Non-inductive Current Research Articles

Controlling three-dimensional magnetic island appearance with external current drive in the Chinese first quasi-axisymmetric stellarator

In this study, the impact of a non-inductive current drive, such as electron cyclotron current drive, on three-dimensional (3D) magnetic islands in the high-β equilibrium of the Chinese First Quasi-axisymmetric Stellarator (CFQS) was investigated using the HINT code. In the case of a high-β equilibrium (volume-averaged plasma beta <β> ∼ 0.74% and bootstrap current I bs∼ 24.5 kA), two m/n = 4/2 rational surfaces with large magnetic islands develop (Wang et al 2021 Nucl. Fusion 61 036021). The islands can be effectively controlled using a constant or a Gaussian current density profile, depending on the direction and amplitude of the current. With a constant current density amounting to a total current of −6 kA, the rotational transform profile can be modified such that the m/n = 4/2 rational surface is eliminated and the island is suppressed. For the Gaussian current density profile, the magnetic island can also be suppressed using a smaller total current of ∼−2 kA to adjust the iota profile. These results suggest that in the CFQS stellarator, the external current drive might be an efficient approach for controlling 3D magnetic islands and consequently improving plasma confinement.

Generation of energetic electrons by an electron cyclotron wave through stochastic heating in a spherical tokamak

This study presents novel findings on stochastic electron heating via a random electron cyclotron wave (ECW) in a spherical tokamak. Hard x ray measurements demonstrate the time evolution of hard x ray counts at different energy bands, consistent with predictions from the stochastic heating model. The ECW heating rate shows a positive correlation with applied power, confirming the effectiveness of stochastic heating. Remarkably, the ECW-driven plasma current remains insensitive to ECW incidence angle, consistent with model predictions. The observed stochastic heating of electrons offers potential for exploring innovative non-inductive current drive modes in spherical tokamaks. This research contributes to the understanding of plasma behaviour and motivates the development of new models for non-inductive current drive in fusion devices.

RF conditioning to suppress multipactor discharge for helicon wave current drive in KSTAR

The efficient off-axis non-inductive current drive using auxiliary power is essential for the steady state operation of tokamaks. In previous studies, a helicon wave current drive has been proposed for high beta plasma in several fusion devices, and the 476-MHz radio-frequency (RF) system for the proposed helicon wave has been installed in KSTAR. However, multipactor discharge occurred inside the waveguide system, including the vacuum feedthrough (VFT), and it interrupted the normal RF propagation to plasma. Thus, for successful helicon wave current drive, suppressing the multipactor discharge in the waveguide system, including the 6–1/8″ coaxial type VFT and the newly manufactured helical type traveling wave antenna (TWA) is essential before installation in KSTAR. RF conditioning utilizing a short pulse modulation during a long period helps suppress the multipactor discharge, enabling the 1-MW 476-MHz RF to pass through the entire waveguide system despite the existence of multipactor discharge. Following RF conditioning in VFT with a short pulse modulation, the newly manufactured helical type TWA and VFT for helicon wave current drive have been successfully installed in KSTAR.

Realization of T e0 > 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 excellent experimental platform on the timescale of hundreds of seconds for studying multiscale instabilities, electron-dominant transport and particle recycling (plasma-wall interactions) under weak collisionality.

