Non-inductive Start-up Research Articles

Efficient ECCD non-inductive plasma current start-up, ramp-up, and sustainment for an ST fusion reactor

The elimination of the need for an Ohmic heating solenoid may be the most impactful design driver for the realization of economical compact fusion tokamak reactor systems. However, this would require fully non-inductive start-up and current ramp-up from zero plasma current and low electron temperature of sub-keV to the full plasma current of ∼10–15 MA at 20–30 keV electron temperature. To address this challenge, an efficient solenoid-free start-up and ramp-up scenario utilizing a low-field-side-launched extraordinary mode at the fundamental electron cyclotron harmonic frequency (X–I) is proposed, which has more than two orders of magnitude higher electron cyclotron current drive (ECCD) efficiency than the conventional ECCD for the sub-keV start-up regime. A time dependent model was developed to simulate the start-up scenarios. For the Spherical Tokamak Advanced Reactor (STAR) (Menard et al 2023 Next-Step Low-Aspect-Ratio Tokamak Design Studies (IAEA)), it was found that to fully non-inductively ramp-up to 15 MA, it would take about 25 MW of EC power at 170 GHz. Because of the relatively large plasma volume of STAR, radiation losses must be considered. It is important to make sure that high Z impurities are kept sufficiently low during the early current start-up phase where the temperature is sub-keV range. Since the initial current ramp up takes place at a factor of ten lower density compared to the sustained regimes, it is important to transition into a higher bootstrap fraction discharge at lower density to minimize the ECCD power requirement during the densification. For the sustainment phase an array of eight gyrotron launchers with a total of about 60 MW of fundamental O-mode was found to be sufficient to provide the required axis-peaked external current drive. High efficiencies between 19–57 kA MW−1 were found with optimal aiming, and these were resilient to small changes in aiming angles and density and temperature profiles.

Multi-harmonic electron cyclotron heating and current drive scenarios for non-inductive start-up and ramp-up in high field ST-40 spherical tokamak

Non-inductive start-up and ramp-up is an important topic for spherical tokamak reactor design as the central solenoid implementation is highly restrictive particularly for the low-aspect-ratio tokamak configuration. In the high field spherical tokamak (ST), ST-40 with B T0 ⩽ 3 T, a preparation is underway for high power ECH and ECCD current start-up/ramp-up experiments utilizing two MW-class 140/105 GHz gyrotrons. Here, we explored various ECH/ECCD scenarios for a low-field-side (LFS) launch-angle steerable waveguide launcher placed near the mid-plane region. Due to the large toroidal field variation of ST configuration, multiple cyclotron harmonic resonance layers could exist within the plasma. In this start-up and ramp-up regime, both fundamental and second harmonic ECH resonances must be considered. We find that even with the presence of X-II resonance layer in the plasma, an efficient X-I ECH and ECCD regime can be accessed for the low electron temperature T e0 as low as 200 eV which is a typical starting temperature of ECH heated plasmas in an open-field-line configuration. The presence of X-II resonance could become significant at higher T e0 as X-II absorption increases with T e0 which could reduce the current ramp-up efficiency as the power reaching X-I is reduced. Finally for the pure X-I regime where the 2Ωe resonance is moved outside the plasma with B T0 ∼ 3.4 T, we find that it is possible to reach the full current of I p ∼ 1 MA fully non-inductively with the ECH power of ∼1 MW at n e0 ∼ 1.0 × 1019 m−3 using 105 GHz frequency gyrotron. By reducing the outer limiter position R L ∼ 78 cm to 70 cm, the pure X-I regime is recovered at the rated ST-40 magnetic field of B T0 ∼ 3.0 T. This X-I regime is accessible with a relatively broad range of launched n II or the launching angles. A survey of X-mode X-II ECH and ECCD at higher density regimes is also shown for completeness.

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.

Development of equilibrium fitting code using finite element method in versatile experiment spherical torus

An equilibrium fitting code using finite element method is developed for various plasma studies, especially on non-inductive start-up and ECH heating & current drive experiments in Versatile Experiment Spherical Torus. The magnetic diagnostics are applied to provide the basic equilibrium fitting constraints. As severe eddy currents are induced by not only the PF coils but also the plasma current itself, the eddy current component from the magnetic signals is dealt with the vessel wall model in the eddy current solver and carefully compensated, which is essential for accurate equilibrium fitting. The fast camera image analysis method is adopted and the limiter current monitoring system is developed for validating the results of the plasma position and shape from the code. The equilibrium fitting code is applied to the limited and diverted plasma configuration in VEST and the time evolution of the plasma position and shape and even averaged magnetohydrodynamic parameters such as the edge safety factor and the internal inductance are extracted. Additionally, the equilibrium fitting code is used to analyze the non-inductive plasma start-up experiment using the direct current helicity injection in VEST. Especially, it is the first time to apply the finite element method to direct current helicity injection experiment and ECH heated plasmas which have significant amount of plasma current flowing along the open field line.

