Present Status of the KSTAR Superconducting Magnet System Development

The mission of Korea Superconducting Tokamak Advanced Research (KSTAR) project is to develop an advanced steady-state superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Because one of the KSTAR mission is to achieve a steady-state operation, the use of superconducting coils is an obvious choice for the magnet system. The KSTAR superconducting magnet system consists of 16 Toroidal Field (TF) coils and 14 Poloidal Field (PF) coils. Internally-cooled Cable-In-Conduit Conductors (CICC) are put into use in both the TF and PF coil systems. The TF coil system provides a field of 3.5 T at the plasma center and the PF coil system is able to provide a flux swing of 17 V-sec. 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 coil system for background magnetic-field generation , the construction of a large-scale superconducting magnet and CICC test facility. TF and PF coils are in the stage of fabrication to pave the way for the scheduled completion of KSTAR by the end of 2006.     

Fabrication of the KSTAR vacuum vessel and ports

The construction of the steady-state-capable superconducting KSTAR tokamak is in close proximity to the finalization. As one of the main components of the KSTAR device, the vacuum vessel is designed and manufactured during the construction period. The KSTAR vacuum vessel is composed of two large sectors forming the 337.5° of a full torus, and the remaining 22.5° section consisting of 24 small pieces. The large two sectors were welded at the site, and the 22.5° space was used for toroidal field coil assembly. The remaining 22.5° section of the vacuum vessel was assembled after 16 toroidal field coils assembly. The total 72 penetration ports were used to connect the vacuum vessel body and the cryostat. The major fabrication activity started in January 2003 after the finalization of fabrication design. The final components and structures were warehoused in June 2004 and site assembly is finished in 2007. Details of analysis, shop fabrication, and inspection results of the vacuum vessel including ports are summarized in the present work.     

MHD stability analysis for advanced tokamak modes in the KSTAR device

The Korea Superconducting Tokamak Advanced Research (KSTAR) device aims to demonstrating the steady-state operation of high-performance advanced tokamak (AT) modes. In order to meet this research goal it is critical to have a good magnetohydrodynamic (MHD) stability, so that KSTAR adopted a strong plasma shape and a conducting wall close to plasma for such stability. An early calculation during the KSTAR design phase had shown that a target AT mode stable up to β_N above 5 can be then obtained. A recent work by Katsuro-Hopkins et al. [O. Katsuro-Hopkins, S.A. Sabbagh, J.M. Bialek, H.K. Park, J.G. Bak, J. Chung, et al., Equilibrium and global MHD stability study of KSTAR high beta plasmas under passive and active mode control, Nucl. Fusion 50(2010) 025019] showed, however, that the maximum β_N value can be substantially lower than 5, unlike the earlier result. In this work, we present a more detailed study on the MHD stability limit of the KSTAR target AT mode and try to clarify the discrepancy observed in the previous two works. It is shown that in the reverse-shear plasma the target mode with β_N above 5 can be obtained if the pressure profile is relatively peaked, but the maximum β_N value is substantially reduced below 5 if the pressure profile becomes broader. This result suggests the importance of a proper control of the pressure profile to get the high-beta AT mode in KSTAR.     

The operational results of the KSTAR PF cryo-circuit

The Koreas Superconducting Tokamak Advanced Research (KSTAR) PF cryo-circuit is designed for cooling the fourteen superconducting magnets (Nb3Sn, NbTi) and structures. Those are cooled down by the supercritical helium (4.5 K and 5.5 bar) of a forced flow (pressure gradient: 2 bar) in order to maintain the supercritical state of the helium. To supply a large amount of supercritical helium (>370 g/s), a circulator was inserted into the PF cryo-circuit. The compressed supercritical helium is distributed to five helium manifolds with cryogenic valves and supplied to each PF magnet. While the PF magnets had been operating, the mass flow rate reduced and the pressure head of the circulator was fluctuated depending on the PF magnet operation scenario. These phenomena could damage the circulator and could stop it during operation. Therefore, by-pass valve, which is parallel with in-line valve and is connected with inlet and outlet of the magnet, was opened in order to reduce of the circulator's pressure head. In this paper, we focused on the helium behavior of the superconducting magnet when the by-pass valve was opened in order to release the pressure head of the circulator and the results will be presented.     

