• Proceedings of the Korean Vacuum Society Conference
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• 2012.08a
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• pp.431-431
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• 2012

A compact tokamak reactor concept as a 14 MeV neutron source is desirable from an economic viewpoint for a fusion-driven transmutation reactor. LAR (Low Aspect Ratio) tokamak allows a potential of high "see full txt" operation with high bootstrap current fractions and can be used for a compact fusion neutron source. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components and are constrained to use ITER physics and technology. In a transmutation reactor, the blanket should produce enough tritium for tritium self-sufficiency and the neutron multiplication factor, keff should be less than 0.95 to maintain sub-criticality. The shield should provide sufficient protection for the superconducting toroidal field (TF) coil against radiation damage and heating effects of the fusion neutrons, fission neutrons, and secondary gammas. In this work, characteristics of transmutation reactor based on LAR tokamak is investigated by using the coupled system analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.473-478
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• 1997

A severe accident management strategy, in-vessel retention corium through external reactor vessel cooling(ERVC) is being studied worldwide as a means to prevent reactor vessel failure following a core melt accident. An evaluation of feasibility of this ERVC for a large Advanced Light Water Reactor (ALWR) is presented. To account for the coolability of corium and metal in the reactor vessel, a thermal analysis is performed using an existing method. Results show that the peak heat flux along the inner surface of the reactor vessel lower head has a relatively smaller margin than a small capacity reactor such as AP600 in regards with the critical heat flux attainable at the outer surface of the reactor vessel lower head.

• Journal of Engineering Education Research
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• v.15 no.1
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• pp.45-52
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• 2012

The success history of Light Water Reactors (PWR and BWR) showed how a dominant technology could be shaped in a political and economical context. The american nuclear politics, the interest of american nuclear industry, and the accumulated technological know-hows made it possible that the not inherently safe reactor-Light Water Reactor- became a prominent reactor model. The path dependency of reactor technology on LWR kept the engineers from developing a new safer reactor, even if the severe reactor accidents occurred. In oder to increase safety of nuclear power system, we should understand the social shaping process of nuclear technology.

• 한국연소학회:학술대회논문집
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• 2012.11a
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• pp.79-82
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• 2012

The gas-solid reactor, such as rotary kiln, sintering bed, incinerator and CFB boiler, is the one of most widely used industrial reactors for contacting gases and solids. the gas-solid reactor are mainly used for drying, calcining and reducing solid materials. In the gas-solid reactor, heat is supplied to the outside of the wall or inside of the reactor. The heat transfer in gas-solid reactor encompasses all the modes of transport mechanisms, that is, conduction, convection and radiation. The chemical reactions occurring in the bed are driven by energy supplied by the heat transfer. This paper deal with the effect of heat transfer and chemical reaction in the gas-solid reactor.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2001.07a
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• pp.499-502
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• 2001

In this experiment, an attempt to use the sludge pellets as catalyst for NO removal from simulated gas is experimentally investigated by using cylinder type reactor and packed-bed reactor. An experimental investigation has been conducted for NO concentration of 50[ppm], 100[ppm], 200[ppm] balanced with air, a gas flow rate of 5[1/min]. Ac voltage to discharge the gases was supplied. In the result, NOx removal rate in packed bed reactor is higher than that in cylinder type reactor. it is thought that plasma density in contact point of BaTiO$_3$ is significantly higher than that in cylinder reactor.

• Proceedings of the KIEE Conference
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• 2005.05a
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• pp.81-83
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• 2005

The Reactor Protection System (RPS) is a very important system in a nuclear power plant because the system shuts down the reactor to maintain the reactor core integrity and the reactor coolant system pressure boundary if the plant conditions approach the specified safety limits. This paper describes the unavailability assessment of a digital reactor protection system using the fault tree analysis technique. The fault tree technique can be expressed in terms of combinations of the basic event failures. In this paper, a prediction method of the hardware failure rate is suggested for a digital reactor protection system. and applied to the reactor protection system being developed in Korea.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.148-156
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• 2015

Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs) that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C) strategies for a large 1,000 MWe iPWR is described. Reactor system modeling-which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum-is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.10a
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• pp.859-862
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• 2002

The Core Support Barrel (CSB) is a major component of Reactor Internals, and is designed to support and protect the Reactor Core. In this study, Reactor Core, Core Shroud and CSB were simplified to coaxial cylinders and then the offset of Reactor Core & Core Shroud to the dynamic characteristic of CSB was analyzed. For the beam modes, natural frequencies of the cantilevered cylinder are compared with those of the cantilevered beam. And it was found out that shear modulus must be used correctly to convert the shell model to the equivalent beam model. From the dynamic characteristics of the beam model, it was found out that natural frequencies are proportional to the length of Reactor Core & Core Shroud and inversely proportional to the mass. From the comparison with the dynamic characteristics of a beam model and a lumped-mass model it was found out that the size of lumped-mass must be determined considering both the length and the mass of Reactor Core & Core Shroud.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.16 no.12S
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• pp.1292-1298
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• 2003

In this paper, dc reactor lot the inductive high-Tc superconducting fault current limiter (SFCL) was optimally designed by finite element method(FEM). The Prototype high-Tc do reactor was manufactured and compared to the results of design. This dc reactor consists of 4∼stacked double pancake coils which are wounded with Bi-2223 wire coated with SUS315L. Kapton tape is used for the insulation of turn to turn and layer to layer. Each pancake is connected in series by soldering Finally, optimal design and manufacture method lot the dc reactor is suggested in this paper. Through the comparison of result of optimal design and experimental result of prototype high-Tc superconducting dc reactor, reliance on the design of the high-Tc dc reactor tot the 1.2 kV/80 A SFCL is proved.

• Nuclear Engineering and Technology
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• v.35 no.1
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• pp.25-34
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• 2003

The research on the development of the fault monitoring system for the thermal reduction reactor has been performed preliminarily in order to support the successful operation of the thermal reduction reactor. The final task of the development of the fault monitoring system is to assure the integrity of the thermal$_3$ reduction reactor by the acoustic emission (AE) method. The objectives of this paper are to identify and characterize the fault-induced signals for the discrimination of the various AE signals acquired during the reactor operation. The AE data acquisition and analysis system was constructed and applied to the fault monitoring of the small- scale reduction reactor, Through the series of experiments, the various signals such as background noise, operating signals, and fault-induced signals were measured and their characteristics were identified, which will be used in the signal discrimination for further application to full-scale thermal reduction reactor.