• Transactions of the Korean Society of Automotive Engineers
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• v.16 no.5
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• pp.179-189
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• 2008

The performance of Lean NOx Trap (LNT) based on the catalysts of Pt/K/Ba/$\gamma-Al_2O_3$ with proprietary washcoat formulation is studied using a bench flow reactor system. To investigate the effect of temperature and gas hourly space velocity (GHSV) on the nitrogen oxides (NOx) trapping capacity as well as NOx breakthrough time and final ratio of $NO_2$ to NO of LNT, series of adsorption isotherms are carried out with simulated exhaust gases of the lean burn engines. Since typical operation of LNT requires periodic regeneration with a short rich excursion, where the stored or trapped NOx is released and subsequently reduced to $N_2$, the effect of the duration of lean and rich phase and type of reductants on the NOx conversion is investigated. NOx storage capacity and breakthrough time obtained from adsorption isotherms shows a volcano-type dependence on the temperature with a maximum NOx storage capacity occurring $350^{\circ}C$ and with a maximum breakthrough time occurring $400^{\circ}C$ at all GHSVs investigated in this study. Also, maximum ratio of $NO_2$ to NO is obtained at $400^{\circ}C$ with a GHSV of $75,000\;hr^{-1}$ Lean/rich cycle of 100 s lean and 5 s rich used with a concentration of 1.33% of $H_2$ and 4% of CO in the rich phase is found to be optimum at operating temperature of $350^{\circ}C$ and a GHSV of $50,000\;hr^{-1}$.

• Nuclear Engineering and Technology
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• v.39 no.1
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• pp.9-20
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• 2007

Design study on the Gas Turbine High Temperature Reactor 300-Cogeneration (GTHTR300C) aiming at producing both electricity by a gas turbine and hydrogen by a thermochemical water splitting method (IS process method) has been conducted. It is expected to be one of the most attractive systems to provide hydrogen for fuel cell vehicles after 2030. The GTHTR300C employs a block type Very High Temperature Reactor (VHTR) with thermal power of 600MW and outlet coolant temperature of $950^{\circ}C$. The intermediate heat exchanger (IHX) and the gas turbine are arranged in series in the primary circuit. The IHX transfers the heat of 170MW to the secondary system used for hydrogen production. The balance of the reactor thermal power is used for electricity generation. The GTHTR300C is designed based on the existing technologies of the High Temperature Engineering Test Reactor (HTTR) and helium turbine power conversion and on the technologies whose development have been well under way for IS hydrogen production process so as to minimize cost and risk of deployment. This paper describes the original design features focusing on the plant layout and plant cycle of the GTHTR300C together with present development status of the GTHTR300, IHX, etc. Also, the advantage of the GTHTR300C is presented.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.323-334
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• 2013

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.

• Journal of the Korean Institute of Gas
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• v.2 no.3
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• pp.60-69
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• 1998

Methane, the major constituent of natural gas, had been converted to higher hydrocarbons by microwave and radio-frequency plasma in vacuum condition. Methane had been activated to plasma by suppling high energy then converted to ethane, ethylene, acetylene. The direct conversion process of methane had produced few by-products and demanded low-energy. The plasma sources were microwave and radio-frequency. Two types of reactor had been used to activate methane. One is common single tubular-type reactor and the other is series coil-type reactor which used for the first time in this study. To produce more C2 products, methane had been converted by a plasma and catalyst. The results of this study could be used to study mechanism of plasma reaction of methane, design the plant-scale reactor.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.58 no.2
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• pp.232-238
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• 2009

Flicker, also known as voltage fluctuation, is a newest problem of power quality issues, because it is caused by nonlinear loads such as electrical arc furnace and large-scale induction motor, which are country-widely used as the heavy industries of a country develop. An international standard, International Electrotechnical Commission (IEC) 61000-4-15, was published in 1997 and revised in 2003. With increasing concerns about flicker, its mitigation methods have been also studied. General countermeasures for flicker are divided into three categories: a) enhancing the capacity of supplying system, b) Series elements including series reactor and series capacitor and c) power electronic devices including static VAR compensator (SVC) and static synchronous compensator (STATCOM). This paper introduces how to mitigate the voltage flicker at the point of common coupling (PCC) and presents how to simulate and compare the flicker alleviating effects by each mitigation method, using IEC flickermeter based on the Matlab/Simulink program.

