EAST ion cyclotron resonance heating system for long pulse operation

Radio frequency (RF) power in the ion cyclotron range of frequencies (ICRF) is one of the primary auxiliary heating techniques for Experimental Advanced Superconducting Tokamak (EAST). The ICRF system for EAST has been developed to support long-pulse high-β advanced tokamak fusion physics experiments. The ICRF system is capable of delivering 12 MW 1000-s RF power to the plasma through two antennas. The phasing between current straps of the antennas can be adjusted to optimize the RF power spectrum. The main technical features of the ICRF system are described. Each of the 8 ICRF transmitters has been successfully tested to 1.5 MW for a wide range of frequency (25–70 MHz) on a dummy load. Part of the ICRF system was in operation during the EAST 2012 spring experimental campaign and a maximum power of 800 kW (at 27 MHz) lasting for 30 s has been coupled for long pulse H mode operation.     

Parametric study of one triplet of the ITER ICRH antenna by numerical modeling

Each of the two ITER ICRF antennas consists of a close-packed array of 24 straps arranged in a 6 poloidal by 4 toroidal array. Three poloidally adjacent straps (a “triplet” of straps) are fed together through a 4-port junction from one 20 Ohm feeding line. The complete array has to radiate 20 MW of RF power over a frequency range of 40 MHz to 55 MHz and for different toroidal phasings. The RF optimization of the antenna has been performed numerically on one triplet of straps (1/8th of the antenna) [1], [2]. In parallel a number of reduced-scale mock-ups of one triplet of the ITER ICRH antenna were constructed in order to validate the results of the numerical optimization [1], [3].The aim of this work is primarily to benchmark the CST MWS^® [4] numerical modeling against numerous measurements done on the mock-up of the 2007 design. Moreover MWS calculates the 3D distribution of the currents and of the fields of the triplet. Hence it gives the possibility to check the fields and current distributions resulting from the optimisation study of the ITER ICRH antenna triplet done by changing geometrical parameters of the straps and antenna box of the mock-up of 2007 design [1], [2], [3]. The considered parameters are: strap width, antenna box depth and vertical septum recess with respect to the front of the current strap. The impact of the presence of the Faraday screen is also evaluated.Excellent agreement between modeled and measured S parameters is obtained. Analysis of the fields and currents distributions on the straps is reported. Excellent current balance is confirmed.     

Current Status of ICRF Heating Experiments on EAST

Radio frequency (RF) heating in the ion cyclotron range of frequencies (ICRF) is one of the primary auxiliary heating methods for EAST. The ICRF system provides 6 MW power in primary phase and will be capable of 10 MW later. Three 1.5 MW ICRF systems in a frequency range of 25 to 70 MHz have already been in operation. The ICRF heating launchers are designed to have two current straps with each driven by a RF power source of 1.5 MW. In this paper a brief introduction of the ICRF heating system capability in EAST and the preliminary results in EAST are presented.     

Recent plasma control progress on EAST

In recent 2 years, various algorithms to control plasma shape, current and density have been implemented or improved for EAST tokamak. These plasma control performances have been verified by either simulated or actual experimental operation, and thus plasma control basis has been established for the long pulse operation and high performance H-mode plasma operation with low hybrid wave (LHW) and ion cyclotron resonance frequency (ICRF) heating. Startup simulation has been done by using TOKSYS code for the plasma breakdown in either 3.1 Wb or 4.5 Wb initial poloidal flux state and the scenarios proved to be robust and used for routine operation. Various shape configurations have been well feedback controlled by using ISOFLUX limited, double-null or single null algorithms based on RTEFIT equilibrium reconstruction. For the long pulse operation, strike point control and magnetics drift compensation have been implemented in the plasma control system (PCS). To improve the operation safety and efficiency, the verification of magnetic diagnostics before plasma breakdown has been demonstrated adequate to prevent a discharge in case of key sensor failure.     

