• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
• /
• 2003.10a
• /
• pp.301-303
• /
• 2003

The KSTAR(Korea Superconducting Tokamak Advanced Research) superconducting magnet system consist of 16 TF(Toroidal Field) and 14 PF(Poloidal Field) coils. Internally-cooled cabled superconductors will be used for the magnet system. The magnet system adopt a superconducting CICC (Cable-In-Conduit Conductor) type. The KSTAR PF 6, 7 CICCs use NbTi Superconducting cable with stainless steel 316LN conduit while the other PF CICC use Incoloy 908 conduit. For the fabrication of PF CICC, superconducting cables have been fabricated and the cable has the diameter of 22.3mm. A continuous CICC jacketing system is developed for the CICC jacketing and the jacketing system uses the tube-mill process, which consists of forming, welding, sizing and squaring procedures. The cabling and the jacketing process is described. The welding condition and design specification of CICCs are also discussed. The fabrication results including the geometrical specification and the void fraction will be discussed.

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

• Proceedings of the Korean Institute of IIIuminating and Electrical Installation Engineers Conference
• /
• 2005.11a
• /
• pp.151-155
• /
• 2005

Though the nuclear fusion system has been fused into hydro-nuclear based on thermodynamics by tokamak system, there has been no success story. Because it's impossible to confine high temperatured plasma long Time actually. New nuclear-fusion-system using this nuclear-fusion-method will gather toroidal-magnetic-field by putting Core Block(C shaped torus iron) and toroidal-aluminium coil into toroidal magnetic-field-aluminium. That will arrange the nuclear-fusion-route on a gap fallen out by a part of cut torus-core and mkee the toroidal-an electric-current flow and electrolyze the fusioned-material (an electrolyte) into troidal-electrocity. That consists of troidal-magnetic-fild coil, toroidal-coial fusioned- material, series circuit. So toroidal-electocity will be changed into filament-electrocity and be introjected into fusioned-material. In a sapce on filament-electrocity, the magnetic inhaling-powr will enlarge to input-electrocity outside. This will exceed the Coulomb force and reache the nuclear-fusion. By this phenomenon there be quantity-loss. By this process I could confirmed that Einstein euation$(E=mC^2)$ releases into thermal energy.

• Proceedings of the Korean Vacuum Society Conference
• /
• 2010.02a
• /
• pp.295-295
• /
• 2010

KSTAR(Korea Superconducting Tokamak Advanced Research) 토카막에 설치되어 있는 ICRF(Ion Cyclotron Range Frequency) 시스템을 이용한 방전세정을 2008년에 이어 2009 KSTAR 플라즈마 campaign 동안에도 시행하였다. ICRF 시스템을 이용한 방전세정인 ICWC(Ion Cyclotron Wall Cleaning)는 ITER와 DEMO 같은 초전도 자석을 이용하는 토카막에서 토카막 shot 중간에 자장을 낮추지 않고 바로 방전 세정을 할 수 있는 방법이다. 토카막에서 방전세정은 탄소나 산소 화합물과 같은 불순물을 제거하여 방사에 의한 플라즈마 냉각을 막고 토카막 초기 start-up시 진공 챔버 벽면으로부터 의도하지 않은 연료주입을 제거하는 역할을 한다. 본 연구에서는 ICWC 방전 세정 플라즈마의 밀도특성과 균일도를 간섭계와 $H_{\alpha}$ line 세기를 통해 관측하고 RGA를 통해서 C, $H_2O$, $O^2$ 불순물의 제거량을 파악하는 한편 토카막의 신뢰성 있는 start-up을 위해 요구되는 벽면에서 토카막 방전가스의 제거량을 HD양을 통해서 조사하였다. 플라즈마 선적분 밀도는 약 $1{\sim}3{\times}10^{17}#/m^2$로 측정되었는데 이는 보통 He을 이용한 방전세정 플라즈마의 밀도에 해당한다. 한편 $H_{\alpha}$ line의 세기를 통해 ICWC 방전 플라즈마의 균일도를 살펴본 결과 안테나 전류띠의 중간이 아닌 끝부분에서 $H_{\alpha}$의 세기가 큰 것으로 나타났는데 이는 ICWC 플라즈마가 Inductive 방전보다는 capacitive 방전에 의해 생성되는 것으로 추정된다. ICWC 방전에서 C, $H_2O$, $O_2$ 불순물의 제거율은 각각 약 $4.2{\times}10^{-5}\;mbar{\cdot}l/sec$, $1.4{\times}10^{-3}\;mbar{\cdot}l/sec$ 그리고 $1.72{\times}10^{-4}\;mbar{\cdot}l/sec$로 각각 나타났는데 ICWC shot이 진행될수록 이 양은 점점 줄어들었다. 대표적인 He/$H_2$, He ICWC 방전 shot인 2118, 2123 shot에서 벽면에서 $D_2$의 제거율은 각각 약 $0.12\;mbar{\cdot}l/sec$와 $3.9{\times}10^{-3}\;mbar{\cdot}l/sec$로 나타났다. 이는 수소의 첨가로 인해 HD의 형태로 $D_2$의 제거율이 증가되었기 때문이다. 한편 $H_2$의 첨가는 챔버 벽면에 흡착되는 $H_2$ 양을 또한 증가시키므로 차후에 $H_2$ 만을 제거하는 He ICWC를 수행해야 할 것이다.

