• Proceedings of the KIPE Conference
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• 2015.07a
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• pp.223-224
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• 2015

본 논문에서는 국제 핵융합실험로용 TF 컨버터의 전류제어에 대하여 서술하였다. TF 컨버터는 도넛형 진공용기 내부에 직류 자기장을 발생시켜 플라즈마를 진공용기 내에 가두어 주는 18개의 TF 코일에 전류를 공급한다. 68kA의 직류전류를 17H의 초전도 코일에 공급하기 위해 TF 컨버터는 666V의 계통전원으로 급속 충전 방전의 동작과 333V의 계통전원으로 완속 충전 방전의 동작을 수행한다. 이러한 전류제어 프로파일을 만족하는 TF 컨버터의 전류제어기를 설계하였고, 이를 실제 제어기와 RTDS를 연동하는 HIL 시스템을 구축하여 검증하였다.

• Nuclear Engineering and Technology
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• v.16 no.2
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• pp.89-96
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• 1984

A toroidal-field (TF) coil with a pure tension D-shape curve is designed for the confinement of high-temperature plasmas in the SNUT-79, which is a tokamak being built at Seoul National University. A toroidal assembly of 16 D-shape TF coils is designed to produce the magnetic field of up to 3T, of which ripples appear to be below 4% of the average toroidal field in the plasma region. Exact positions and currents in six equilibrium coils distributed symmetrically in the z=0 plane are found by the solution of a set of linear equations which is transformed from a Fredholm integral equation of the first kind. The decay indices resulted from equilibrium field indicate that the stability condition for vertical and horizontal displacements is satisfied.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.08a
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• pp.314-314
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• 2010

For the optimal design of a tokamak-type reactor, self-consistent determination of a radial build of reactor systems is important and the radial build has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor systems. In a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil, the shield should provide sufficient protection for the superconducting TF coil and the shield plays a key role in determining the size of a reactor. To determine the radial build of a reactor, neutronic effects such as tritium breeding in the blanket, nuclear heating, and radiation damage to toroidal field (TF) coil has to be included in the systems analysis. In this work, the outboard blanket only is considered where tritium self-sufficiency is possible by using an inboard neutron reflector instead of breeding blanket. The reflecting shield should provide not only protection for the superconducting TF coil but also improved neutron economy for the tritium breeding in outboard blanket. Tungsten carbide, metal hydride such as titanium hydride and zirconium hydride can be used for improved shielding performance and thus smaller shield thickness. With the use of advanced technology in the shield, conceptual design of a compact superconducting LAR reactor with aspect ratio of less than 2 will be presented as a viable power plant.

• Proceedings of the SAREK Conference
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• 2008.11a
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• pp.413-418
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• 2008

The toroidal field magnet power supply (TF MPS) for the KSTAR was constructed in August, 2007 and started the operation for the commissioning in March, 2008. The main role of the TF MPS is to supply the electric power to the TF magnet of the KSTAR. The water cooling components of the TF MPS are 16 stacks, busbar of 70 meters. For the cooling of the TF MPS, the D I water cooling system was designed and installed. The water cooling system consists of several pumps, heat exchangers, D I water polishing system and so on. The water cooling system for the TF MPS was tested under the 15 kA current charging condition. In this paper be discussed about the system performance and other parameters.

• Journal of Electrical Engineering and Technology
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• v.7 no.2
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• pp.178-183
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• 2012

The design of the integrated quench protection (QP) system for the high current superconducting magnet (SCM) has been fabricated and tested for the toroidal field (TF) coil system of the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The QP system is capable of protecting the TF SCM, which consists of 16 identical coils serially connected with a stored energy of 495 MJ at the design operation level at 35.2 kA per turn. Given that the power supply for the TF coils can only ramp up and maintain the coil current, the design of the QP system includes two features. The first is a basic fast discharge function to protect the TF SCM by a dump resistor circuit with a 7 s time constant in case of coil quench event. The second is a slow discharge function with a time constant of 360 s for a daily TF discharge or for a stop demand from the tokamak control system. The QP system has been successfully tested up to 40 kA with a short circuit and up to 34 kA with TF SCM in the second campaign of KSTAR. This paper describes the characteristics of the TF QP systems and test results of the plasma experiment of KSTAR in 2009.