Synergistic current drive of helicon wave and lower hybrid wave in HL-2M

Non-inductive current drive plays a crucial role in tokamak, especially for its steady state operations. Recently, the helicon wave (HW) has been regarded as a promising tool for driving off-axis plasma current in reactor-grade machine. The lower-hybrid wave (LHW) is the most effective radio-frequency current drive method, however, it has the drawback, which is limited by the conditions of wave accessibility in the high parameter tokamak, making the wave power usually damped at the plasma edge. HW can spiral towards the plasma centre directly under a high electron density. To obtain a long pulse steady state operation of reactor tokamak, the complementarity of HW and LHW in the aspect of driven current distribution in the high parameter tokamak is considered. The synergy current drive of the HW and the LHW is studied numerically in the steady-state scenario of HL-2M. According to the fast wave dispersion relation of plasma, the HW parameters, including its wave frequency and launched parallel refractive index, are obtained firstly. Results of GENRAY code simulation show that a single pass wave power absorption of the HW can be obtained generally through the electron Landau damping and transit time magnetic pumping effects. On the other hand, the LHW parameters are adopted from the equipped system on the machine. Results of single pass wave absorption are also obtained in the case of LHW. And then, the synergy effects of HW and LHW are studied numerically based on the GENRAY/CQL3D models. The cooperation of these two waves results in a broad plasma current distribution along the radial direction (&lt;inline-formula&gt;&lt;tex-math id="M2"&gt;\begin{document}$\rho = $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20231077_M2.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20231077_M2.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt;0.2－0.9) in the machine. Taking the electron distribution functions of these waves into account, it is clear that the electrons are accelerated by the HW in the parallel magnetic field direction, resulting in more electrons entering the region of LHW resonance area. As the consequence, a net plasma current appears. Furthermore, a fine-grained parametric scan is performed by changing the launched parallel refractive index of HW, and the results indicate that positive synergy effects can be generally observed once the related wave current drive profiles are overlapped. Finally, the synergy factor is shown to be proportional to this overlap and reaches its maximum value of 1.18 in HL-2M.

Improvement of zero-dimensional system model and its analysis and prediction of steady-state operating regime on EAST

The zero-dimensional system model has been widely used for predicting and analyzing plasma performance in fusion reactors and designing next-generation tokamaks. These models can quickly scan and calculate various parameter, and can be used for the design of device reference operation point and preparation for more accurate one-dimensional numerical simulations. They can also be used to predict device operational parameters and heating/ current drive conditions, providing a quick reference for experimental design. However, relying on physical approximations and empirical formulas can lead to significant systematic errors. In this work we introduce a plasma equilibrium program to obtain the main plasma profile parameters and their calculations based on magnetic surface information. The bootstrap current calculation is improved by introducing the relationship between the bootstrap current coefficient of the Sauter model and the collision rate change. The improved model is validated by using experimental results from EAST, and the results of the zero-dimensional system model calculations are found to be consistent with the results of kinetic equilibrium analysis. Based on the improved model and existing experimental results, the required heating/current drive power and achievable normalized beta for steady-state, long-pulse operation of the 500 kA plasma current on EAST are analyzed and predicted. The calculation results show that EAST can achieve steady-state operation at the 500 kA plasma current with bootstrap current fraction over 50% in the parameter range of 7.0–9.5 MW heating/driving power, &lt;inline-formula&gt;&lt;tex-math id="M5"&gt;\begin{document}$ {H}_{98} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M5.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M5.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt;is 1.25–1.35, and &lt;inline-formula&gt;&lt;tex-math id="M6"&gt;\begin{document}$ {f}_{{\rm{n}}{\rm{G}}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M6.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M6.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt;~0.9. Additionally, to maintain the total non-inductive current, the total heating/current drive power needs to be highly sensitive to plasma confinement and density, which is the most effective way to increase the bootstrap current fraction and reduce the peak heat loads on the divertor. Improving plasma confinement is the most effective way to achieve high bootstrap current fraction and reduce the peak heat load on the divertor. In this work, we also analyze the effect of heating power ratio on the bootstrap current, showing that adjusting the power ratio can change the bootstrap current fraction, and we further analyze the long-pulse operating region of EAST with a plasma current of 500 kA. In the range of 9.5 MW total heating/current driving power, &lt;inline-formula&gt;&lt;tex-math id="M7"&gt;\begin{document}$ {H}_{98} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M7.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M7.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; is 1.0–1.4, and normalized electron density &lt;inline-formula&gt;&lt;tex-math id="M8"&gt;\begin{document}$ {f}_{{\rm{n}}{\rm{G}}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M8.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M8.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; is 0.8–1.0, high-performance long-pulse or fully non-inductive steady-state operation can be achieved, supporting the research on the physics of ITER and CFETR steady-state operation modes. In general, improving the plasma confinement performance can achieve fully non-inductive operation at lower heating/driving power while maintaining the same plasma parameters, and expand the plasma operating regime, which is the most effective way to achieve high-parameter steady-state operation of the plasma.

