• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.583-588
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• 1997

For the efficient current drive, the structure of ICRF wave propagation and absorption in a tokamak plasma should be first investigated. In this paper, two dimensional study on FWCD as well as ICRF minority ion heating for the KSTAR [Korea Superconducting Tok Amak Research] [1] plasma was performed using the full wave code of TORIC [2]. The ICRF wave propagation and absorption structures, the competitive power absorption between electrons and ions and the coupling of antenna/plasma are investigated.

• 한국수소및신에너지학회논문집
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• 제23권1호
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• pp.56-64
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• 2012

The ITER fuel cycle is designed for DT operation in equimolar ratio. It involves not only a group of fuelling system and torus cryo-pumping system of the exhaust gases through the divertor from the torus in tokamak plant, but also from the exhaust gas processing of the fusion effluent gas mixture connected to the hydrogen isotope separation in cryogenic distillation to the final safe storage & delivery of the hydrogen isotopes in tritium plant. Tritium plant system supplies deuterium and tritium from external sources and treats all tritiated fluids in ITER operation. Every operation and affairs is focused on the tritium inventory accountancy and the confinement. This paper describes the major fuel cycle processes and interfaces in the tritium plant in aspects of upcoming technologies for future hydrogen and/or hydrogen isotope utilization.

• 한국진공학회:학술대회논문집
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• 한국진공학회 1999년도 제17회 학술발표회 논문개요집
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• pp.39-39
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• 1999

The KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak is a nuclear fusion experimental device for a long pulse/steady-state plasma operation, adopting fully superconducting magnets. In accordance with completion of the basic design of the torus vacuum vessel and the enclosing cryostat, the vacuum pumping and gas fueling basic design has been developed to fulfil the physics requirements. The ultra-high vacuum pumping and sophisticated gas fueling system of the machine is essential to achieve such roles for optimized plasma performance and operation. Recently the vacuum exhaust system using dedicated pumping ports for the vacuum vessel and cryostat has been modified to meet more reliable and successful performance of the KSTAR[Fig. 1].In order to achieve the required base pressure of 5 x 10-9 torr, the total impurity load to the vessel internal is limited to ~5 x 10-5 torr-1/x, while the cryostat base pressure is kept as ~5 x 105 torr to mitigate the thermal load applied to the superconducting magnets. Each KSTAR fueling system will be separately capable of fueling gas at a rate of 50 torr-1/x, consistent with the given pumping throughput. In order to initiate a plasma discharge in KSTAR, the vacuum vessel is filled to a gas pressure of few 10-6 to few 10-4 torr, and additional gas injection is required to maintain and increase the plasma density during the course of the discharge period.

• Nuclear Engineering and Technology
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• 제40권6호
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• pp.467-476
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• 2008

To keep the superconducting (SC) magnet coils of KSTAR at proper operating conditions, not only the coils but also other cold components, such as thermal shields (TS), magnet structures, SC bus-lines (BL), and current leads (CL) must be maintained at their respective cryogenic temperatures. A helium refrigeration system (RRS) with an exergetic equivalent cooling power of 9 kW at 4.5 K without liquid nitrogen ($LN_2$) pre-cooling has been manufactured and installed. The main components of the KST AR helium refrigeration system (HRS) can be classified into the warm compression system (WCS) and the cryogenic devices according to the operating temperature levels. The process helium is compressed from 1 bar to 22 bar passing through the WCS and is supplied to cryogenic devices. The main components of cryogenic devices are consist of cold box (C/B) and distribution box (D/B). The C/B cool-down and make the various cryogenic helium for the KSTAR Tokamak and the various cryogenic helium is distributed by the D/B as per the KSTAR requirement. In this proceeding, we will present the commissioning results of the KSTAR HRS. Circuits which can simulate the thermal loads and pressure drops corresponding to the cooling channels of each cold component of KSTAR have been integrated into the helium distribution system of the HRS. Using those circuits, the performance and the capability of the HRS, to fulfill the mission of establishing the appropriate operating condition for the KSTAR SC magnet coils, have been successfully demonstrated.