EASTICRF天线电流带电磁分析

为了实现EAST托卡马克1000s以上的稳态先进模式运行的最终物理目标,两电流带双环共振(RDL)离子回旋共振（ICRF）天线被选择用来加热,电流带是ICRF天线关键部件,它通过近场区的耦合把能量传输到等离子体中。本文通过有限元方法对电流带在等离子体破裂和等离子体垂直位移事件两种工况下进行了电磁计算,给出了电流带感应电流密度大小分布情况、磁感应强度大小分布情况以及电流带所受的电磁力。利用电流带所受的电磁力作为载荷对电流带进行了结构分析,分析结果为验证电流带结构的可行性提供理论依据,分析方法对未来更高功率的ICRF天线电流带进行电磁分析具有一定的借鉴价值。     

Commissioning of the ITER-like ICRF antenna for JET

The new JET ion cyclotron resonance frequency (ICRF) ITER-like antenna (ILA), which was assembled during 2006, was commissioned on the JET RF testbed prior to installation on the JET torus. The 4 resonant double loops (RDL) of the ILA were tested at high power at 42 MHz up to 42 kV for 5 s in 10 min intervals. Low power matching studies using a saltwater load placed in front of the ILA have allowed testing and optimizing proposed matching algorithms on single RDLs, paired RDLs and finally on the full array. The upper limit of the frequency range of the ILA appears to be limited to 47–49 MHz due to the effect on the electrical lengths of the connection between the capacitors and the conjugate T point. Capacitor position scans have allowed obtaining the necessary data to confirm the RF model of the RDL which is necessary for the scattering matrix arc detection. The latter is deemed necessary in order to detect arcs at the low impedance conjugate T of the circuit. The antenna was installed onto JET during August 2007 and commissioning on plasma started May 2008. At present the commissioning of the ILA on JET is ongoing in a series of dedicated experimental campaigns.     

ICRF physics aspects of wall conditioning plasma characterization in TEXTOR

This paper focuses on encouraging results obtained on the characterization of RF produced plasmas during pulsed-mode wall conditioning discharges in ion cyclotron resonance frequency (ICRF) regime in the limiter tokamak TEXTOR. Recent Ion Cyclotron Wall Conditioning (ICWC) experiment carried out in TEXTOR tokamak, lead to the identification of various dependences of the antenna-plasma coupling efficiency on the plasma parameters for possible ICWC-discharge cleaning in ITER at half field. Our ICWC experiments emphasize on (i) study of antenna coupling during the mode conversion scenario, (ii) reproducible generation of ICRF plasmas for wall conditioning, by coupling RF power from one or two ICRF antennas and (iii) effect of application of an additional (along with toroidal magnetic field) stationary vertical (B_V ? B_T) or oscillating poloidal magnetic field (B_p ? B_T) on antenna coupling and relevant plasma parameters.     

A distributed timing and synchronization system for EAST

The distributed timing and synchronization system (DTSS) plays an important role in Experimental Advanced Superconducting Tokamak (EAST), which is one of the national key fusion research facilities in China. This system synchronizes each subsystem of EAST by using reference clock and trigger. A prototype DTSS module has been developed based on PXI bus and RIO (reconfigurable I/O) devices. The DTSS can provide reference clock in frequency up to 80 MHz. The trigger can be pre-defined from 1 ms to 6872 s with 10 ns accuracy. In addition, this system can acquire, process signals, and send output or command to other systems. The DTSS has been successfully applied to 2010 fall EAST experiment, and the results confirmed its accuracy and reliability. After the analysis of system requirement, the architecture of the DTSS and the technical implementation based on PXI are presented in this paper.     

Analysis of dynamic matching networks for the ICRF system at ASDEX Upgrade

The ICRF (Ion Cyclotron Range of Frequencies) system, used to heat the plasma of ASDEX Upgrade, consists of RF generators, 3 dB hybrids, coaxial transmission lines, matching networks and inductive loop antennas. The maximum power achievable by the generator strongly depends on the amplitude and phase of the reflection coefficient. Hence, matching of the antenna input impedance to the generator output impedance is essential for the ICRF heating system. The coupling between the ICRF antennas and the plasma is subject to relatively fast variations (few ms). The changes are caused by the modification of plasma edge conditions, such as L–H mode transitions, gas puffing or ELMs (Edge Localized Modes). These variations change the impedance of the ICRF antennas. For optimal operation of the ICRF system, a continuous, and fast matching is therefore preferable. A MATLAB^® based simulation tool has been developed to analyse matching networks for the ICRF system at ASDEX Upgrade. The program is highly flexible, and can solve the matching calculations for different conditions and configurations of the system. The results are presented in a Smith chart.     