Spectral Intensity of Electron Cyclotron Radiation Emerging from the Plasma to the First Wall in ITER

It is predicted that in ITER, due to high values of electron temperature and magnetic field strength, electron cyclotron (EC) radiation emitted by plasma will be a significant source (together with external EC radiation injected for auxiliary plasma heating and non-inductive current drive) of additional thermal and electromagnetic loads for microwave and optical diagnostics. The spectral distribution of plasma EC radiation is particularly important to consider in millimeter-wave diagnostics, namely for high- and low-magnetic-field side reflectometry, plasma position reflectometry, and collective Thomson scattering diagnostic, because the transmission lines of these diagnostics yield the transport of EC waves emitted by the plasma. The development of semi-analytical methods used to describe the spectral distribution of plasma-generated EC radiation in tokamaks, starting from the work of S. Tamor, is based on the dominance of multiple reflections of this radiation from the first wall in a toroidal axially symmetric vacuum chamber. Here, we present calculations using the CYNEQ code of the spectral intensity of the EC radiation emerging from the plasma to the first wall and port plugs for five scenarios of ITER operation. This code uses the symmetry-based effect of approximate isotropy and homogeneity of radiation intensity in a substantial part of the phase space and has been successfully tested by comparison with first-principles codes. The energy flux density in the range of 30–200 kW/m2 is predicted for wall reflectance in the range of 0.6–0.95. The possible effect of this radiation on in-vessel components and diagnostics is assessed by calculating the surface density of the energy absorbed by various materials of the ITER first wall.

Impact of Neutral Beam and Plasma Shaping on the Current Drive Efficiency and Fast Ion Losses in a Compact Tokamak FNS-ST

Fusion neutron source (FNS)-spherical tokamak (ST) will be a FNS based on ST ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$R =0.5$ </tex-math></inline-formula> m, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$a =0.3$ </tex-math></inline-formula> m, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$k =2.75$ </tex-math></inline-formula> , and <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$B =1.5$ </tex-math></inline-formula> T) with a steady-state neutron yield ~1018 n/s. The noninductive current drive and neutron generation is to be maintained by neutral beam injection (NBI). Beam in toroids (BTOR) code (Python) is used to reproduce NBI and plasma shaping with high degree of detail and to perform parametric studies of neutral beam (NB) driven effects. The beam model light neutral beam (LNB) is inherited from the original beam transmission (BTR) code, used since 2005 for NBI design and optimizations. The BTOR methodology combines NB statistical description with high-performance analytical models of particle tracing, which allows one to get results several orders faster, if compared to the conventional (Monte–Carlo) approach, and still with a good agreement. The results obtained for the FNS-ST device prove that the entire beam efficiency highly depends on the beam–plasma mutual size and shaping, including the beam cross section and aiming, plasma aspect ratio, elongation, and triangularity. The results are especially important for a steady-state current drive and fusion control in low aspect ratio tokamaks; the BTOR method is also applicable to a conventional tokamak design.

Observation of a new type of self-generated current in magnetized plasmas

A tokamak, a torus-shaped nuclear fusion device, needs an electric current in the plasma to produce magnetic field in the poloidal direction for confining fusion plasmas. Plasma current is conventionally generated by electromagnetic induction. However, for a steady-state fusion reactor, minimizing the inductive current is essential to extend the tokamak operating duration. Several non-inductive current drive schemes have been developed for steady-state operations such as radio-frequency waves and neutral beams. However, commercial reactors require minimal use of these external sources to maximize the fusion gain, Q, the ratio of the fusion power to the external power. Apart from these external current drives, a self-generated current, so-called bootstrap current, was predicted theoretically and demonstrated experimentally. Here, we reveal another self-generated current that can exist in a tokamak and this has not yet been discussed by present theories. We report conclusive experimental evidence of this self-generated current observed in the KSTAR tokamak.

Multi-harmonic Rutherford island theory

Rutherford island theory, which governs the nonlinear evolution of tearing modes in tokamak plasmas, is generalized to take into account situations in which the conventional one-harmonic approximation is not valid. The analysis incorporates non-inductive currents driven by radio frequency (RF) electromagnetic waves injected into the plasma. A multi-harmonic tearing mode dispersion relation is derived that takes the form of a nonlinear inhomogeneous matrix eigenvalue problem. The dispersion relation is solved in the so-called two-harmonic approximation, in which only the principal Fourier harmonic of the perturbed magnetic flux and its first overtone are included in the calculation. In the absence of RF current drive, the nonlinear behavior of a tearing mode predicted in the two-harmonic approximation does not differ substantially from that predicted in the one-harmonic approximation. On the other hand, RF current drive that is sufficiently localized in the vicinity of the O-points of the mode's magnetic island chain is capable of triggering bifurcations of the O-points (which is impossible in the one-harmonic approximation). However, the current drive is incapable of triggering bifurcations of the island X-points. This finding is significant because Bardóczi and Evans [Phys. Rev. Lett. 126, 085003 (2021)] recently observed bifurcations of magnetic island chain O-points in the presence of RF current drive in the DIII-D tokamak but did not observe bifurcations of the X-points. Finally, the changes in the topology of the magnetic island flux-surfaces induced by RF current drive are found to facilitate the stabilization of the tearing mode.

