Investigation of electron cyclotron current drive efficiency on the J-TEXT tokamak

Electron cyclotron current drive (ECCD) efficiency research is of great importance for the neoclassical tearing mode (NTM) stabilization. Improving ECCD efficiency is beneficial for the NTM stabilization and the ECCD power threshold reduction. ECCD efficiency has been investigated on the J-TEXT tokamak. The electron cyclotron wave (ECW) power scan was performed to obtain the current drive efficiency. The current drive efficiency is derived to be approximately η₀ = (0.06–0.16) × 10¹⁹ A m⁻² W⁻¹ on the J-TEXT tokamak. The effect of the residual toroidal electric field has been included in the determination of the current drive efficiency, which will enhance the ECCD efficiency. At the plasma current of I_p = 100 kA and electron density of n_e = 1.5 × 10¹⁹ m⁻³, the ratio of Spitzer conductivity between omhic (OH) and ECCD phases is considered and the experimental data have been corrected. The correction results show that the current drive efficiency η1 caused by the fast electron hot conductivity decreases by approximately 79%. It can be estimated that the driven current is approximately 24 kA at 300 kW ECW power.     

Optimization of X-mode electron cyclotron current drive in high-electron-temperature plasma in the EAST tokamak

A discharge with electron temperature up to 14 keV has been achieved in EAST. Analysis of the electron cyclotron current drive(ECCD) efficiency at high electron temperature under EAST parameters is presented using C3PO/LUKE code. Simulation results show that the ECCD efficiency of X-mode increases with central electron temperature up to 10 keV and then starts to decrease above 10 keV, at a specific magnetic field and toroidal angle. The efficiency degradation is due to the presence of the third ...     

Recent progress of the ECRH system and initial experimental results on the J-TEXT tokamak

In order to broaden the range of the plasma parameters and provide experimental conditions for physical research into high-performance plasma, the development of the electron cyclotron resonance heating (ECRH) system for the J-TEXT tokamak was initiated in 2017. For the first stage, the ECRH system operated successfully with one 105 GHz/500 kW/1 s gyrotron in 2019. More than 400 kW electron cyclotron (EC) wave power has been injected into the plasma successfully, raising the core electron temperature to 1.5 keV. In 2022, another 105 GHz/500 kW/1 s gyrotron completed commissioning tests which signifies that the ECRH system could generate an EC wave power of 1 MW in total. Under the support of the ECRH system, various physical experiments have been carried out on J-TEXT. The electron thermal transport in ECRH plasmas has been investigated. When ECRH is turned on, the electron thermal diffusivity significantly increases. The runaway current is elevated when a disruption occurs during ECRH heating. When the injected EC wave power is 400 kW, the conversion efficiency of runaway current increases from 35% to 75%. Fast electron behavior is observed in electron cyclotron current drive (ECCD) plasma by the fast electron bremsstrahlung diagnostic (FEB). The increase in the FEB intensity implies that ECCD could generate fast electrons. A successful startup with a 200 kW ECW is achieved. With the upgrade of the ECRH system, the J-TEXT operational range could be expanded and further relevant research could be conducted.     

EAST电子回旋波与低杂波双波协同的数值模拟

为在EAST装置中优化电子回旋波与低杂波双波协同的电流驱动效率,从而获得更大的协同驱动电流以维持长脉冲运行,本文以双波协同驱动电流的物理机制为基础,运用模拟程序C3PO/LUKE对EAST参数下电子回旋电流驱动与低杂波电流驱动的协同效果进行了数值模拟计算,给出了协同电流和协同因子。计算结果表明：当两波驱动电流密度峰值的位置一致时协同效果最佳;而电子密度和电子温度的增加可能导致两波重叠区域的变化进而影响协同效果。     

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.     

Predictive Simulations of Plasma Heating and Current Drive by Neutral Beam Injection on EAST

Long pulse and high performance steady-state operation is the main scientific mission of experimental advanced superconducting tokamak (EAST). In order to achieve this objective, high-power auxiliary heating systems are essential. Radio frequency (RF) wave heating and neutral beam injection (NBI) are two principal methods. NBI is an effective method of plasma heating and current drive, and it has been used in many magnetic confinement fusion devices. Based on the plasma equilibrium of EAST (Li et al., Plasma Phys Control Fusion 55:125008, 2013) plus previous EAST experimental data used as initial conditions, the NBI module (Polevoi et al., JAERI-Data, 1997) employed in automated system for transport analysis (ASTRA) code (Pereverzev et al., IPP-Report, 2002) is applied to predict the effects of plasma heating and current drive with different neutral beam injection power levels. At certain levels of plasma densities and plasma current densities, the simulation results show that the NBI heats plasma effectively, also increases the proportions of NB current and bootstrap current among total current significantly.     