• Proceedings of the Korean Vacuum Society Conference
• /
• 2013.08a
• /
• pp.179.2-179.2
• /
• 2013

A large-area RF-driven ion source is being developed at Germany for the heating and current drive of ITER device. Negative hydrogen ion sources are major components of neutral beam injection (NBI) systems in future large-scale fusion experiments such as ITER and DEMO. The RF sources for the production of positive hydrogen ions have been successfully developed at IPP (Max-Planck-Institute for Plasma Physics), Garching, for the ASDEX-U and W7-AS neutral beam heating systems. Ion sources of the first NBI system (NBI-1) for the KSTAR tokamak have been developed successfully with a bucket plasma generator based on the filament arc discharge, which have contributed to achieve a good plasma performance such as 15 sec H-mode operation with an injection of 3.5 MW NB power. There is a development plan of RF ion source at the KAERI to extract the positive ions, which can be used for the second NBI system (NBI-2) of the KSTAR and to extract the negative ions for future fusion devices such as Fusion Neutron Source and Korea-DEMO. The development progresses of RF ion source at the KAERI are described in this presentation.

• Progress in Superconductivity and Cryogenics
• /
• v.24 no.2
• /
• pp.7-18
• /
• 2022

European efforts to design superconducting conductors for a future tokamak have involved Nb₃Sn cable-in-conduit conductor (CICC). Nb₃Sn coils which undergo heat treatment to activate the Nb₃Sn material are mostly produced through the wind-then-react route. However, some Nb₃Sn coils have been proposed with CICCs of the react-then-wind route. The latter CICCs are physically constrained due to handling limitations which if not adhered to will result in irrecoverable damage to the Nb₃Sn cable inside, nullifying any performance advantage. A group at the Swiss Plasma Center has proposed such CICC designs, constructing samples and testing them for performance. The characteristics and performance of these react & wind (R&W) CICCs are compared with the more common wind & react (W&R) CICCs, and it is found that the R&W designs show more extreme characteristics than typical W&R Nb₃Sn CICCs for some parameters that are known to influence CICC performance. Where the R&W CICCs extend the range of those parameters, they also continue trends formed by the W&R CICCs with the parameters. The main observation, however, is that although the current sharing temperature performances of the R&W samples are above the average of the W&R samples they were compared to, they are not the highest. A similar observation applies to a cost comparison of the superconducting material where the R&W CICCs are found to be relatively cheap but not the cheapest. Given these results, clear practical advantages to the R&W CICC design is not evident.