Electromagnetic disruption analysis of 2.45 GHz LHW antenna in EAST under different plasma configurations

The 2.45 GHz lower hybrid wave (LHW) antenna is one of the key components for plasma heating and current drive on experimental advanced superconducting tokamak (EAST). In the lower hybrid current drive (LHCD) experiment, the microwave power is delivered to the plasma through the LHW antenna. During a plasma disruption, the eddy currents are induced in the antenna because of plasma current decay. These induced currents interact with the strong static magnetic field to produce forces and torques in the antenna which are one of key factors determining the design of the antenna. Therefore, this paper presents the key results of a transient electromagnetic (EM) analysis of the antenna during disruption events under different plasma configurations. Two plasma centered disruption scenarios are taken into account: exp quench and linear quench. The analysis was performed with MAXWELL, a computer code based on the finite element method. All the results are presented and discussed which will offer guidance for the design and manufacture of the antenna in future.     

Design of improved compact decoupler based on adjustable capacitor for EAST-ICRF antenna

The Ion Cyclotron Radio Frequency (ICRF) heating antenna on EAST adopts a decoupling device to constrain power coupling among the radiation straps, which was discovered shortcomings such as long size, poor contact, and etc. In order to improve these weak points, a new type decoupler with terminal-loaded tunable capacitor is designed to replace the previous design. Besides the capability of the tunable admittance parameters of decoupler, the withstand voltage of the capacitor is the most significant consideration for working under high power. Therefore, the theoretical analysis carefully elaborates the capacitor withstand voltage, and the detailed analytical equations and criteria for design are given. After the comparative analysis of theoretical calculation and 3D simulation results, the decoupler design scheme is finalized. The capacitor-loaded decoupler has been successfully adopted for ICRF antenna at port N on EAST, and achieved the optimization of adjacent port isolation from −22 to −58 dB at 37 MHz without plasma to restrict mutual coupling. The new design of the decoupler has greatly improved its compactness and automatic adjustment performance, and could be good solution for the decoupling network of ICRF antennas.     

RF calculation and thermal-stress analysis of the ECRH launcher on EAST tokamak

EAST is a medium sized superconducting tokamak with major radius R = 1.8 m, minor radius a = 0.45 m, plasma current I_p ≤ 1 MA, toroidal field B_T ≤ 3.5 T and expected plasma pulse length up to 1000 s. An electron cyclotron resonance heating (ECRH) launcher for four-beam injection is being installed on EAST tokamak. Four electron cyclotron wave beams which are generated from four sets of 140 GHz/1 MW/1000 s gyrotrons will be injected into the plasma by the spherical focusing mirrors and plane mobile mirrors. The focusing mirrors are spherical to focus Gaussian beams after reflection. Four plane mobile mirrors independently steer continuously in the poloidal and toroidal direction controlled by motors. With the suitable distance between mirrors and appropriate focal length of focusing mirror, the beam radius in the resonance layer of plasma is 31.145 mm. The heat from plasma radiation and metal losses is loaded on the mobile mirror. In order to decrease the temperature and thermal stress, the inner equivalent diameter of water channels is 8 mm and the suggested water velocity is 4 m/s.     

Development of KSTAR in-vessel components and heating systems

In-vessel components of the Korea Superconducting Tokamak Advanced Research (KSTAR) were developed for 2010 campaign to provide a crucial circumstance for achieving the strongly shaped and diverted plasma. Moreover, the in-vessel components such as limiter, divertor, passive stabilizer, in-vessel control coil (IVCC) system demonstrated good performances satisfying the original design concepts. In addition to the plasma facing components and the IVCC, in-vessel cryo-pump (IVCP) system was also installed to leverage divertor operation. Besides the in-vessel components, there have been substantial progresses in development of the heating and current drive system. The KSTAR heating and current drive system includes all kinds of the major heating systems such as neutral beam injection (NBI), ion cyclotron range of frequency (ICRF), electron cyclotron resonance heating and current drive (ECH and ECCD), lower hybrid current drive (LHCD) systems. As an initial stage for full equipment of the heating systems to total power of 26 MW, several key systems such as 1st NBI (called NBI-1), ICRF, and ECH-assisted startup system successfully demonstrated their excellent feasibilities in the design and performances for dedication to the 2010 campaign.     