Neutral Beam Coupling with Plasma in a Compact Fusion Neutron Source

FNS-ST is a fusion neutron source project based on a spherical tokamak (R/a = 0.5 m/0.3 m) with a steady-state neutron generation of ~1018 n/s. Neutral beam injection (NBI) is supposed to maintain steady-state operation, non-inductive current drive and neutron production in FNS-ST plasma. In a low aspect ratio device, the toroidal magnetic field shape is not optimal for fast ions confinement in plasma, and the toroidal effects are more pronounced compared to the conventional tokamak design (with R/a &gt; 2.5). The neutral beam production and the tokamak plasma response to NBI were efficiently modeled by a specialized beam-plasma software package BTR-BTOR, which allowed fast optimization of the neutral beam transport and evolution within the injector unit, as well as the parametric study of NBI induced effects in plasma. The “Lite neutral beam model” (LNB) implements a statistical beam description in 6-dimensional phase space (106–1010 particles), and the beam particle conversions are organized as a data flow pipeline. This parametric study of FNS-ST tokamak is focused on the beam-plasma coupling issue. The main result of the study is a method to achieve steady-state current drive and fusion controllability in beam-driven toroidal plasmas. LNB methods can be also applied to NBI design for conventional tokamaks.

Effective current drive in the pedestal region of high-confinement tokamak plasma using electron cyclotron waves

Large-amplitude edge-localized modes (ELMs) are a major concern in the development of tokamak fusion energy. According to the peeling–ballooning theoretical model, effective current drive in the pedestal region of high-confinement (H-mode) tokamak plasmas can alter the current profile in this region and hence affect ELM instabilities. Using ray-tracing and Fokker–Planck quasilinear codes, effective and localized current drive by electron cyclotron (EC) waves in the H-mode pedestal region is studied in this paper. Numerical investigations are performed under present-day medium-sized tokamaks with an aspect ratio (A = R/a) covering a typical range of [2.7, 4.0]. Localized non-inductive current can be generated effectively by the Ohkawa mechanism of EC waves in the low-electron collisionality H-mode pedestal region. The Ohkawa mechanism-dominated current drive (OKCD) is much more effective than traditional electron cyclotron current drive (ECCD), which is the Fisch–Boozer mechanism-dominated one. Under 1–2 MW EC power, the current density of both co-OKCD and counter-OKCD has the same magnitude as edge bootstrap current density, which is calculated using the Sauter model. However, edge ECCD is much smaller than the bootstrap current. Consequently, the use of edge ECCD to control ELM failed in experiments, and only the EC heating can affect ELM behavior. Simulations on OKCD are performed in the H-mode pedestal region of realistic DIII-D tokamak plasma, and the results also support the above conclusions. The dependence of effective edge OKCD on the ratio of magnetic field and EC frequency is also found in tokamaks with a typical aspect ratio range. This work opens up a new possible field for active control of ELMs (triggering or suppressing an ELM) by edge current drive in present-day medium-sized tokamaks.

Accurate Current Drive Modeling on EAST

A nonlinear transformer model is used for describing the tokamak discharge with fraction of noninductive current. The nonlinear model accurately identifies the tokamak as a toroidal transformer by a voltage balance equation. In order to verify the feasibility of the application of the nonlinear model in the experimental advanced superconducting tokamak (EAST), the voltage balance equation is checked for the different scenarios and applied to calculate and analyze the noninductive current in shots #73999 and #66740, respectively. Calculation results indicate that the #73999 shot H-mode plasma discharge is full noninductive current drive and occupies approximately 99% of plasma current. It is a good agreement with the simulation result (98%) by the integrated modeling under the framework OMFIT with equilibrium code EAST equilibrium fitting (EFIT), and transport codes TGYRO and ONETWO in EAST. For the #66740 shot, it is nearly full noninductive current drive.