• Nuclear Engineering and Technology
• /
• v.53 no.7
• /
• pp.2174-2183
• /
• 2021

Helium cooled solid breeder blanket as an important blanket candidate of the Tokamak fusion reactor uses ceramic pebble bed for tritium breeding. Considering the poor effective thermal conductivity of the ceramic breeder pebble bed, thin structure of tritium breeder pebble bed is usually adopted in the blanket design. The container wall has a great influence on the thin pebble bed packing structure, especially for the assembly of mono-sized particles, and thin pebble bed will appear anisotropic effective thermal conductivity phenomenon. In this paper, thin ceramic pebble beds composed of 1 mm diameter Li₄SiO₄ particles are generated by the EDEM 2.7. The effective thermal conductivity of different thickness pebble beds in the three-dimensional directions are analyzed by three-dimensional thermal network method. It is observed that thin Li₄SiO₄ pebble bed showing anisotropic effective thermal conductivity under the practical design size. Normally, the effective thermal conductivity along the bed vertical direction is higher than the horizontal direction due to the gravity effect. As the thickness increases from 10 mm to 40 mm, the effective thermal conductivity of the pebble bed gradually increases.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4134-4145
• /
• 2022

The collimator is one of the high-heat-flux components used to avoid a series of vacuum and thermal problems. In this paper, the heat load distribution throughout the collimator is first calculated through experimental data, and a transient thermodynamic simulation analysis of the original model is carried out. The error of the pipe outlet temperature between the simulated and experimental values is 1.632%, indicating that the simulation result is reliable. Second, the model is optimized to improve the heat transfer performance of the collimator, including the contact mode between the pipe and the flange, the pipe material and the addition of a twisted tape in the pipe. It is concluded that the convective heat transfer coefficient of the optimized model is increased by 15.381% and the maximum wall temperature is reduced by 16.415%; thus, the heat transfer capacity of the optimized model is effectively improved. Third, to adapt the long-pulse steady-state operation of the experimental advanced superconducting Tokamak (EAST) in the future, steady-state simulations of the original and optimized collimators are carried out. The results show that the maximum temperature of the optimized model is reduced by 37.864% compared with that of the original model. The optimized model was changed as little as possible to obtain a better heat exchange structure on the premise of ensuring the consumption of the same mass flow rate of water so that the collimator can adapt to operational environments with higher heat fluxes and long pulses in the future. These research methods also provide a reference for the future design of components under high-energy and long-pulse operational conditions.

• Nuclear Engineering and Technology
• /
• v.55 no.7
• /
• pp.2621-2627
• /
• 2023

In Experimental Advanced Superconducting Tokamak (EAST) experiments, to achieve long pulse and high-power ICRH system operation, a new kind of ICRH antenna has been designed. One of the most critical factors in limiting the operation of long pulse and high power is the intense heat load in the front face of the ICRH antenna, especially the Faraday Screen (FS). Therefore, the cooling channels of FS need to be designed. According to thermal-hydraulic analysis, the FS tubes are divided into several groups to achieve more excellent water cooling capability. The number of series and parallel tubes in one group is chosen as six. This antenna went into service in the spring of 2021, and it is delightful that the temperature distribution of the FS tube is below 400 ℃ in 14.5 s and 1.8 MW ICRH system operation. However, the active water-cooling design was not carried out on the upper and lower plates of FS, which led to severe ablations on that region under long pulse and high power operation, and the temperature is up to 800. Therefore, the upper and lower side plates of the FS were designed with water cooling based on thermal-hydraulic analysis. During the 2022 winter experiments, the temperature of ICRH antenna FS was lower than 400 in the pulse of 200s and the power of 1 MW operation.

• Korean Chemical Engineering Research
• /
• v.50 no.4
• /
• pp.595-603
• /
• 2012

International Thermonuclear Experimental Reactor (ITER) will be constructed in 2019 according to the JIA (Joint Implementation Agreement) of 7 countries. The ITER fusion fuel cycle consists of fusion vacuum vessel, tritium plant and fuelling system. The tritium plant provides the functions of storage, delivery, separation, removal and recovery of the deuterium and tritium used as fusion fuels for the ITER. The tritium plant systems supply deuterium and tritium from external sources and treat all tritiated fluids from ITER operation through Storage and Delivery System (SDS), Tokamak Exhaust Processing (TEP), Isotope Separation System (ISS), Water Detritiation System & Atmosphere Detritiation System (WDS & ADS) and Analysis System (ANS). In this paper, the functions and design requirements of the major systems in the tritium plant and the status of R&D are described. Korean party is developing the SDS for ITER tritium plant and partially attaining the WDS technology through the construction and operation experience of the Wolsong Tritium Removal Facility (WTRF). Now it is expected that researchers in other fields such as chemical engineering take part in the development of upcoming technologies for ISS and TEP.