed for VHTR (Very High Temperature Reactor) application by SNU (Seoul National University). Since radiation heat transfer is the major process of passive heat removal in a RCCS, it is important to determine the precise emissivity of the reactor vessel. Review studies have used a constant emissivity in the passive heat removal analysis, even though the emissivity depends on many factors such as temperature, surface roughness, oxidation level, wavelength, direction, atmosphere conditions, etc. Therefore, information on the emissivity of a given material in a real RCCS is essential in order to properly analyze the radiation heat transfer in a VHTR. The objectives of this study are to develop a method for compensation of the factors affecting the emissivity measurement using an infrared thermometer and to estimate the true emissivity from the measured emissivity via the developed method, especially in the SNU-RCCS environment. From this viewpoint, we investigated factors such as the attenuation effect of the window, filling gas, and the effect of background radiation on the emissivity measurements. The emissivity of the vessel surface of the SNU-RCCS facility was then measured using a sight tube. The background radiation was subsequently removed from the measured emissivity by solving a simultaneous equation. Finally, the calculated emissivity was compared with the measured emissivity in a separate emissivity measurement device, yielding good agreement with the emissivity increase with vessel temperature in a range of 0.82 to 0.88.

• Journal of Power System Engineering
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• v.20 no.2
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• pp.91-96
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• 2016

A fleet consists of a main vessel, light vessels and carrying vessels for purse seine fishery. Carrying vessels contains fish storages to maintain freshness of catches. Currently most carrying vessels applies the cooling system using plain ice though accompanied various shortcomings. Seawater cooling system directly chilling seawater are now in use on carrying vessels in some developed countries to make up for these shortcomings and maximize advantages. This research deals with necessity of seawater cooling systems and establishes system criteria using Aspentech HYSYS program, prior to an experiment of compact-scale seawater cooling system which now in progress of manufacture. Performance comparison on condensation capacity, mass flow rate of working fluid, compressor power input, pump power input and others of the seawater cooling system applying a flooded evaporator is conducted with respect to the temperature of surface seawater varying according to seasons. The result presents that mass flow rate circulating the system is increased about 16.7% as the temperature of surface seawater increases. At the same condition, condensation capacity and compressor input work also increase about 9.8% and 91.2%, respectively.

• Proceedings of the KWS Conference
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• 2002.10a
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• pp.418-424
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• 2002

Korean Fusion Reactor(KSTAR) system consists of a vacuum vessel, in-vessel components, cryostat, thermal shield, super-conducting magnets and magnet supporting structures. These systems are in the final stage of engineering design with the involvement of industrial manufacturers. The overall configuration and the detailed dimensions of the KSTAR structure have been determined and the first stage of manufacturing is progressing now. In this study, the fabrication and assembly sequence were evaluated in viewpoint of high strengthening joints and very high accuracy. Especially for this purpose, the special cleaning process and welding process were proposed for high strengthening austenitic stainless steel which shall be used at cryogenic temperature. The draft procedure qualification data for welding process are presented with precise welding data including special narrow groove design. For the cooling line attachment on the surface of inside wall of magnet structure case, Induction brazing technology is introduced with some special jigging system and some consumables.