Thermal analysis and optimization of the EAST ICRH antenna

The ion cyclotron resonance of frequency heating(ICRH) plays an important role in plasma heating.Two ICRH antennas were designed and applied on the EAST tokamak.In order to meet the requirement imposed by high-power and long-pulse operation of EAST in the future,an active cooling system is mandatory to be designed to remove the heat load deposited on the components.Thermal analyses for high heat-load components have been carried out,which presented clear temperature distribution on each component and provided the reference data to do the optimization.Meanwhile,heat pipes were designed to satisfy the high requirement imposed by a Faraday shield and lateral limiter.     

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.     

Design and control of coaxial switching system for ion cyclotron wave resonance heating system of EAST

Ion cyclotron wave resonance heating system(ICRH) which is one of the most important auxiliary system in EAST provides conditions for heating the plasma. In order to make the whole transmission network of ICRH form the required state, the coaxial switching system was developed, which mainly consists of 11 same coaxial switches. Each coaxial switch has a controller with an RS485 communication interface and is able to switch between two states. All controllers are integrated to an RS485 hub which connects to the computer. A master computer software is developed to control the coaxial switching system so as to achieve the state needs of ICRH. Moreover, several rounds of experiments show that the coaxial switching system operates reliably. The coaxial switching system has already been applied to normal use currently.     

EAST离子回旋加热系统高功率射频放大器阳极电源的研制

离子回旋共振加热是EAST超导托卡马克核聚变实验中重要的辅助加热手段。高性能的高功率射频放大器阳极电源对整个加热系统的稳定运行起重要作用。本工作设计了基于脉冲阶梯调制（PSM）技术的阳极电源及其控制保护系统，通过采集电源的实验数据对电源的设计进行了验证。实验结果证明，本阳极电源的设计和参数选择均是合理正确的，电源的开通和关断以及控制保护的数据指标完全达到设计要求。     

Development and Preliminary Commissioning Results of a Long Pulse 140 GHz ECRH System on EAST Tokamak(Invited)

A long pulse electron cyclotron resonance heating(ECRH)system has been developed to meet the requirements of steady-state operation for the EAST superconducting tokamak,and the first EC wave was successfully injected into plasma during the 2015 spring campaign.The system is mainly composed of four 140 GHz gyrotron systems,4 ITER-Like transmission lines,4 independent channel launchers and corresponding power supplies,a water cooling,control &inter-lock system etc.Each gyrotron is expected to deliver a maximum power of 1 MW and be operated at 100-1000 s pulse lengths.The No.1 and No.2 gyrotron systems have been installed.In the initial commissioning,a series of parameters of 1 MW 1 s,900 k W 10 s,800 k W 95 s and650 k W 753 s have been demonstrated successfully on the No.1 gyrotron system based on calorimetric dummy load measurements.Significant plasma heating and MHD instability suppression effects were observed in EAST experiments.In addition,high confinement(H-mode)discharges triggered by ECRH were obtained.     

Effects of fast ions produced by ICRF heating on the pressure at EAST

Ion cyclotron resonance heating(ICRH),which can produce fast ions,is an important auxiliary heating method at EAST.To analyze the effect of ICRH-induced fast ions on the plasma pressure at EAST,simulations are performed using TRANSP and TORIC codes.It is found that the ICRF-induced fast ion pressure cannot be negligible when the ICRF power is sufficiently high.The magnitude of the total ion pressure can be raised up to 60%of the total pressure as the input power rises above 3 MW.The pressure profile is also significantly modified when the resonant layer is changed.It is shown that by changing the wave frequency and antenna position,the total ion pressure profile can be broadened,which might provide an option for profile control at EAST.     

Results of ICRF Heating Experiments from the EAST 2010 Campaign

Radio frequency(RF) plasma heating in ion cyclotron range of frequencies(ICRF)was successfully performed on the Experimental Advanced Superconducting Tokamak(EAST).This is mainly because lithium wall conditioning was routinely used to reduce both impurity and hydrogen(H) recycling and to improve the ICRF power absorption.Mainly ICRF heating of the H minority regime at 27 MHz has been applied in deuterium plasmas.The ion cyclotron resonance heating(ICRH) is found to depend strongly on plasma preheating.The ICRH efficiency can be much improved in conjunction with the lower hybrid wave(LHW).Effective ion and electron heating was observed with the H minority heating mode.The increase of the stored energy reached30 kJ in L-mode plasma by using the ICRF power of 1.0 MW alone when the H cyclotron resonance layer was at plasma center.     