Identification of kink instability in 3D helical flux ropes at VEST

Local helicity injection (LHI) is a non-inductive startup and current drive method via Taylor relaxation for the spherical torus. In achieving Taylor relaxation, it has been suggested that kink instability in 3D helical flux ropes plays an important role. However, the role and occurrence of kink instability during LHI have yet to be validated. Experimentally, determining the kink mode in a flux rope relies on measuring internal information using a probe. However, for LHI, the 3D geometry complicates this measurement process. Here, we propose a new approach for determining the kink modes of 3D helical flux ropes without any internal probe measurements. It is confirmed by this approach that flux ropes exhibit two different kink modes. With increasing plasma current in the flux ropes, a transition from the coherent internal kink mode to the external kink mode is observed. Kink mode properties such as rotating frequency calculated from the kink theory agree well with the magnetic signature driven by the kink mode. During the LHI experiment in the versatile experiment spherical torus, three distinguishable phases are confirmed by the approach, consistent with NIMROD simulation. Before driving the toroidal plasma current, the external kink mode is observed for 3D helical flux ropes. As the toroidal plasma current increases, the external kink mode disappears while generating broadband internal modes instead of coherent internal kink of flux ropes. Decoupling between the toroidal plasma and flux rope results in both decay of toroidal plasma current and re-appearance of the external kink mode in the flux ropes.

Current drive experiments in SST1 tokamak with lower hybrid waves

The steadystate superconducting tokamak (SST1) is aimed to demonstrate long pulse plasma discharges employing non-inductive current drive by means of lower hybrid current drive (LHCD) system. The major and minor radius of the machine is 1.1 m and 0.2 m, respectively. The LHCD system for SST1 comprises of klystrons, each rated for 0.5 MW-CW rf power at a frequency of 3.7 GHz. The grill antenna comprises of two rows, each row accommodating 32 waveguide elements. Electron cyclotron resonance breakdown assisted Ohmic plasma is formed in SST1 to overcome the issues associated with low loop voltage start-ups. With recent modifications in the poloidal coils configuration, even with narrow EC pulse (∼50 ms), good repeatable and consistent Ohmic plasmas could be produced which helped in carrying out LHCD current drive experiments on SST1. These experiments demonstrated both fully as well as partially driven non-inductive plasma current in SST1 tokamak. Discharges with zero loop voltages were obtained. The interaction of lower hybrid waves with plasma and generation of suprathermal electrons could be established using energy spectra measured by CdTe detectors. Various other signatures like drop in loop voltages, negative loop voltages, spikes in hard x-rays and increase in second harmonic ECE signal, further confirmed the current drive by LHW’s. The beneficial effect of LHW’s in suppressing hard x-rays was also demonstrated in these experiments. The longest discharge of ∼650 ms could be obtained in SST1 with the help of LHW’s. In this paper, the experimental results obtained with LHCD experiments on SST1 is reported and discussed in more details.

A 1D quasilinear equation describing the current drive excitation by helicons in a tokamak plasma

A quasilinear equation which allows describing evolution of the electron distribution function and generation of non-inductive currents by helicons is obtained. It is shown that in the analyzed case the Fokker-Planck equation can be approximated by a one-dimensional equation in the longitudinal electron velocity space with a diffusion coefficient proportional to the helicon power absorbed by electrons due to Landau damping. Keywords: tokamak, current drive, helicon, quasilinear diffusion

Одномерное квазилинейное уравнение для описания генерации токов увлечения в плазме токамака геликонами

A quasi-linear equation which allows describing evolution of electron distribution function and generation of non-inductive currents by helicons is obtained. It is shown that in the analysed case the Fokker-Planck equation can be approximated by a one-dimensional equation in the longitudinal electron velocity space with a diffusion coefficient proportional to the helicon power absorbed by electrons due to Landau damping.

Design of Cooling Water System (CWS) for CRAFT Lower Hybrid Current Driven (LHCD) System

Lower hybrid current driven (LHCD) system, as the most efficient non-inductive current drive method in the tokamak, is an integral part of the Comprehensive Research Facility for Fusion Technology (CRAFT). The cooling water system (CWS) is necessary to be designed to remove the thermal power generated by the clients of the LHCD system so that system could operate safely. Therefore, a thermal hydraulic model is developed by AFT Fathom aimed at investigating thermal hydraulic behavior of the system under the normal operation. According to the calculation results, the CWS can provide required pressure, temperature and flow rate to address the client requirements of LHCD system. The study could provide a design reference for the construction of cooling water system for LHCD system and other CRAFT subsystems.