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
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• 2003.10a
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• pp.307-310
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• 2003

The KSTAR (Korea Superconducting Tokamak Advanced Research) prototype TF (Toroidal Field) coil was tested in the superconducting coil test facility in KBSI (Korea basic Science Institute). The test was divided into several campaigns according to the objectives. The objectives of the first campaign were to cool the coil into operating temperature and to find any defect in the coil such as cold leaks. From the results of the first campaign, which was carried out during Jan. 2003, any defect in the TF prototype coil was not found. At the second campaign, the large-current charging experiment was one of the major issues, and was carried out during Aug. 2003 In this paper, the test preparation, and the test results of the second campaign were presented.

• Progress in Superconductivity
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• v.10 no.2
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• pp.108-115
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• 2009

As a participant taking part in the ITER TF conductor R&D program, we developed two toroidal field conductors with variations of conduit thickness resulting in the different void fraction of the conductors. The estimated void fractions of the conductors are 31% and 33%. In this paper we present the details of the TF conductor development and performance test results of them carried out by the measurement of current sharing temperature under cyclic loading. Regarding the conductor development, the internal-Sn-processed $Nb_3Sn$ strand characteristics, strand cabling, twist pitch and characteristics of the conduit materials are presented. For the understanding of the conductor design and performance, the conductor test results are presented and the effect of the conductor design parameters such as void fraction and twist pitch is discussed based on the results.

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
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• 2003.10a
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• pp.298-300
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• 2003

The KSTAR superconducting magnet system consists of 16 TF (Toroidal Field) and 14 PF (Poloidal Field) coils. Both of the TF and PF coil system use internally-cooled Cable-In-Conduit Conductors (CICC). The major achievement in KSTAR magnet system development includes the development of CICC, the development of a full size TF model coil, the development of a background magnetic field generation coil system, the construction of a large scale superconducting magnet. TF and PF coils are in the stage of the fabrication for the KSTAR completion in the year 2005.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.02a
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• pp.314-314
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• 2010

The design of the blanket and shield play a key role in determining the size of a reactor since it has an impact on the various reactor components. The blanket should produce enough tritium for tritium self-sufficiency and the shield should provide sufficient protection for the superconducting TF coil. Neutronic optimization of the blanket and the shield is necessary, and we coupled the system analysis with a neutronic calculation to account for the interrelation of the blanket and shield with the plasma performance of a reactor system in a self-consistent manner. By using the coupled system analysis code, the operational space for a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil is investigated with an spect ratio in the range of 1.5 - 2.5. The minimum major radius which satisfies all the physics and engineering requirements increases with the magnetic field at the magnetic axis. A required inboard shield thickness is mainly determined by the requirement on the protection of the TF coil against radiation damage. It is shown that to have a fusion power bigger than 3,000 MW in the LAR tokamak with a superconducting TF coil, a major radius bigger than 4.0 m is required.

• Progress in Superconductivity and Cryogenics
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• v.8 no.2
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• pp.25-28
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• 2006

The magnet system of KSTAR(korean Superconducting Tokamak Advanced Research) is consisted of 16 TF (Toroidal Field) coils and 14 PF (Poroidal Field) coils. Internal cooling CICC(Cable in Conduit Conductor) type conductor is used for both of TF and PF coil systems. The conduit material for $Nb_3Sn$ cable is Incoloy 908 and 316LN stainless-steel was used as conduit material for NbTi cable. $Nb_3Sn$ CICC is used for all TF coils and PF1-5 coils while NbTi CICC is used for PF6 and 7 coils. $Nb_3Sn$ and NbTi strands were made for KSTAR superconducting strand. They are satisfied with KSTAR superconducotr requirements. The $Nb_3Sn$ strands supplied from three companies; MELCO (Mitsubishi Electric Co.), OAS (Outokumpu Advanced Superconductor) and KAT (Kiswire Advanced Technology) were used. A special CICC jacketing system is developed for the KSTAR CICC fabrication which uses the tube-mill process consisted of forming, welding, sizing and squaring procedures. The. procedures for cabling and jacketing of CICC for TF and PF coils and their results including the geometrical specification and characteristics of strands are described.