Recent ICRF coupling experiments on EAST

Recent ion cyclotron resonance frequency(ICRF) coupling experiments for optimizing ICRF heating in high power discharge were performed on EAST. The coupling experiments were focus on antenna phasing and gas puffing, which were performed separately on two ports of the ion cyclotron resonance heating(ICRH) system of EAST. The antenna phasing was performed on the I-port antenna, which consists of four toroidally spaced radiating straps operating in multiple phasing cases; the coupling performance was better under low wave number ∣k_‖∣(ranging from 4.5 to 6.5). By fuelling the plasma from gas injectors, placed as uniformly spaced array from top to bottom at each side limiter of the B-port antenna, which works in dipole phasing, the coupling resistance of the B-port antenna increased obviously.Furthermore, the coupling resistance of the I-port antenna was insensitive to a smaller rate of gas puffing but when the gas injection rate was more than a certain value(1021 s~(-1)), a sharp increase in the coupling resistance of the I-port antenna occurred, which was mainly caused by the toroidal asymmetric boundary density arising from gas puffing. A more specific analysis is given in the paper.     

A repetitive pellet injection system for steady state fuelling in EAST superconducting tokamak

A new hydrogen/deuterium pellet injector has been developed for Experimental Advanced Superconducting Tokamak (EAST). The pellet injector based on a screw extruder is able to fire pellets (∅2 mm × 2 mm; frequency 1–10 Hz and velocity 150–300 m/s) in steady state mode with reliability greater than 95%. An injection line was designed for pumping propellant gas and for diagnostic purpose also. A guide tube for magnetic high-field side (HFS) injection was developed and theoretical calculation has been done. After successful engineering commissioning, the injection system served at EAST 2012 campaign and first experimental results were obtained.     

Thermal analysis of toroidal field coil in EAST at 3.7 K

The thermal performance of toroidal field (TF) coil is studied at 3.7 K in Experimental Advanced Superconducting Tokamak device (EAST) to obtain the higher stability for the higher plasma parameters operation. It is a good way to lower the operating temperature of TF coil to acquire the higher stability margin. This paper describes the structure and cooling process design of TF coil and case firstly. Based on the thermal load in the case, the thermal performance of the TF coil is performed at the plasma disruption state. The helium temperature in the cable-in-conduit conductor (CICC) and case is evaluated during the 1.5 MA plasma disruptions. Then, the experimental results of TF coil which has been cooled at 3.7 K and discharged in 10 kA are shown including the thermal loss evaluation. Finally, the thermal stability performance of TF coil is analyzed according to the 3.7 K experimental results and the stability prediction is performed at 1.5 MA plasma current operations.     

ASDEX Upgrade enhancements in view of ITER application

The aim of the ASDEX Upgrade (AUG) programme is to support the design, prepare the physics base and develop regimes beyond the baseline of ITER and for DEMO. Its ITER-like geometry, poloidal field system, versatile heating system and power fluxes make AUG particularly suited.After the transition to fully tungsten coated plasma facing components AUG could be operated without prior boronizations and a low permanent deuterium retention was found qualifying W as wall material. ITER-like baseline H-modes (H₉₈ ～ 1, β_N ～ 2) were routinely achieved up to 1.2 MA plasma currents. W concentrations could be kept at an acceptable level of <5 × 10^?5 by central wave heating (enhancing impurity outward transport) and ELM pacing with gas puffing. The compatibility of high performance improved H-modes, the ITER hybrid scenario, with an un-boronized W wall was demonstrated achieving H₉₈ ～ 1.1 and β_N up to 2.6 at modest triangularities δ ≤ 0.3. This performance is reached despite the gas puffing needed for W influx control. Increasing δ to 0.35 allowed at even higher puff rates still a H₉₈ ～ 1.1.Reliable plasma operation in support of ITER comprised the demonstration of ECRF assisted low voltage plasma start-up and current rise at toroidal electric fields below 0.3 V/m resulting in a ITER compatible range of plasma internal inductance of 0.71–0.97. Disruption mitigation is feasible using strong gas puffs, and the achieved electron densities approach values needed for runaway suppression.Present hardware extensions in support of ITER include the upgrading of ECRH by a 4 MW/10 s system with large deposition variability (tuneable frequency between 105 and 140 GHz, real-time steerable mirrors) for central heating and MHD mode control. A powerful system of 24 in-vessel coils produces error fields up to toroidal mode number n = 4 for ELM suppression and mode rotation control. In connection with a close conducting wall they will open up the road for RWM stabilization in advanced scenarios. For those we are considering LHCD for current drive and profile control with up to 500 kA driven current. The tungsten sources are dominated by sputtering from intrinsic light impurities, and the W influx from the outboard limiters are the main source for the core plasma. ICRH induced electric fields accelerate light impurities, restricting the use of ICRH to just after boronization. 4-strap antennas imbedded in extended wall structures might solve this problem. Finally, doubling the plasma volume with plasma currents above 2 MA in AUG could be the solution for a needed ITER satellite.     

