Y.S. Bae

Design and construction of the KSTAR tokamak

The extensive design effort for KSTAR has been focused on two major aspects of the KSTAR project mission - steady-state-operation capability and advanced tokamak physics. The steady state aspect of the mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long pulse current drive and heating systems. The advanced tokamak aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double null divertor and passive stabilizers, internal control coils and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, the vacuum vessel, plasma facing components and power supplies. The new KSTAR experimental facility with cryogenic system and deionized water cooling and main power systems has been designed, and the construction work is under way for completion in 2004.

A lower hybrid current drive system for ITER

A 20 MW/5 GHz lower hybrid current drive (LHCD) system was initially due to be commissioned and used for the second mission of ITER, i.e. the Q = 5 steady state target. Though not part of the currently planned procurement phase, it is now under consideration for an earlier delivery. In this paper, both physics and technology conceptual designs are reviewed. Furthermore, an appropriate work plan is also developed. This work plan for design, R&D, procurement and installation of a 20 MW LHCD system on ITER follows the ITER Scientific and Technical Advisory Committee (STAC) T13-05 task instructions. It gives more details on the various scientific and technical implications of the system, without presuming on any work or procurement sharing amongst the possible ITER partners(b). This document does not commit the Institutions or Domestic Agencies of the various authors in that respect.

Validation of the ITER-relevant passive-active-multijunction LHCD launcher on long pulses in Tore Supra

A new ITER-relevant lower hybrid current drive (LHCD) launcher, based on the passive-active-multijunction (PAM) concept, was brought into operation on the Tore Supra tokamak in autumn 2009. The PAM launcher concept was designed in view of ITER to allow efficient cooling of the waveguides, as required for long pulse operation. In addition, it offers low power reflection close to the cut-off density, which is very attractive for ITER, where the large distance between the plasma and the wall may bring the density in front of the launcher to low values. The first experimental campaign on Tore Supra has shown extremely encouraging results in terms of reflected power level and power handling. Power reflection coefficient <2% is obtained at low density in front of the launcher, i.e. close to the cut-off density, and very good agreement between the experimental results and the coupling code predictions is obtained. Long pulse operation at ITER-relevant power density has been demonstrated. The maximum power and energy reached so far is 2.7 MW during 78 s, corresponding to a power density of 25 MW m −2 , i.e. its design value at f = 3.7 GHz. In addition, 2.7 MW has been coupled at a plasma–launcher distance of 10 cm, with a power reflection coefficient <2%. Finally, full non-inductive discharges have been sustained for 50 s with the PAM.

ECH pre-ionization and assisted startup in the fully superconducting KSTAR tokamak using second harmonic

This letter reports on the successful demonstration of the second harmonic electron cyclotron heating (ECH)-assisted startup in the first plasma experiments recently completed in the fully superconducting Korea Superconducting Tokamak Advanced Research (KSTAR) device whose major and minor radii are 1.8 m and 0.5 m, respectively. For the second harmonic ECH-assisted startup, an 84 GHz EC wave at 0.35 MW was launched before the onset of the toroidal electric field of the Ohmic system. And it was observed that this was sufficient to achieve breakdown in the ECH pre-ionization phase, allow burn-through and sustain the plasma during the current ramp with a low loop voltage of 2.0 V and a corresponding toroidal electric field of 0.24 V m−1at the innermost vacuum vessel wall (R = 1.3 m). This is a lower value than 0.3 Vm−1 which is the maximum electric field in ITER. Due to the limited volt-seconds and the loop voltage of the Ohmic power system, the extended pulse duration of the ECH power up to 180 ms allowed the plasma current to rise up to more than 100 kA with a ramp-up rate of 0.8 MA s−1.

KSTAR equilibrium operating space and projected stabilization at high normalized beta

