• Journal of the Korean Vacuum Society
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• v.16 no.5
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• pp.388-396
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• 2007

KSTAR (Korea Superconducting Tokamak Advanced Research) current lead system (CLS) has a role to interconnect magnet power supply (MPS) in room temperature (300 K) and superconducting (SC) bus-line, electrically. For the first plasma experiments, it should be assembled 4 current leads (CL) on toroidal field (TF) current lead box (CLB) and 14 leads on poloidal field (PF) CLB. Two current leads, with the design currents 17.5 kA, and SC bus-lines are connected in parallel to supply 35 kA DC currents on TF magnet. Whereas, it could supply $20\;{\sim}\;26\;kA$ to each pairs of PF magnets during more than 350 s. At the cold terminals of the leads, there are joined SC bus-lines and it was constructed helium coolant control system, aside from main tokamak system, to protect heat flux through current leads and enhanced Joule heat due to supplied currents. Throughout the establishment processes, it was tested the high vacuum pumping, helium leak of the helium lines and hardwares mounted between the helium lines, flow controls for CL, and liquid nitrogen cool-down of possible parts (current leads, CL helium lines, and thermal shield helium lines for CLB), for the accomplishment of the required performances.

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.296-301
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• 2003

A study on the engineering design and numerical thermo-hydraulic analysis for KSTAR TF coil structure cooling system has been conducted. The numerical analyses have been done to verify the engineering design of cooling using the commercial code, FLUENT and in-house code for calculating helium properties which varies with cooling tube's heat transfer. Through the engineering design process based on the steady heat balance concepts, the circular stainless steel tube with inner diameter of 4 mm for TF coil has been selected as cooling tube. From normal operation mode analysis results, total 28 cooling tubes were finally chosen. Also, three dimensional cool down analysis for TF coil with designed cooling tube was satisfied with next three design criteria. First is cooling work termination within a month, second is maximum temperature difference within 50 K in TF coil structure and third is exit helium pressure above 2 bar. Consequently, these cool down scenario results can afford to adopt as operating scenario data when KSTAR facilities operate.

• Proceedings of the Korean Vacuum Society Conference
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• 2011.02a
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• pp.397-397
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• 2011

In a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil, the radial build of TF coil and the shield play a key role in determining the size of a reactor. For self-consistent determination of the reactor components and physics parameters, a system analysis code is coupled with one-dimensional radiation transport code. Conceptual design study of a compact superconducting LAR tokamak reactor with aspect ratio less than 2.5 was conducted and the optimum radial build was identified. It is shown that the use of an improved shielding material and high temperature superconducting magnets with high critical current density opens up the possibility of a fusion power plant with compact size and small re-circulating power simultaneously at low aspect ratio, and that by using an inboard neutron reflector instead of breeding blanket, tritium self-sufficiency is possible with outboard blanket only and thus compact sized reactor is viable.

• Proceedings of the KIPE Conference
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• 2006.06a
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• pp.435-437
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• 2006

KSTAR (Korea Superconducting Tokamak Advanced Research) 장치는 Tokamak 개념의 핵융합 연구 장치로서 플라즈마를 가두기 위한 자장을 발생하는 토로이달 자장(Toroidal Field, TF) 코일과 플라즈마 발생 및 형상 조정을 위한 폴로이달 자장 (Poloidal Filed, PF) 코일로 구성되며, 초전도 코일을 이용한다. TF코일의 전원장치로는 40 kA급의 안정된 직류 전원장치가 필요하며, PF 코일의 전원장치로는 빠른 전류상승 및 피드백 기능을 갖춘 정밀 대전류 전원을 필요로 한다. 또한 초전도 코일의 ??????치현상 발생 시 코일과 전원장치 보호를 위한 대전류 직류 차단시스템을 필요로 한다. KSTAR 장치의 설계에 의하면 상하 7쌍의 초전도 PF 코일에 약 1MA/sec급의 고속 전류구동을 운전 시나리오에 따라 인가하여 핵융합 연구를 위한 플라즈마를 생성한다. 본 논문은 TF 및 PF 코일에 대전류를 인가하기 위해서 개발된 전원장치 (Power Supply, PS)에 관한 연구이다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.425-431
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• 1996