Simulation of EAST off-axis neutral beam heating and current drive

The capability of off-axis neutral beam heating and current drive has been investigated with NUBEAM for Experimental Advanced Superconducting Tokamak (EAST). Three different approaches to realize off-axis Neutral Beam Injection (NBI) have been studied. Simulation results for on- and off-axis NBI are reported. The effects of the alignment of NBI relative to the magnetic field pitch on off-axis neutral beam heating and current drive are observed and discussed qualitatively. By comparing the numerical results, a most favorable off-axis NBI configuration is recommended. The capability to control sawtooth is also investigated by comparing locations of the q = 1 rational surface and the peak of the fast ion density profile.     

Study of electromagnetic noise influence on quench detection system under different discharge conditions for EAST

EAST is the first Tokamak device whose toroidal and poloidal magnet are superconducting. The enormous magnetic field energy stored in the magnet system will transfer into thermal energy and cause the damage of superconducting magnet, if a quench happened. Therefore, reliable quench detection is a key issue for steady-state operation. In addition to electromagnetic noise from poloidal magnet fields and plasma current which will experience fast current ramp rate, radio frequency noise from heating system also have some interference on quench detection system to a certain degree. The most difficult point for quench detection system is required to have more detail evaluation on electromagnetic noise interference.Recently experiments have been carried out successfully in EAST device. The steady-state operation with 1 MA of plasma current and more than 100-s plasma duration has been obtained. In the paper, the electromagnetic noise interference on quench detection system under different discharge conditions are analyzed and relative process methods are also introduced. The technological experience and experimental data are significant for the constructing ITER and similar superconducting device have been mentioned which will supply significant technological experience and experimental data for constructing ITER and similar superconducting device.     

Preliminary design and performance study of EAST liquid lithium limiter based on CPS

Lithium has the ability of H recycling suppression and impurities absorption and it can be used as plasma facing material (PFM) in tokamaks. Lithium conditioning experiments were launched on EAST, HT-7 and some other tokamaks for many years by using the methods of GDC, IRCF and evaporation. Liquid lithium has better performances in effective lifetime and heat removal aspects compared to non-liquid lithium. While, applying liquid lithium in the tokamak would cause the safety problem as the lithium can react with many substances violently and the magnetohydrodynamic behavior is difficult to be handled. EAST liquid lithium limiter (LLL) system is under developing and will be applied in EAST to study the main technologies of the liquid lithium application. The normal operation temperature of the limiter is expected as 230–550 °C under the active cooling of water. Capillary porous system (CPS) is used to prevent the lithium from splashing under large electromagnetic force by increasing the surface tension of the lithium. In order to investigate the cooling performance of the cooling design, the thermal-hydraulic analysis was done which shows that with 3 m/s flowing velocity, the water can keep the limiter under 550 °C all the time if the heat flux is lower than 0.7 MW/m². Under heat flux of 1 MW/m², the limiter should be retreated within 7 s to avoid erosion. The pressure drop of the coolant under 3 m/s is less than 40 kPa with temperature difference nearly 34 °C which meet the design requirements very well. The key manufacture process and technologies like vacuum bonding between the CuCrZr heat sink and 316L guide plate were well studied in the R&D process. The heating test on the test bench showed that the CPS can be heated efficiently by the heaters attached into the heat sink.