• Nuclear industry
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• no.9_10 s.9
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• pp.26-32
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• 1982

• Proceedings of the Korean Vacuum Society Conference
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• 1998.06a
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• pp.152-152
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.511-511
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• 2001

• Progress in Superconductivity and Cryogenics
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• v.3 no.1
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• pp.45-49
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• 2001

A magnet supporting post installed between the lower TF coil tooled by 4.5 K supercritical helium and the cryostat base is one of the most important components of the superconducting magnet supporting structure for KSTAR Tokamak. This structure should be flexible to absorb thermal shrink of the magnet and also should be rigid to support the magnet weight and the Plasma disruptions load. The Post was designed with stainless steel 316LN and CFRP that have low thermal conductivity and high structural strength at low temperature. In order to verify the possibility of fabrication and the structural safety. a whole scale prototype of the KSTAR magnet supporting post was manufactured and tested. Static and compressive cyclic load tests under the maximum Plasma vertical disruption load and the magnet dead weight were performed. The teat results showed that the magnet supporting post of KSTAR Tokamak was possible to manufacture and structurally rigid.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.545-550
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• 2004

KSTAR cryostat is a 8.8 m diameter vacuum vessel that provides the necessary thermal barrier between the ambient temperature test cell and the supercritical helium cooled superconducting magnet providing the base pressure of 1 ${\times}$ $10^{-3}Pa$. The cryostat is a single walled vessel consisting of central cylindrical section and two end closures, a flat base structure with external reinforcements and a dome-shaped lid structure. The base structure has 8 equally spaced support legs anchored on the concrete base. The cryostat vessel design was executed to satisfy the performance and operation requirements. The major loads considered in the structural analysis were vacuum pressure, dead weight, electromagnetic load driven by plasma disruption, and seismic load. Based on the fabrication and inspection procedures for the vessel, cryostat vessel was fabricated and inspected. It was confirmed that the inspection results were acceptable.

• Vacuum Magazine
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• v.4 no.1
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• pp.16-23
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• 2017

Recently, KSTAR, Korea's superconducting fusion energy research and development device, has succeeded in driving the high performance plasma for 70 seconds for the first time in the world. Continuous plasma operation technology is an essential factor for commercialization of fusion energy power generation. Therefore, this achievement is expected to play a major role in the research of fusion technology required for future fusion power plants. In order to operate the KSTAR, the discharge process in which the neutral gas is turned into the plasma should be preceded in the start-up (breakdown) phase of tokamak operation. This process essentially involves the vacuum environment in the tokamak device. KSTAR has successfully operated a vacuum pumping system to achieve the target level of the vacuum environment through a high temperature baking operation, a discharge cleaning process and boronization.

• Journal of the Korean Vacuum Society
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• v.8 no.4A
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• pp.397-402
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• 1999

The base pressure of the vacuum vessel of KSTAR tokamak is to be ultra high vacuum, $10^{-6}\sim10^{-7}Pa$, to produce a clean plasma with low impurity concentrations. For this purpose, vessel and plasma facing components need to be baked up to $250^{\circ}C$, $350^{\circ}C$ respectively to remove impurities from the plasma-material interaction surfaces. Here the required heating power to be supplied for baking has been calculated according to pre-assumed different temperature profiles (baking scenario and proper baking plan for KSTAR tokamak has been proposed. Mass flow rate and temperature of nitrogen gas for baking has also been calculated.

• Journal of the Korean Vacuum Society
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• v.8 no.4B
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• pp.497-506
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• 1999

Carbon has been widely used for the material of plasma facing components in fusion experiment devices like a tokamak, because carbon has good thermal and mechanical properties. However carbon gas a relatively high ougassing rate. Therefore the amount and the surface area of the carbon material used in the vessel will determine the background pressure of the vacuum vessel. In this experiment influences of carbon on the vacuum performance was investigated by measuring chamber pressure, ougassing rater and gas spectrum of carbon fiber composite (CFC) samples in various situations, pumping out, chamber baking, carbon heating (250~$500^{\circ}C$), exposure to atmosphere for maintenance of in-vessel components, etc., occurring routinely during tokamak operations.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.348-359
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• 1997

This paper reviews cooling aspects of the diverter system in Tokamak fusion devices with primary emphasis on the critical heat flux (CHF) issues for oater-cooled designs. General characteristics of four (4) coolant options for diverter cooling gases, oater, liquid metal, and organic liquid - are discussed first, focusing on the comparison of advantages and disadvantages of those options. Then results of recent studies on the high-heat-flux CHF of water at subcooled high-velocity conditions are reviewed to provide a general idea on the feasibility of the water-cooled diverter concept for future Tokamak fusion reactors. Water is assessed to be the most viable and practical coolant option for diverters of future experimental Tokamaks.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.323-327
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• 2020

In this study, the neutronic calculation to obtain tritium breeding ratio (TBR) in a deuterium-tritium (D-T) fusion power reactor using Monte Carlo MCNPX is done. In addition, by using COMSOL software, an efficient cooling system is designed. In the proposed design, it is adequate to enrich up to 40% ⁶Li. Total tritium breeding ratio of 1.12 is achieved. The temperature of helium as coolant gas never exceed 687℃. As regards the tolerable temperature of beryllium (650℃), the design of blanket module is done in the way that beryllium temperature never exceed 600℃. The main feature of this design indicates the temperature of helium coolant is higher than other proposed models for blanket module, therefore power of electricity generation will increase.