Design and test of the KSTAR vacuum pumping system

The Korea Superconductor Tokamak Advanced Research (KSTAR) device is a tokamak mainly composed of a vacuum vessel, superconducting magnets, and cryostat. The internal volume of the vacuum vessel is about 110 m³ with a target pressure of 1 × 10⁻⁶ Pa, while the volume of the cryostat is 450 m³ with a target pressure of 5 × 10⁻³ Pa. To attain these target pressures, two identical vacuum pumping systems consisting of dry pumps, mechanical booster pumps, turbo-molecular pumps, and cryopumps were installed. The control system of the vacuum pumping systems was built using the experimental physics and industrial control system (EPICS), which has various merits such as easy access, convenient extension and flexible integration. The pump-down test of the pumping ducts was successfully executed under the control of the EPICS system.     

Design and operation results of nitrogen gas baking system for KSTAR plasma facing components

A baking system for the Korea Superconducting Tokamak Advanced Research (KSTAR) plasma facing components (PFCs) is designed and operated to achieve vacuum pressure below 5 × 10^?7 mbar in vacuum vessel with removing impurities. The purpose of this research is to prevent the fracture of PFC because of thermal stress during baking the PFC, and to accomplish stable operation of the baking system with the minimum life cycle cost. The uniformity of PFC temperature in each sector was investigated, when the supply gas temperature was varied by 5 °C per hour using a heater and the three-way valve at the outlet of a compressor. The alternative of the pipe expansion owing to hot gas and the cage configuration of the three-way valve were also studied. During the fourth campaign of the KSTAR in 2011, nitrogen gas temperature rose up to 300 °C, PFC temperature reached at 250 °C, the temperature difference among PFCs was maintained at below 8.3 °C, and vacuum pressure of up to 7.24 × 10^?8 mbar was achieved inside the vacuum vessel.     

Development status of KSTAR 5 GHz LHCD system

A steady-state lower hybrid current drive (LHCD) system is under development for advanced tokamak experiments of the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The KSTAR 5 GHz steady-state LHCD system is being designed to couple an input power of 2 MW for 300 s generated by four 5 GHz klystrons. For the development of this system, there are two critical issues. One is the development of a 5 GHz CW klystron for the RF source of the system. The other is the design of a steady-state LH launcher with active water cooling. In this paper, the current status of the development and design for the KSTAR steady-state LHCD system is described. For the LHCD system, aiming at a basic experimental study of 5 GHz LH wave propagation and operational experience with an LHCD system, the installation of an initial LHCD system with a capacity of 0.5 MW for 2 s is scheduled in 2010 using a 5 GHz prototype klystron and an un-cooled 1 MW launcher. The design and progress for the initial LHCD system are also presented.     

Static and Dynamic Mechanical Analyses for the Vacuum Vessel of EAST Superconducting Tokamak Device

EAST （experimental advanced superconducting tokamak） is an advanced steadystate plasma physics experimental device, which is being constructed as the Chinese National Nuclear Fusion Research Project. During the plasma operation the vacuum vessel as one of the key component will withstand the electromagnetic force due to the plasma disruption, the Halo current and the toroidal field coil quench, the pressure of boride water and the thermal load due to 250℃ baking by pressurized nitrogen gas. In this paper a report of the static and dynamic mechanical analyses of the vacuum vessel is made. Firstly the applied loads on the vacuum vessel were given and the static stress distribution under the gravitational loads, the pressure loads, the electromagnetic loads and thermal loads were investigated. Then a series of primary dynamic, buckling and fatigue life analyses were performed to predict the structure＇s dynamic behavior. A seismic analysis was also conducted.     