• Journal of Electrical Engineering and Technology
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• 제7권2호
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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.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.207-219
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• 2022

The suppression of high-intensity blob structures in the scrape-off layer (SOL) by ion-cyclotron range of frequencies (ICRF) power, leading to a decrease in the turbulent fluctuation level, is observed first in the Experimental Advanced Superconducting Tokamak (EAST) experiment. This suppression effect from ICRF power injection is global in the whole SOL at EAST, i.e. blob structures both in the regions that are magnetically connected to the active ICRF launcher and in the regions that are not connected to the active ICRF launcher could be suppressed by ICRF power. However, more ICRF power is required to reach the full blob structure suppression effect in the regions that are magnetically unconnected to the active launcher than in the regions that are magnetically connected to the active launcher. Studies show that a possible reason for the blob suppression could be the enhanced E_r × B shear flow in the SOL, which is supported by the shaper radial gradient in the floating potential profiles sensed by the divertor probe arrays with increasing ICRF power. The local RF wave power unabsorbed by the core plasma is responsible for the modification of potential profiles in the SOL regions.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3717-3723
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• 2022

For the first time, two-dimensional temporal behavior of the edge localized mode (ELM) filament is measured in the edge tokamak plasma with a multi-channel electrical probe array (MCEP). MCEP, which has 16 floating probes (4 × 4), is mounted at the far scrape-off layer (SOL) region in the KSTAR. An electron temperature and an ion flux are measured by sideband method (SBM), which can achieve two-dimensional measurements with high time resolution. Furthermore, temporal evolutions of the electron temperature and the ion flux are obtained during the ELM occurrence. In the H-mode period, short spikes from ELM bursts are observed in measured plasma parameters, and the trend is similar to that of typical Hα signal. Interestingly, when blob-like ELM filaments crash the probe, the heat flux is significantly higher in a local region of the probe array. The results show that our probe array using the SBM can measure the ELM behavior and the plasma parameters without the effect of the stray current caused by the huge device. This study can provide valuable data needed to understand the interaction between the SOL plasma and the plasma facing components (PFCs).

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.216-221
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• 2024

Ion cyclotron range of frequency (ICRF) heating system is an important auxiliary heating method in the experimental Advanced Superconducting Tokamak (EAST). In EAST, several megawatts of power are transmitted with coaxial transmission lines and coupled to the plasma. For the long pulse and high power operation of the ICRF waves heating system, it is very important to effectively control the power and initial phase of the ICRF signals. In this paper, a power and phase feedback control system is described based on field programmable gate array (FPGA) devices, which can realize complicated algorithms with the advantages of fast running and high reliability. The transmitted power and antenna phase are measured by a power and phase detector and digitized. The power and phase feedback control algorithms is designed to achieve the target power and antenna phase. The power feedback control system was tested on a dummy load and during plasma experiments. Test results confirm that the feedback control system can precisely control ICRF power and antenna phase and is robust during plasma variations.

• 한국진공학회:학술대회논문집
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• 한국진공학회 2016년도 제50회 동계 정기학술대회 초록집
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• pp.425.2-425.2
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• 2016

토카막(TOKAMAK) 장치의 진공용기 및 용기내벽은 플라즈마(Plasma)에 의한 고열과 높은 에너지의 이온 입자들에 항상 노출되어 있는 환경이다. 토카막의 일종인 KSTAR장치의 진공용기는 스테인레스강(STS316)계열의 재질로 이루어져 있고, 플라즈마와 면하는 용기 벽면은 플라즈마에 대해 견딜 수 있도록 그라파이트 타일(graphite tile)로 구성되어 있다. 고에너지의 이온 입자들과 열플럭스(Heatflux)는 용기벽면과 용기를 침식시키고, 또한 이렇게 생겨난 분진(dust)들은 진공용기 내 여기저기를 떠다니게 되고, 플라즈마에 대해서 불순물로서 작용하게 된다. 본 연구에서는 감마분석법으로 플라즈마에 의해 진공용기 내에 집적된 분진들의 구성 성분을 분석하여 주요 출처를 규명할 수 있는 방법을 제시하고, KSTAR 플라즈마의 불순물 제어에 유용하게 활용 할 수 있는 데이터를 제공하여 향후 KSTAR의 고성능 플라즈마 기술개발에 일조할 수 있도록 하고자 한다.

• 한국초전도ㆍ저온공학회논문지
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• 제18권3호
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• pp.10-20
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• 2016

The Cable-In-Conduit-Conductor (CICC) for the ITER tokamak Central Solenoid (CS) has undergone design change since the first prototype conductor sample was tested in 2010. After tests showed that the performance of initial conductor samples degraded rapidly without stabilization, an alternate design with shorter sub-cable twist pitches was tested and discovered to satisfy performance requirements, namely that the minimum current sharing temperature ($T_{cs}$) remained above a given limit under DC bias. With consistent successful performance of ITER CS conductor CICC samples using the alternate design, an attempt is made here to revisit the internal electromagnetic properties of the CICC cable design to identify any correlation with conductor performance. Results of this study suggest that there may be a simple link between the $Nb_3Sn$ CICC internal self-field and its $T_{cs}$ performance. The study also suggests that an optimization process should exist that can further improve the performance of $Nb_3Sn$ based CICC.