• Journal of the Korean Vacuum Society
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• v.16 no.6
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• pp.489-495
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• 2007

The assembly of the main system of the KSTAR tokamak has been recently completed, and the preparation for the 1st plasma and test operations is progressed. The fueling system established for these purposes uses only one port placed at the opposite side of the pumping duct, and has a difficulty of attaining a uniform and fast supply of fuel particles to the plasma. At the base operation stage after finishing the test operation, the fueling system must be improved to provide a uniform fueling and a feed-back control in accordance with a high-density tokamak plasma maintained for a long period. As a part for understanding the points to be improved in the fueling system, a Monte Carlo simulation on the gas fueling into the tokamak plasma has been executed. After modeling the vacuum vessel and the plasma of quasi-D shapes as tori of rectangular cross-sections, the influences of the position and the number of the fueling inputs on the particle density distribution for a given pumping probability and mean free path were investigated.

• 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.

• Journal of Sensor Science and Technology
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• v.25 no.1
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• pp.46-50
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• 2016

During KSTAR plasma experiments, torsional $Alfv\acute{e}n$ waves in the frequency of few GHz or below were detected. To understand this plasma waves during the crash of MHD instabilities, an RF spectrometer has been developed for detection of the radiated RF signals in the KSTAR Tokamak. It has the capability of broadband RF spectral measurement (50 ~ 400 MHz). To detect the broadband RF signals which are radiated from the KSTAR systems, a broadband antenna is the key feature of the RF spectrometer. In this paper, a broadband bowtie antenna for detection of $Alfv\acute{e}n$-waves in the KSTAR Tokamak is presented. Planar-type bowtie antenna is designed and fabricated on an FR4 substrate with thickness of 1.6 mm. The antenna consists of bowtie shaped balanced radiators and broadband planar balun. The antenna is designed to have an input impedance of 50 Ohm, and a taper-shaped balun is adopted for field and impedance matching between 50 Ohm transmission line to 110 Ohm feeding network of balanced radiators. The implemented antenna provides around -3 to 3 dBi of gain for the whole frequency band. The VSWR of the bowtie antenna is less than 12:1 over the frequency bandwidth of 50 to 2000 MHz.

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.1098-1102
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• 2003

Vacuum vessel of the KSTAR(Korea Superconducting Tokamak Advanced Research) tokamak is a fully welded structure with D-shaped cross-section. According to the requirements of the physics design, sixteen horizontal ports, sixteen slanted ports, sixteen baking and cooling ports, and twenty-four top and bottom vertical ports are designed for the diagnostics, plasma heating, vacuum pumping, and baking and cooling. Bellows on these ports are used for flexible components to absorb the relative displacement due to the vacuum vessel thermal expansion and the electromagnetic force between the vacuum vessel and the cryostat ports. Fatigue strength evaluation was performed to decide the dimension of the bellows. In order to assure the quality of the bellows, a prototype bellows for the neutral beam injection port has been fabricated and tested prior to main fabrication. It was conformed that the prototype bellows has sufficient fatigue strength and vacuum reliability in the expected load conditions.

• 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.

• Nuclear Engineering and Technology
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• v.29 no.6
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• pp.459-467
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• 1997

The steady-state behavior of recycling neutral atoms in a tokamak edge region has been analyzed through a two-dimensional Monte Carlo simulation. A particle tracking algorithm used in earlier research on the neutral particle transport is applied to this Monte Carlo simulation in order to perform more accurate calculations with the EDGETRAN code which was previously developed for a two-dimensional edge plasma transport in the authors' laboratory. The physical model of neutral recycling includes charge-exchange and ionization interactions between plasmas and neutral atoms. The reflection processes of incident particles on the device wall are described by empirical formulas. Calculations for density, energy, and velocity distributions of neutral deuterium-tritium atoms have been carried out for a medium-sized tokamak with a double-null configuration based on the KT-2 conceptual design. The input plasma parameters such as plasma density, ion and electron temperatures, and ion fluid velocity are provided from the EDGETRAN calculations. As a result of the present numerical analysis, it is noticed that a significant drop of the neutral atom density appears in the region of high plasma density and that the similar distribution of neutral energy to that of plasma ions is present as frequently reported in other studies. Relations between edge plasma conditions and the neutral recycling behavior are discussed from the numerical results obtained herein.

• The Transactions of the Korean Institute of Electrical Engineers C
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• v.54 no.3
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• pp.120-126
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• 2005

The Korean Superconducting Tokamak Advanced Research (KSTAR) tokamak device is being constructed to perform long-pulse, high-beta, advanced tokamak fusion physics experiments. The long-pulse operation requires the non-inductive current drive system such as the Lower-Hybrid Current Drive (LHCD) system. The LHCD system drives the non-inductive plasma current by means of C-band RF with 2-MW CW power and 5-GHz frequency. For the LHCD test experiments, an RF test system is developed. It is composed of a 5-GHz, 1.5-MW pulsed magnetron and a compact pulse modulator with $4\;{\mu}s$ of pulse width. The pulse modulator provides the maximum output voltage of 45 kV and the maximum current of 90 A. It is composed of 7 stages of Pulse Forming Network (PFN), a thyratron tube (E2V, CX1191D), and a pulse transformer with 1:4 step-up ratio. In this paper, the detailed design and the performance test of the pulse modulator are presented.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.136-147
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• 1993

The paper describes : ⅰ) Mathematical modeling of poloidal flux to define and calculate the tokamak plasma position based on a property of the plasma boundary which is always a flux surface. Controlling the plasma boundary position is therefore equivalent to equalizing the flux value on several points belonging to a curve tangent to the limiter. ⅱ) Experimental method for determining the outmost poloidal isoflux surface by a linear combination of measurements of magnetic fluxes, fields and field gradients, without requiring knowledge of internal plasma parameters for the feedback control, i.e., with neither corrections for variation in the poloidal beta and the plasma current distribution, nor compensations for the induced currents in the vacuum vessel. ⅲ) Feedback control algorithm for the regulation of plasma boundary position and its electronics hardware based on the PID control theory. ⅳ) Experimental results obtained from the RTP tokamak experiments using the present plasma control system.

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.32-39
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• 1989

A method for determining the optimum ideal MHD $\beta$limit and the operation conditions is presented for an axisymmetric tokamak reactor with a circular cross section. The $\beta$limit is calculated under the constraints of ideal MHD instability criteria with varying the operation conditions which depend on the toroidal current density distributions. Semiempirical laws deduced from experimental observations are used for the toroidal current density distributions. Analytic derivations of various equations required to determine the $\beta$limit are carried out from the empirical equations. Various distributions of the $\beta$limit are obtained by the numerical calculations for different distributions of the toroidal current density. The resulting values of the maximum $\beta$limited by ideal MHD instabilities are expressed by a scaling law in terms of the tokamak geometry and the safety factor.

• 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.