First Engineering Commissioning of EAST Tokamak

Experimental Advanced Superconducting Tokamak （EAST） is the first fully superconducting tokamak. The first commissioning started on Feb. 1st of 2006 and finished on March 30TM of 2006 at the Institute of Plasma Physics, Chinese Academy of Sciences. It consists of leakage testing at both room temperature and low temperature, pumping down, cooling down all coils, current leads, bus bar and the thermal shielding, exciting all the coils, measuring magnetic configuration and warming up the magnets. The electromagnetic, thermal hydraulic and mechanical performance of EAST Toroidal Field （TF） and Poloidal Field （PF） magnets have also been tested. All sub-systems, including pumping system, cryogenic system, PF＆ TF power supply systems, magnet instrumentation system, quench detection and protection system, water cooling system, data acquisition system, main control system, plasma control system （PCS）, interlock and safety system have been successfully tested.     

Achievement of 1.9 MW for 300 s during the commissioning of the KSTAR ICRF transmitter

KSTAR (Korea Superconducting Tokamak Advanced Research) is a national tokamak aiming at the high beta operation based on AT (Advanced Tokamak) scenarios in Korea and ICRF (Ion Cyclotron Ranges of Frequency) is one of the essential heating and current drive tools to achieve this goal. The ICRF heating and current drive scenario requires 4 units of 2 MW transmitters with a frequency range from 25 to 60 MHz. The first KSTAR transmitter is a modified FMIT (Fusion Material Irradiation Test) transmitter consisting of four amplifier stages. An amplitude-modulated 1 mW frequency source drives a 500 W solid state wideband amplifier, which in turn drives three tuned triode/tetrode amplifier stages. The tube employed in the final power amplifier is a 4CM2500KG tetrode fabricated by CPI (Communications & Power Industries). After the fabrication of the cavity and power supply was completed in 2004, several failures of the tube during a factory and a site acceptance test occurred before eventually achieving 1.9 MW for 300 s at 33 MHz in 2007. The electrical efficiency of the FPA (Final Power Amplifier) is about 70%. Although this is a very encouraging result for the development of an ICRF transmitter for ITER (International Thermonuclear Experimental Reactor), continued efforts for a reliable operation are required to achieve the final goals of the KSTAR and ITER ICRF system.     

Water-cooling effects on the high-voltage and long-pulse capabilities of the KSTAR ICRF antenna

A new ICRF antenna originating from the prototype antenna was constructed for the KSTAR tokamak in 2002. The performance of the antenna was experimentally estimated at the RF test stand without a plasma. Recently three series of RF tests were performed at a frequency of 30 MHz; without any cooling, with a water-cooling for only the antenna, and with a water-cooling of the antenna and the transmission line connected to the antenna. In the tests, a half of the current strap was connected to a RF source via a matching circuit with the other half one connected to an open terminated coaxial line, and the other three straps were shorted at the input ports. During the RF pulse, the temperatures at several positions of the antenna cavity wall were measured by embedded thermocouples and the temperature profile of the front face of the antenna was measured by an IR camera. The line voltage, forward and reflected powers, and the RFTC pressure were also measured. The water-cooled antenna showed several enhanced performances in a comparison with the non-cooled case, and the standoff voltage was significantly increased. By utilizing a water-cooling of the antenna and the transmission line, we achieved a standoff voltage of 41.3 kVp for a pulse length of 300 s, and we could extend the pulse length up to 600 s at a maximum voltage of 35.0 kVp without encountering any problems, which considerably exceeds the design requirements.     