Computational studies on fast wave current drive in high beta SST-1 and SST-2 plasmas

The non-inductive current drive in low toroidal beta (β) steady-state superconducting tokamak-1 (SST-1) is obtained using slow waves or Lower Hybrid waves (LHW). If β of the plasma is increased either by increasing the plasma parameters (density/temperature) or by decreasing the toroidal magnetic field, the non-inductive current drive may be supported by Fast Waves (FWs), as wave accessibility is limited for slow waves at these high β plasmas. Non-inductive current drive using FWs is explored for these high β plasmas to identify the optimized parameters (operating frequency and parallel refractive index (n||)) of the fast wave. The studies indicate that FWs may be launched with an operating frequency of 0.5 GHz RF source and n|| around 4 in SST-1 plasma. The GENRAY code is employed to carry out detailed studies on Fast Wave Current Drive (FWCD) analysis using ray tracing techniques. Computational studies on FWCD are carried out where central electron density (ne0) and central electron temperature (Te0) of 6×1019m−3 and 3 keV respectively is assumed. The FWCD displays central depositions of RF power and is attributed to fast radial propagation compared to the parallel propagation along the toroidal magnetic field, resulting in ray transits with large number of reflections. The off-axis current drive in SST-1 plasma requires reduced radial penetration of the wave and demands unrealistic plasma parameter in terms of β. However, for SST-2 plasmas, having enhanced plasma parameters (βN∼1.8), it is found that if FWs are launched with n||∼3 and f∼1.5 GHz, the radial penetration of the wave is impeded and single pass absorption in toroidal direction is observed, thus providing conducive condition for off axis current drive. It indicates that FWs would drive plasma current of 11 kA/MW at ρ∼0.7 for ne0∼2.2×1020m−3 and Te0∼7.7 keV values.

Progress of Experimental Studies in the HL-2A Tokamak

During the last several years, the HL-2A experiment has made significant progress in the following areas: (1) lower-hybrid wave (LHW) heating and current drive, (2) plasma confinement and turbulent transport, (3) magnetohydrodynamic (MHD) instabilities and energetic particle physics and (4) H-mode and edge localized mode (ELM) control. The results show that the LHW system working in the co-current mode can reach higher driving efficiency and full non-inductive lower-hybrid current drive (LHCD) has been achieved. The intrinsic poloidal torque characterized by the divergence of the residual stress is deduced from synthesis for the first time. The dynamics of spectral symmetry breaking in drift wave turbulence is in good agreement with the development of the poloidal torque to drive the edge poloidal flow. The influence of the cross-phase dynamics on turbulent stress was also investigated. The ion internal transport barrier has been observed in the NBI-heated plasma, and inside the barrier the ion thermal transport is reduced to the neoclassical level. Besides, micro-turbulence is modulated by the rotation frequency of the magnetic island, and this modulation effect is related to a critical island width. Strong E × B shear is found at the island boundary. Three kinds of axisymmetric modes, beta-induced Alfven eigenmode (BAE), toroidal Alfven eigenmode (TAE) and the ellipticity-induced Alfven eigenmode (EAE), are found to be driven unstable by nonlinear mode coupling between Alfven eigenmodes and tearing mode which is well explained by the nonlinear gyrokinetic theory. The fishbone and tearing modes were actively controlled by the electron cyclotron resonance heating (ECRH). The dynamics of the edge plasma flows and turbulence during the L–I–H transition have been dedicatedly investigated. The geodesic acoustic mode (GAM) and limit cycle oscillation (LCO) coexist for a short time and disappear in the H-mode plasma with the increasing of E × B shear flow before the I–H transition, which plays an important role in the turbulence suppression. Different techniques, such as LHW, ECRH, resonant magnetic perturbation (RMP), and impurity seeding by the laser blow-off (LBO) and supersonic molecular beam injection (SMBI), have been successfully applied to control the large ELMs. It has been found that pedestal turbulence enhancement might be responsible for the observed mitigation effect.