A scoping study of the application of neutral beam heating on the TCV tokamak

The TCV tokamak contributes to the physics understanding of fusion plasmas, broadening the parameter range of reactor relevant regimes, by investigations based on an extensive use of the existing main experimental tools: flexible shaping and high power real time-controllable electron cyclotron heating (ECH) and current drive (ECCD) systems. A proposed implementation of direct ion heating on the TCV by the installation of a 20–35 keV neutral beam injection (NBI) with a total power of 1–3 MW would permit an extension of the accessible range of ion to electron temperatures (T_i/T_e ～ 0.1–0.8) to well beyond unity, depending on the NBI/ECH mix and the plasma density. A NBI system would provide TCV with a tool for plasma study at reactor relevant T_i/T_e ratios ～1 and in investigating fast ion and MHD physics together with the effects of plasma rotation and high plasma β scenarios. The feasibility studies for a NBI heating on TCV presented in this paper were undertaken to construct a specification for the neutral beam injectors together with an experimental geometry for possible operational scenarios.     

Observation and characterization of the effect of electron cyclotron waves on toroidal rotation in EAST L-mode discharges

The change in the toroidal rotation of plasma caused by electron cyclotron wave (ECW) injection has been observed in EAST.It is found that the response of the rotation is similar for all possible ECW toroidal injection angles.The core toroidal rotation velocity increases in the co-current direction along with a rise in the plasma temperature and stored energy.The profile of the electron temperature,ion temperature and toroidal rotation velocity gradually become peaked.The change in toroidal rotation in the core increases with the ECW injection power.Different behavior is observed when the ECWs are injected into low hybrid current drive (LHCD) target plasmas,where the electron temperature and rotation profile become peaked,while the ion temperature profile flattens after ECW injection,suggesting different transport characteristics in energy and momentum.     

Effect of Plasma Rotation on Neutral Beam Heating and Current Drive in Tokamaks

For a rapidly rotating plasma, the effects of the resulting Doppler shift have to be included in the neoclassical theory of neutral beam heating, current drive, and plasma transport. In this paper, an improved simulation of neutral beam injection (NBI) and current drive in rotating plasmas is introduced. NBI is simulated using the Monte Carlo code NUBEAM along with the transport code ONETWO. The physical characteristics of heating and current drive for co- and counter-NBI are investigated for non-rotating, co-rotating, and counter-rotating plasmas, all of which can take place in the experiments. In general, it is found that rotation of the plasma can increase the NBI power deposition on the plasma electrons but has little effect on the ions. Moreover, plasma heating by co-NBI is more efficient than that by counter-NBI. For neutral beam current drive, because of the Doppler shift, co-rotation (counter-rotation) of the bulk plasma tends to decrease the co-NBI (counter-NBI) driven current. On the other hand, due to trapping and orbit loss of the fast ions, co-rotation (counter-rotation) has little effect on the counter-NBI (co-NBI) driven current. The results are applied to the forthcoming NBI heating and current drive experiments of the EAST tokamak and should also be useful in the design of experiments in ITER.     

Numerical Simulation of Non-Inductive Current Driven Scenario in EAST Using Neutral Beam Injection

For achieving the scientific mission of long pulse and high performance operation,experimental advanced superconducting tokamak(EAST) applies fully superconducting magnet technology and is equiped with high power auxiliary heating system.Besides RF(Radio Frequency) wave heating,neutral beam injection(NBI) is an effective heating and current drive method in fusion research.NBCD(Neutral Beam Current Drive) as a viable non-inductive current drive source plays an important role in quasi-steady state operating scenario for tokamak.The non-inductive current driven scenario in EAST only by NBI is predicted using the TSC/NUBEAM code.At the condition of low plasma current and moderate plasma density,neutral beam injection heats the plasma effectively and NBCD plus bootstrap current accounts for a large proportion among the total plasma current for the flattop time.     

EAST托卡马克的中性束注入方案

高能中性束注入(Neutral beam injection,NBI)是核聚变装置托卡马克采用的芯部辅助加热和非感应电流驱动主要手段之一.本文介绍了国家大科学工程全超导托卡马克实验装置(Experimental advanced super-conductingtokamak,EAST)上的高能NBI加热方案及注入器的工程要求,并讨论了中性束在EAST等离子体中的传输等相关问题.     