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.     

Polarization and Mode Control of EAST 140 GHz ECRH&CD System

The 140 GHz electron cyclotron heating and current drive (ECRH&CD) project was launched in 2011 on EAST tokamak facility, which is designed to launch 4 MW of total power for the duration up to 1,000 s into the plasma. The heating and current drive efficiencies depend on the wave coupling mode in plasma and the coupling performance relies on polarization characteristics of injected beam, so polarization control is necessary for efficient plasma heating and current drive. Two polarizer miter bends will be used to control the wave polarization for each transmission line on EAST ECRH&CD system, any required wave polarization can be produced by adjusting the mirror rotation angle of each polarizer miter bend. This work mainly shows the calculated polarizer mirror settings as a function of the injection angles for pure second extraordinary harmonic mode coupling.     

Theoretical Analysis of Triple Liquid Stub Tuner Impedance Matching for ICRH on Tokamaks

The impedance matching is crucial for continuous wave operation of ion cyclotron resonance heating(ICRH) antennae with high power injection into plasmas.A sudden increase in the reflected radio frequency power due to an impedance mismatch of the ICRH system is an issue which must be solved for present-day and future fusion reactors.This paper presents a method for theoretical analysis of ICRH system impedance matching for a triple liquid stub tuner under plasma operational conditions.The relationship of the antenna input impedance with the plasma parameters and operating frequency is first obtained using a global solution.Then,the relations of the plasma parameters and operating frequency with the matching liquid heights are indirectly obtained through numerical simulation according to transmission line theory and matching conditions.The method provides an alternative theoretical method,rather than measurements,to study triple liquid stub tuner impedance matching for ICRH,which may be beneficial for the design of ICRH systems on tokamaks.     

Water Flow Calorimetry System of EAST Neutral Beam Injector

Neutral beam injection (NBI) is recognized as one of the most effective means of plasma heating. The EAST NBI water flow calorimetry system (WFCS) based on PCI extensions for instrumentation (PXI) was established, it can measure temperature rise and flow rate of cooling water of the heat load components, and achieve beam power distribution and neutralization efficiency. Experimental data obtained from WFCS are feedback of the ion source operation state and direct the operation parameter optimization of the ion source. Experimental results show that the WFCS is stable, reliable, and meet the experimental requirements fully.     

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.     

Design of Main Control Console Software in EAST Neutral Beam Injector’s Control System for the First Beam Line

Neutral beam injector is one of the main plasma heating and plasma current driving methods for experimental advanced superconducting tokomaks (EAST). In order to realize visual operation of EAST neutral beam injector’s control system (NBICS), main control console (MCC) is developed to work as the human–machine interface between the NBICS and physical operator. It can meet the requirements of visual control of NBICS by providing a user graphic interface. With the specific algorithms, the setup of power supply sequence is relatively independent and simple. Displaying the real-time feedback of the subsystems provides a reference for operators to monitor the status of the system. The MCC software runs on a Windows system and uses C++ language code while using client/server (C/S) mode, multithreading and cyclic redundancy check technology. The experimental results have proved that MCC provides a stability and reliability operation of NBICS and works as an effective man–machine interface at the same time.     

Integrated Operating Scenario to Achieve 100-Second,High Electron Temperature Discharge on EAST

Stationary long pulse plasma of high electron temperature was produced on EAST for the first time through an integrated control of plasma shape,divertor heat flux,particle exhaust,wall conditioning,impurity management,and the coupling of multiple heating and current drive power.A discharge with a lower single null divertor configuration was maintained for 103 s at a plasma current of 0.4 MA,q_(95)≈7.0,a peak electron temperature of 4.5 keV,and a central density n_e(0)~2.5×10~(19) m~(-3).The plasma current was nearly non-inductive(V_(loop) 0.05 V,poloidal beta ~0.9) driven by a combination of 0.6 MW lower hybrid wave at 2.45 GHz,1.4 MW lower hybrid wave at 4.6 GHz,0.5 MW electron cyclotron heating at 140 GHz,and 0.4 MW modulated neutral deuterium beam injected at 60 kV.This progress demonstrated strong synergy of electron cyclotron and lower hybrid electron heating,current drive,and energy confinement of stationary plasma on EAST.It further introduced an example of integrated "hybrid" operating scenario of interest to ITER and CFETR.     