Along with an expanded evaluation of the equilibrium operating space of the Korea Superconducting Tokamak Advanced Research, KSTAR, experimental equilibria of the most recent plasma discharges were reconstructed using the EFIT code. In near-circular plasmas created in 2009, equilibria reached a stored energy of 54kJ with a maximum plasma current of 0.34MA. Highly shaped plasmas with near double-null configuration in 2010 achieved H-mode with clear edge localized mode (ELM) activity, and transiently reached a stored energy of up to 257kJ, elongation of 1.96 and normalized beta of 1.3. The plasma current reached 0.7MA. Projecting active and passive stabilization of global MHD instabilities for operation above the ideal no-wall beta limit using the designed control hardware was also considered. Kinetic modification of the ideal MHD n = 1 stability criterion was computed by the MISK code on KSTAR theoretical equilibria with a plasma current of 2MA, internal inductance of 0.7 and normalizedbetaof4.0withsimpledensity,temperatureandrotationprofiles. Thesteepedgepressuregradientofthis equilibrium resulted in the need for significant plasma toroidal rotation to allow thermal particle kinetic resonances to stabilize the resistive wall mode (RWM). The impact of various materials and electrical connections of the passive stabilizing plates on RWM growth rates was analysed, and copper plates reduced the RWM passive growth rate by a factor of 15 compared with stainless steel plates at a normalized beta of 4.4. Computations of active RWM control using the VALEN code showed that the n = 1 mode can be stabilized at normalized beta near the ideal wall limit via control fields produced by the midplane in-vessel control coils (IVCCs) with as low as 0.83kW control power using ideal control system assumptions. The ELM mitigation potential of the IVCC, examined by evaluating the vacuum island overlap created by resonant magnetic perturbations, was analysed using the TRIP3D code. Using a combinationofallIVCCswithdominant n = 2fieldandupper/lowercoilsinanevenparityconfiguration,aChirikov parameter near unity at normalized poloidal flux 0.83, an empirically determined condition for ELM mitigation in DIII-D, was generated in theoretical high-beta equilibria. Chirikov profile optimization was addressed in terms of coil parity and safety factor profile. (Some figures in this article are in colour only in the electronic version)

Characteristics of the First H-mode Discharges in KSTAR

Typical ELMy H-mode discharges have been obtained in the KSTAR tokamak with the combined auxiliary heating of neutral beam injection (NBI) and electron cyclotron resonant heating (ECRH). The minimum external heating power required for the L?H transition is about 0.9?MW for a line-averaged density of ~2.0 ? 1019?m?3. There is a clear indication of the increase in the L?H threshold power with decreasing density for densities lower than ~2 ? 1019?m?3. The L?H transitions typically occurred shortly after the beginning of plasma current flattop (Ip = 0.6?MA) period and after the fast shaping to a highly elongated double-null divertor configuration. The maximum heating power available was marginal for the L?H transition, which is also implied by the relatively slow transition time (>10?ms) and the synchronization of the transition with large sawtooth crashes. The initial analysis of thermal energy confinement time (?E) indicates that ?E is higher than the prediction of multi-machine scaling laws by 10?20%. A clear increase in electron and ion temperature in the pedestal is observed in the H-mode phase but the core temperature does not change significantly. On the other hand, the toroidal rotation velocity increased over the whole radial range in the H-mode phase. The measured ELM frequency was around 10?30?Hz for the large ELM bursts and 50?100?Hz for the smaller ones. In addition, very small and high frequency (200?300?Hz) ELMs appeared between large ELM spikes when the ECRH is added to the NBI-heated H-mode plasmas. The drop of total stored energy during a large ELM is up to 5% in most cases.

Status of KSTAR Electron Cyclotron Heating System

An 84-GHz, 500-kW electron cyclotron (EC) heating (ECH) system is under installation for ECH-assisted start-up in the Korea Superconducting Tokamak Advanced Research (KSTAR) facility. An 84-GHz, 500-kW gyrotron, and 1.5-MVA power supply system have been installed at KSTAR, and the initial test of the gyrotron has been carried out with a short-pulse condition of 20 μs and maximum beam parameters of 80 kV and 25 A that generate an output radio-frequency (rf) power of 500 kW. The planned 2-s-long operation with 500-kW rf output power is beginning with a long-pulse test of the gyrotron power supply. The launcher system was fabricated in collaboration with Princeton Plasma Physics Laboratory. It will inject 500-kW rf power into the KSTAR plasma with a highly flexible steering mirror system, allowing toroidal and poloidal beam deposition scans. KSTAR will employ 170-GHz EC current drive (CD) in ITER-relevant experiments such as the suppression of the neoclassical tearing modes and the creation of an electron internal transport barrier. The Japan Atomic Energy Agency will provide a 170-GHz, 1-MW gyrotron on loan in 2008 in accordance with a Korea-Japan fusion collaboration agreement, and it will be used for the 170-GHz, 1-MW ECCD system in 2010. This paper describes the current status of the installation and initial conditioning tests of the 84-GHz gyrotron system as well as the development plan of the 170-GHz ECH and CD system. Also, this paper discusses the CD efficiency and the steering range of the second-harmonic X-mode injection at 170 GHz and 5 MW from an equatorial launcher.