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.     

Real-time mirror steering for improved closed loop neoclassical tearing mode suppression by electron cyclotron current drive in DIII-D

The development and operation of the neoclassical tearing mode (NTM) avoidance and control system for DIII-D, which uses six sets of real-time steerable mirrors in order to move the electron cyclotron current drive (ECCD) deposition location in plasma, is described. The real-time DIII-D NTM control algorithm residing in the Plasma Control System (PCS) automatically detects an NTM by analysis of the Mirnov diagnostics, employs motional Stark effect (MSE) EFIT MHD equilibrium reconstruction to locate the rational q-surface where the NTM island can be found, then calculates the appropriate mirror position for alignment of the ECCD with the island using ray tracing. The control commands from PCS are sent to the electron cyclotron system to switch on and off or modulate the gyrotrons and to the steerable mirror system to move the steerable mirrors to the requested positions. Successful NTM suppression has been achieved in DIII-D using this control system to rapidly align the NTM island and the ECCD deposition location, and to actively maintain the alignment as plasma conditions change.     

First Achievement of Plasma Heating for EAST Neutral Beam Injector

As one of the most effective methods for plasma heating,a neutral beam injector(NBI) achieved plasma heating and current driving for the first time in EAST 2014 experimental campaign.According to the research plan of the EAST physics experiment,the first NBI(EASTNBI-1) has been built and become operational in 2014.In this article,the latest experiment results of EAST-NBI-1 are reported as follows:(1) EAST achieves H-mode plasma in the case of NBI heating alone,(2) EAST achieves 22 s long pulse stable H-mode plasma in the case of sinndtaneous NBI and lower hybrid wave(LHW) heating.The measurement data show that the loop voltage decreased and the plasma stored energy increased obviously.The results indicate that EAST-NBI-1 has achieved plasma heating and current driving,and thus lays a foundation for the construction of EAST-NBI-2,which will be built in a few months this year.     

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.     

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.     

Evaluation of electron cyclotron current drive performance for CFETR

A number of simulations of electron cyclotron current drive (ECCD) have been carried out for the China Fusion Engineering Test Reactor (CFETR) using the C3PO/LUKE code to investigate the performance and optimize schemes of power injection for the design of the launcher. The operation ranges of the toroidal field, cutoff density, and resonance layer location are given at different source frequencies in CFETR phases I and II. A comparison of ECCD performance between the horizontal and top port launch is presented. ECCD efficiency (r_EC) estimated for CFETR phase I isr_EC=0.21 for top port launch andr_EC=0.20 for horizontal port launch. The ECCD efficiency and second-harmonic absorption is calculated at different wave frequencies (from 170 to 230 GHz) in CFETR phase II. It is found that the highest driven efficiency is obtained at 210 GHz with the toroidal field of 6.5 T, and the second-harmonic absorption increases rapidly with the increase of frequency.     

Realization of Te0>10 keV long pulse operation over 100 s on EAST

In 2021, EAST realized a steady-state long pulse with a duration over 100 s and a core electron temperature over 10 keV. This is an integrated operation that resolves several key issues,including active control of wall conditioning, long-lasting fully noninductive current and divertor heat/particle flux. The fully noninductive current is driven by pure radio frequency(RF) waves with a lower hybrid current drive power of 2.5 MW and electron cyclotron resonance heating of 1.4 MW. This is an excell...     

Development of a Data Acquisition Control System for the First NBI on EAST

Neutral beam injection （NBI） heating is one of the most efficient auxiliary plasma heating methods for fusion devices. The data acquisition control system （DACS） with PXI （pe- ripheral component interconnect extensions for instrumentation） data acquisition card for the first NBI system in the experimental advanced superconducting tokamak （EAST） is presented in this paper. As an important sub-system, DACS is designed to obtain physical measurement signals in the EAST NBI system and to deal and store these data with the Lempel-Ziv-Oberhumer （LZO） lossless data compression algorithm, as well as offer convenient data call-back and access inter- faces to the user for examining and analyzing the data. Experimental results show that accurate data will ensure that researchers correctly analyze it and then properly adjust the experimental parameters or operation, so DACS should take a large step in improving experimental efficiency. Tile hardware and software sections are briefly presented in this paper, and now this system has been tested to be able to work reliably and steadily.     

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.