• Korean Journal of Metals and Materials
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• v.48 no.5
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• pp.410-416
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• 2010

Studies were carried out to establish the technology for sodium-clad compatibility and to analyze the compatibility behavior of the Sodium-cooled Fast Reactor (SFR) cladding material under a flowing sodium environment. The natural circulation facility caused by the thermal convection of the liquid sodium was constructed and the 316-series stainless steels were exposed at $650{^{\circ}C}$ liquid sodium for 1458 hours. The weight change and related microstructural change were analyzed. The results showed that the quasi-dynamic facility represented by the natural convection exhibited similar results compared to the conventional dynamic facility. Selective leaching and local depletion of the chromium, re-distribution of the carbide, and the decarburization process took place in the 316-series stainless steel under a flowing sodium environment. This process decreased as the sodium flowed along the channel, which was caused by the change in the dissolved oxygen and carbon activity in the liquid sodium.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.4
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• pp.457-465
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• 2010

In this study, a Taylor series (T-S) model based on the Arrhenius, McVetty, and Monkman-Grant equations was developed using a mathematical analysis. In order to reduce fitting errors, the McVetty equation was transformed by considering the first three terms of the Taylor series equation. The model parameters were accurately determined by a statistical technique of maximum likelihood estimation, and this model was applied to the creep data of alloy 617. The T-S model results showed better agreement with the experimental data than other models such as the Eno, exponential, and L-M models. In particular, the T-S model was converted into an isothermal Taylor series (IT-S) model that can predict the creep strength at a given temperature. It was identified that the estimations obtained using the converted ITS model was better than that obtained using the T-S model for predicting the long-term creep life of alloy 617.

• Korean Chemical Engineering Research
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• v.49 no.4
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• pp.491-497
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• 2011

In this work, we proposed the three different reactor systems for evaluating of synthetic natural gas(SNG) processes using the synthesis gas consisting of CO and $H_2$ and reactor systems to be considered are series adiabatic reaction system, series adiabatic reaction system with the recirculation and cooling wall type reaction system. The maximum temperature of the first adiabatic reactor in series adiabatic reaction system raised to 800. From the these results, carbon dioxide in product gas as compared to other systems was increased more than that expected due to water gas shift reaction(WGSR) and the maximum $CH_4$ concentration in SNG was 90.1%. In series adiabatic reaction system with the recirculation as a way to decrease the temperature in catalyst bed, the maximum $CH_4$ concentration in SNG was 96.3%. In cooling wall type reaction system, the reaction heat is absorbed by boiling water in the shell and the reaction temperature is controlled by controlling the amount of flow rate and pressure of feed water. The maximum $CH_4$ concentration in SNG for cooling wall type reaction system was 97.9%. The main advantage of the cooling wall type reaction system over adiabatic systems is that potentially it can be achieve almost complete methanation in one reactor.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.678-699
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• 2015

Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for nearterm human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of $100MW_{th}$ and an electricity generation mode of $100MW_{th}$, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and thermohydraulics was carried out. The result indicates that the innovative design has great potential for high propellant efficiency and thrust-to-weight of engine ratio, compared with the existing NTR designs. However, the build-up of fission products in fuel has a significant impact on the bimodal operation of the moderated reactor such as xenon-induced dead time. This issue can be overcome by building in excess reactivity and control margin for the reactor design.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.127-132
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• 2001

An experimental study on penetration integrity of the reactor vessel has been performed under external vessel cooling during a core melt accident. In this study a series of experiments are performed for the verification of the effects of coolant in the annulus between the ICI(In-Core Instrumentation) nozzle and the thimble tube and also the effects of external vessel cooling on the integrity of the penetration using the test section including only one penetration and $Al_{2}O_{3}$ melt as a corium simulant. The experimental results have shown that penetration is more damaged in the case of no external vessel cooling compared with the case of external vessel cooling. It is preliminarily concluded that the external vessel cooling is very effective measure for the improvement of the penetration integrity. Also it is confirmed from the experimental results that the coolant in the annulus reduces the melt penetration distance through the annulus and enhance the integrity of the reactor vessel penetration in the end.