Realization of minute-long steady-state H-mode discharges on EAST

In the 2016 EAST experimental campaign,a steady-state long-pulse H-mode discharge with an ITER-like tungsten divertor lasting longer than one minute has been obtained using only RF heating and current drive,through an integrated control of the wall conditioning,plasma configuration,divertor heat flux,particle exhaust,impurity management,and effective coupling of multiple RF heating and current drive sources at high injected power.The plasma current (Ip ～ 0.45 MA) was fully-noninductively driven (Vloop ＜ 0.0 V) by a combination of ～2.5 MW LHW,～0.4 MW ECH and ～0.8 MW ICRF.This result demonstrates the progress of physics and technology studies on EAST,and will benefit the physics basis for steady state operation of ITER and CFETR.     

Numerical study on the loss of fast ions produced by minority ion cyclotron resonance heating in EAST

The classical prompt loss of fast ions produced by minority ion cyclotron resonance heating(ICRH)is studied by a guiding center orbit following code in the Experimental Advanced Superconducting Tokamak(EAST).It is found that the loss of fast ions produced by ICRH mainly appears in both ends of the resonance layer,while the loss of fast ions in the middle resonance layer is very small.The dominant fast loss comes from trapped ions,rather than from passing ions.Controlling the location of resonance layer at the plasma core may be more beneficial to the EAST tokamak ICRH.In addition,the loss distribution of fast ions is studied.The results show that the fast ions are mainly lost near the midplane in the poloidal direction,but almost uniformly in the toroidal direction.Moreover,we investigate the dependence of fast ion loss on the ICRH power.The simulation results show that the loss fraction of fast ions in both ends of the resonance region increases with the ion cyclotron range of frequencies(ICRF)power,but barely affects the loss of fast ions in the middle region.     

EAST装置高功率宽带射频发射机系统研制

离子回旋波加热系统是EAST装置最重要的辅助加热工具,作为系统最核心的分系统之一,高功率射频发射机为加热等离子体提供射频波能量,对提高等离子体运行参数起着极为重要的作用。基于电路分析、传输线和波导谐振腔等相关工程理论,本文系统地总结了射频发射机系统高功率放大器输入输出回路、放大器级间匹配、寄生振荡抑制、腔体冷却等部分的设计原理和实现方法。在假负载上进行了系统测试,在设计频段内获得了1.5 MW的射频输出功率,测试结果表明系统达到了设计的技术指标。通过两轮EAST射频加热实验验证,发射机系统运行稳定可靠,满足射频加热等相关物理实验要求。     

Mechanical design of the second ICRF antenna for EAST

In order to satisfy the requirements of heating plasma on EAST project, 3 MW ion cyclotron range of frequency (ICRF) heating system will be available at the second stage. Based on this requirement, the second ICRF antenna, has been designed for EAST. The antenna which is planned to operate with a frequency ranging from 30 MHz to 110 MHz, comprises four poloidal current straps. The antenna has many cooling channels inside the current straps, faraday shield and baffle to remove the dissipated RF loss power and incoming plasma heat loads. The antenna is supported via a cantilever support box to the external support structure. Its assembly is plugged in the port and fixed on the support box. External slideway and bellows allow the antenna to be able to move in the radial direction. The key components of the second ICRF antenna has been designed together with structural and thermal analysis presented.     

EAST-ICRH发射机宽频阻抗匹配网络的研究及实现

EAST(Experimental Advanced Superconducting Tokamak)是我国开展受控核聚变研究的新一代实验装置,离子回旋波加热(ICRH)是在该装置中加热等离子体的重要手段之一。论文介绍了离子回旋发射机中阻抗匹配网络的原理及实现方法,并采用LabVIEW软件和PLC相结合的方式开发了一套阻抗匹配网络调节系统。     

The Influence of Neutral Beam Injection on the Heating and Current Drive with Electron Cyclotron Wave on EAST

Both neutral beam injection(NBI) and electron cyclotron resonance heating(ECRH) have been applied on the Experimental Advanced Superconducting Tokamak(EAST)in the 2015 campaign. In order to achieve more effective heating and current drive, the effects of NBI on the heating and current drive with electron cyclotron wave(ECW) are analyzed utilizing the code TORAY and experimental data in the shot #54411 and #54417. According to the experimental and simulated results, for the heating with ECW, NBI can improve the heating efficiency and move the power deposition place towards the inside of the plasma. On the other hand, for the electron cyclotron current drive(ECCD), NBI can also improve the efficiency of ECCD and move the place of ECCD inward. These results will be valuable for the center heating, the achievement of fully non-inductive current drive operation and the suppression of magnetohydrodynamic(MHD) instabilities with ECW on EAST or ITER with many auxiliary heating methods.