Commissioning the ICRF system and an ICRF assisted discharge cleaning at the KSTAR

Korean superconducting tokamak advanced research (KSTAR) is a national superconducting tokamak with the aim of a high beta operation based on advanced tokamak (AT) scenarios, and an ion cyclotron ranges of frequency (ICRF) heating is one of the essential tools to achieve this goal. The fabrication and high voltage (HV) test of the antenna and the matching system were finished in 2006 and the installation of the antenna, matching system and the transmitter at the KSTAR site was completed in 2007. Antenna conditioning was carried out to improve the HV holding condition of the antenna installed on the KSTAR and to check on the electro-magnetic (EM) interference with other equipments such as the superconducting magnet monitoring system and other machine and/or plasma diagnostic systems. The first KSTAR tokamak experimental campaign started by a vacuum pumping, a cryostat cooling and an ICRF system contributed to the successful tokamak shots through an ICRF assisted discharge cleaning of the vacuum vessel. In this paper, the installation processes of the ICRF system (with an emphasis on the quality assurance procedures of KSTAR), as well as the results from the first RF discharge experiment for the discharge cleaning and FWEH (fast wave electron heating) experiment for the KSTAR 1st experimental campaign are outlined.     

Completed installations and the individual commissioning of the KSTAR MG system

Peak power of 200 MVA is required in order to achieve the goal within a long pulse scenario for the final operation of the Korean Superconducting Tokamak Advanced Research (KSTAR). The available grid power is only 100 MVA at the National Fusion Research Institute (NFRI) site. Motor generator (MG) was considered as a method of resolving such problems. The design of the KSTAR MG system was completed in July 2010 and individual devices were produced by relevant manufacturers. The installation of individual devices was completed in December 2012. Specifically, the stator and rotor were assembled at the site due to their large size and weight. The bearings, variable voltage variable frequency (VVVF) and excitation systems were transported and installed on site after being manufactured externally.The building used for MG installation was built in 2011. With the building designed for ease of installation, an overhead crane was designed to allow access to the loading bay.In this paper, we discuss the installation of the MG system and the construction of the building suitable for installation of individual devices. In addition, performance on the test results of individual devices is also discussed.     

Real-time plasma boundary reconstruction in the KSTAR tokamak using finite element method

A study is carried out on the real-time plasma shape identification in the KSTAR device. An improved form of the finite current element (FCE) method is utilized in this study. Results are shown that the plasma boundary can be reproduced in 7 mm accuracy for any plasma configuration in ideal cases without invoking measurement errors. A design guideline for magnetic diagnostics (MD) is established when the measurement signals are subject to Gaussian noise. It is found that the measurement errors in poloidal field (PF) coil currents have substantial influence on the determination of the plasma shape.     

Numerical prediction of pellet injection effects on core plasma fueling in the KSTAR tokamak

Since pellet injection into tokamak plasmas has been found to be an effective method for fueling and profile modification of core plasmas in tokamak experiments, a hypothetical injection of deutrium pellets into the KSTAR tokamak is numerically simulated in this work to investigate its influences on the fueling and transport of the core plasma depending on pellet parameters. A neutral gas shielding model and a pellet drift displacement model are used to describe the ablation and mass deposition from pellets on core plasma profiles. These models are coupled with a 1.5-dimensional (1.5D) core transport code to calculate the plasma density and temperature profiles responding to pellets injected into the target plasma. The simulation results indicate that a HFS (high field side) injection achieves more effective fueling due to a deeper pellet penetration into the core plasma, compared with a LFS (low field side) injection. The plasma density is found to increase during sequential pellet injections from both HFS and LFS, but the HFS case shows better fueling performance owing to a drift of the pellet ablatant in the major radius direction resulting in the deeper pellet penetration. Increasing the size and injection velocity of the pellet contributes to enhance the fueling efficiency. However, raising the power of neutral beam injection heating reduces the fueling efficiency because the pellet mass deposition is shifted toward the edge region in high temperature plasmas. It is concluded that the pellet size and injection direction among pellet and plasma parameters have the most dominant effects on fueling performance while the pellet velocity and heating power have relatively small influences on fueling.     