Recent progress on lower hybrid current drive and implications for ITER

The sustainment of steady-state plasmas in tokamaks requires efficient current drive systems. Lower hybrid current drive is currently the most efficient method to generate a continuous additional off-axis toroidal plasma current and to reduce the poloidal flux consumption during the plasma current ramp-up phase. The operation of the Tore Supra ITER-like lower hybrid (LH) launcher has demonstrated the capability to couple LH power at ITER-like power densities with very low reflected power during long pulses. In addition, the installation of eight 700 kW/CW klystrons at the LH transmitter has allowed increasing the total LH power in long-pulse scenarios. However, in order to achieve pure stationary LH-sustained plasmas, some R&D is needed to increase the reliability of all the systems and codes, from radio-frequency (RF) sources to plasma scenario prediction. The CEA/IRFM is addressing some of these issues by leading a R&D programme towards an ITER LH system and by the validation of an integrated LH modelling suite of codes. In 2011, the RF design of a mode converter was validated at a low power. A 500 kW/5 s RF window is currently under manufacture and will be tested at a high power in 2012 in collaboration with the National Fusion Research Institute. All of this work aims to reduce the operational risks associated with the ITER steady-state operations.

Investigation of MHD instabilities and control in KSTAR preparing for high beta operation

Initial H-mode operation of the Korea Superconducting Tokamak Advanced Research (KSTAR) is expanded to higher normalized beta and lower plasma internal inductance moving towards design target operation. As a key supporting device for ITER, an important goal for KSTAR is to produce physics understanding of MHD instabilities at long pulse with steady-state profiles, at high normalized beta, and over a wide range of plasma rotation profiles. An advance from initial plasma operation is a significant increase in plasma stored energy and normalized beta, with Wtot = 340 kJ, βN = 1.9, which is 75% of the level required to reach the computed ideal n = 1 no-wall stability limit. The internal inductance was lowered to 0.9 at sustained H-mode duration up to 5 s. In ohmically heated plasmas, the plasma current reached 1 MA with prolonged pulse length up to 12 s. Rotating MHD modes are observed in the device with perturbations having tearing rather than ideal parity. Modes with m/n = 3/2 are triggered during the H-mode phase but are relatively weak and do not substantially reduce Wtot. In contrast, 2/1 modes to date only appear when the plasma rotation profiles are lowered after H–L back-transition. Subsequent 2/1 mode locking creates a repetitive collapse of βN by more than 50%. Onset behaviour suggests the 3/2 mode is close to being neoclassically unstable. A correlation between the 2/1 mode amplitude and local rotation shear from an x-ray imaging crystal spectrometer suggests that the rotation shear at the mode rational surface is stabilizing. As a method to access the ITER-relevant low plasma rotation regime, plasma rotation alteration by n = 1, 2 applied fields and associated neoclassical toroidal viscosity (NTV) induced torque is presently investigated. The net rotation profile change measured by a charge exchange recombination diagnostic with proper compensation of plasma boundary movement shows initial evidence of non-resonant rotation damping by the n = 1, 2 applied field configurations. The result addresses perspective on access to low rotation regimes for MHD instability studies applicable to ITER. Computation of active RWM control using the VALEN-3D code examines control performance using midplane locked mode detection sensors. The LM sensors are found to be strongly affected by mode and control coil-induced vessel current, and consequently lead to limited control performance theoretically.

Results of Beam Extraction Performance for the KSTAR Neutral Beam Injector

The first neutral beam injector (NBI-1) has been developed for the Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak. The first long pulse ion source (LPIS-1) has been installed on the NBI-1 for an auxiliary heating and current drive of KSTAR plasmas. The performance of ion and neutral beam extractions in the LPIS-1 was investigated initially on the KSTAR NBI-1 system, prior to the neutral beam injection into the main plasmas. The ion source consists of a magnetic bucket plasma generator with multipole cusp fields and a set of prototype tetrode accelerators with circular apertures. The inner volume of the plasma generator and accelerator column in the LPIS-1 is approximately 123 L. Design requirements for the ion source were a 120 kV/65 A deuterium beam and a 300 s pulse length. The extraction of ion beams was initiated by the formation of arc plasmas in the LPIS-1, called the arc-beam extraction method. A stable ion beam extraction of the LPIS-1 was achieved up to 85 kV/32 A for a 5 s pulse length and 80 kV/25 A for a 14 s pulse length. An optimum beam perveance of 1.15 µperv was observed at an acceleration voltage of 60 kV. Neutralization efficiency was measured by a water-flow calorimetry (WFC) method using a calorimeter and the operation of a bending magnet. The full-energy species of ion beams were detected by using the diagnostic method of optical multichannel analyzer (OMA). An arc efficiency of the LPIS was 0.6–1.1 A/kW depending on the operating conditions of arc discharge. A neutral beam power of ~1.0 MW must be sufficiently injected into the KSTAR plasmas from the LPIS-1 at a beam energy of 80 keV.