KT-2 poloidal-field (PF) system is designed to cope the up-down symmetric double-null (DN) and asymmetric single-null (SN) discharges with typical plasma parameters, in which three sets of "design-basis" scenarios - the ohmic heating (OH), the 5MW and the high bootstrap (HIBS) baseline modes - are applied. The power and energy demand for each cases are also deduced. The peak power and the maximum energy requirements for the KT-2 magnet system, incorporating the PF and the toroidal-field (TF) coils, are proven to be 123MW and 1601MJ, respectively when it is driven in DN configuration. The KT-2 PF system is capable of achieving the machine mission of creating a 500kA heated plasma with a current flattop of $\geq$20 seconds.

• Journal of Electrical Engineering and Technology
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• v.8 no.2
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• pp.326-330
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• 2013

The Korea Superconducting Tokamak Advanced Research (KSTAR) device is an advanced superconducting tokamak to establish scientific and technological bases for attractive fusion reactor. This device requires 3.5 Tesla of toroidal field (TF) for plasma confinement, and requires a strong poloidal flux swing to generate an inductive voltage to produce and sustain the tokamak plasma. KSTAR was originally designed to have 16 serially connected TF magnets for which the nominal current rating is 35.2 kA. KSTAR also has 7 pairs of poloidal field (PF) coils that are driven to 1 MA/sec for generation of the tokamak plasma according to the operation scenarios. The KSTAR Magnet Power Supply (MPS) was dedicated to the superconducting (SC) coil commissioning and $2^{nd}$ plasma experiment as a part of the system commissioning. This paper will describe key features of KSTAR MPS for the $2^{nd}$ plasma experiment, and will also report the engineering and commissioning results of the magnet power supplies.

• 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 Vacuum Society Conference
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• 2012.02a
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• pp.582-582
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• 2012

Characteristics of a fusion-driven transmutation reactor was investigated. A compact reactor concept is desirable from an economic viewpoint. 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. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

• Progress in Superconductivity and Cryogenics
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• v.14 no.2
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• pp.24-27
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• 2012

The superconducting coil system is one of the most important components in Korea Superconducting Tokamak Advanced Research (KSTAR), which has been operated since 2008. $Nb_3Sn$ and NbTi superconductors are being used for cable-in-conduit conductors (CICCs) of the KSTAR toroidal field (TF) and poloidal field (PF) coils. The CICCs are cooled by forced-flow supercritical helium about 4.5 K. The temperature, pressure and mass flow rate of the supercritical helium in the CICCs are interacting with each other during the operation of the coils. The complicate behaviors of the supercritical helium have an effect on the operation and the efficiency of the helium refrigeration system (HRS) by means of, for instance, pressure drop. The hydraulic characteristics of the supercritical helium have been monitored while the TF coils have stably achieved the full current of 35 kA. In other hands, the PF coils have been operated with various pulsed or bipolar mode, so the drastic changes happen in view of hydraulics. The heat load including AC loss on the coils has been analyzed according to the measurement. These activities are important to estimate the temperature margin in various PF operation conditions. In this paper, the latest hydraulic behaviors of PF coils during KSTAR operation are presented.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.975-980
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• 2001

Like composite material. the coil winding pack of the KSTAR (Korea Superconducting Tokamak Advanced Research) consist of multiphase element such as metallic jacket material for protecting superconducting cable, vacuum pressurized imprepregnated (VPI) insulation, and corner roving filler. For jacket material, four CS (Central Solenoid) Coils, $5^{th}$ PF (Poloidal Field) Coil, and TF (Toroidal Field Coil) use Incoloy 908 and $6-7^{th}$ PF coil, Cold worked 316LN. In order to analyze the global behavior of large coil support structure with coil winding pack, it is required to replace the winding pack to monolithic matter with the equivalent mechanical properties, i.e. Young's moduli, shear moduli due to constraint of total nodes number and element numbers. In this study, Equivalent Young's moduli, shear moduli, Poisson's ratio, and thermal expansion coefficient were calculated for all coil winding pack using Finite Element Method.