Progress and improvement of KSTAR plasma control using model-based control simulators

Superconducting tokamaks like KSTAR, EAST and ITER need elaborate magnetic controls mainly due to either the demanding experiment schedule or tighter hardware limitations caused by the superconducting coils. In order to reduce the operation runtime requirements, two types of plasma simulators for the KSTAR plasma control system (PCS) have been developed for improving axisymmetric magnetic controls. The first one is an open-loop type, which can reproduce the control done in an old shot by loading the corresponding diagnostics data and PCS setup. The other one, a closed-loop simulator based on a linear nonrigid plasma model, is designed to simulate dynamic responses of the plasma equilibrium and plasma current (I_p) due to changes of the axisymmetric poloidal field (PF) coil currents, poloidal beta, and internal inductance. The closed-loop simulator is the one that actually can test and enable alteration of the feedback control setup for the next shot. The simulators have been used routinely in 2012 plasma campaign, and the experimental performances of the axisymmetric shape control algorithm are enhanced. Quality of the real-time EFIT has been enhanced by utilizations of the open-loop type. Using the closed-loop type, the decoupling scheme of the plasma current control and axisymmetric shape controls are verified through both the simulations and experiments. By combining with the relay feedback tuning algorithm, the improved controls helped to maintain the shape suitable for longer H-mode (10–16 s) with the number of required commissioning shots largely reduced.     

Optimization of an in-vessel visible inspection system for a long-pulse operation in KSTAR

To monitor the global formation of shaped plasmas, motion, and damage to the internal structures of the vacuum vessel, an in-vessel visible inspection system has been developed and operated on the Korean superconducting tokamak advanced research (KSTAR) device. The system contributed much to research progress on KSTAR such as the plasma start-up, plasma wall interactions, edge-localized modes, and disruptions. Moreover the need to perform inspections became important with high plasma power operation because of the increased frequency of first wall damage following off-normal events. Therefore the system is being improved from its original concept, and its final goal is operation during steady-state operation of the tokamak. The system consists of three fast visible cameras and two light-emitting diode illuminators. They are designed to be controlled fully from the control room to provide inspection capability at any time during hostile operating conditions. In this paper, we describe the upgrade of the system and recent results of the visible inspection system with the images of the KSTAR discharges for the last four years. Finally, we discuss the technical issues for a long pulse steady-state operation.     

Time-dependent simulations of feedback stabilization of neoclassical tearing modes in KSTAR plasmas

A simulation is performed for feedback stabilization of neoclassical tearing mode (NTM) by electron cyclotron current drive (ECCD) for KSTAR in preparation for experiments. An integrated numerical system is constructed by coupling plasma transport, NTM stability, and heating and current drive modules and applied to a KSTAR plasma by assuming similar experimental conditions as ASDEX Upgrade to predict NTM behaviors in KSTAR. System identification is made with database produced by predictive simulations with this integrated numerical system so that three plasma response models are extracted which describe the relation between the EC poloidal launcher angle and the island width in KSTAR. Among them, the P1DI model exhibiting the highest fit accuracy is selected for designing a feedback controller based on the classical Proportional–Integral–Derivative (PID) concept. The controller is coupled with the integrated numerical system and applied to a simulation of NTM stabilization. It is observed that the controller can search and fully stabilize the mode even though the poloidal launch angle is misaligned with the island initially.     

Performance of a New Ion Source for KSTAR Tokamak Plasma Heating

In the experimental campaign of 2010 and 2011 on KSTAR, the NBI-1 system was equipped with one prototype ion source and operated successfully, providing a neutral beam power of 0.7-1.6 MW to the tokamak plasma. The new ion source planned for the 2012 KSTAR campaign had a much more advanced performance compared with the previous one. The target performance of the new ion source was to provide a neutral deuterium beam of 2 MW to the tokamak plasma. The ion source was newly designed, fabricated, and assembled in 2011. The new ion source was then conditioned up to 64 A/100 keV over a 2-hour beam extraction and performance tested at the NB test stand （NBTS） at the Korea Atomic Energy Research Institute （KAERI） in 2012. The measured optimum perveance at which the beam divergence is a minimum was about 2.5μP, and the minimum beam divergent angle was under 1.0° at 60 keV. These results indicate that the 2.0 MW neutral beam power at 100 keV required for the heating of plasma in KSTAR can be delivered by the installation of the new ion source in the KSTAR NBI-1 system.