Regarding the optimization of O1-mode ECRH and the feasibility of EBW startup on NSTX-U

Recently published scenarios for fully non-inductive startup and operation on the National Spherical Torus eXperiment Upgrade (NSTX-U) (Menard et al 2012 Nucl. Fusion 52 083015) show Electron Cyclotron Resonance Heating (ECRH) as an important component in preparing a target plasma for efficient High Harmonic Fast Wave and Neutral Beam heating. The modeling of the propagation and absorption of EC waves in the evolving plasma is required to define the most effective window of operation, and to optimize the launcher geometry for maximal heating and current drive during this window. Here, we extend a previous optimization of O1-mode ECRH on NSTX-U to account for the full time-dependent performance of the ECRH using simulations performed with TRANSP. We find that the evolution of the density profile has a prominent role in the optimization by defining the time window of operation, which in certain cases may be a more important metric to compare launcher performance than the average power absorption. This feature cannot be captured by analysis on static profiles, and should be accounted for when optimizing ECRH on any device that operates near the cutoff density. Additionally, the utility of the electron Bernstein wave (EBW) in driving current and generating closed flux surfaces in the early startup phase has been demonstrated on a number of devices. Using standalone GENRAY simulations, we find that efficient EBW current drive is possible on NSTX-U if the injection angle is shifted below the midplane and aimed towards the top half of the vacuum vessel. However, collisional damping of the EBW is projected to be significant, in some cases accounting for up to 97% of the absorbed EBW power.

Simulations towards the achievement of non-inductive current ramp-up and sustainment in the National Spherical Torus Experiment Upgrade

One of the goals of the National Spherical Torus Experiment Upgrade (NSTX-U) (Menard et al 2012 Nucl. Fusion 52 083015) is the demonstration of fully non-inductive start-up, current ramp-up and sustainment. This work discusses predictive simulations where the available heating and current drive systems are combined to maximize the non-inductive current and minimize the solenoidal contribution. Radio-frequency waves at harmonics higher than the ion cyclotron resonance (high-harmonic fast waves (HHFW)) and neutral beam injection are used to ramp the plasma current non-inductively starting from an initial Ohmic plasma. An interesting synergy is observed in the simulations between the HHFW and electron cyclotron (EC) wave heating. Time-dependent simulations indicate that, depending on the phasing of the HHFW antenna, EC wave heating can significantly increase the effectiveness of the radio-frequency power, by heating the electrons and increasing the current drive efficiency, thus relaxing the requirements on the level of HHFW power that needs to be absorbed in the core plasma to drive the same amount of fast-wave current.

Simulated flux-rope evolution during non-inductive startup in Pegasus

Magnetic flux ropes produced during non-inductive startup of the Pegasus Toroidal Experiment (Eidietis et al 2007 J. Fusion Energy 26 43) are simulated with nonlinear magnetohydrodynamic (MHD) and two-fluid plasma models. A single injector is represented by a localized source-density for magnetic helicity and thermal energy. Results show development of a hollow tokamak-like profile from a sequence of co-helicity merging events that release flux-rope rings from the driven flux rope. Accumulation of poloidal flux over many events redirects the driven flux rope so that its path traces a toroidal surface. Evidence of a quasi-separatrix layer is found from analysis of the squashing-degree parameter during a ring formation event in an MHD simulation. The layer bifurcates twice, once between non-reconnecting passes and again between reconnecting passes of the driven flux rope. Chaotic scattering during ring formation is also apparent from the distribution of field-line lengths. Correlation of flow-velocity and magnetic-field fluctuations, an MHD dynamo-like effect, concentrates symmetric poloidal flux during ring formation.

Physics and Experimental Results of KSTAR ECH

Since the first plasma in Korea Superconducting Tokamak Advanced Research (KSTAR), the electron cyclotron heating (ECH) system has been an essential tool for ECH-assisted start-up using second-harmonic 84-GHz and 110-GHz ECH and for experimental studies of other physics issues such as edge-localized mode control, rotation control, sawtooth control, tearing mode control, and core impurity control for long-pulse discharge and noninductive start-up. The loop voltage in KSTAR is limited by superconducting ohmic coils and a thick vacuum vessel. The ECH-assisted start-up was useful to overcome burn-through during the ramp-up phase with limited loop voltage by the electron cyclotron beam injection before or after the onset of the inductive loop voltage to reduce resistive power consumption. Later, the 170-GHz ECH system is installed as a main electron heating and local current drive for the control of magnetohydrodynamic modes such as sawteeth and neoclassical tearing modes. The 170-GHz gyrotron is an ITER preprototype gyrotron developed by the Japan Atomic Energy Agency. Second-harmonic 170-GHz ECH-assisted start-up was also attempted with a raised toroidal magnetic field of 3 T in the 2011 KSTAR campaign, and flux saving in the ramp-up phase was observed. This lecture describes the physics issues and experimental results of the KSTAR ECH system. The present status and an upgrade plan of the KSTAR ECH system is also described.

Resistive magnetohydrodynamic simulations of helicity-injected startup plasmas in National Spherical Torus eXperiment

The generation of helicity-injected startup plasmas in National Spherical Torus eXperiment (NSTX), including flux surface closure, is studied using resistive-magnetohydrodynamic simulations with plasma flows, currents, ohmic heating and anisotropic thermal conduction. An injection-voltage pulse shape is used that separates the injection and closure phases allowing elucidation of the physics. The formation of an X-point near the helicity-injection gap is triggered as the injector voltage drops to zero. Near the forming X-point, magnetic pressure due to toroidal field entrained in the E × B plasma flow from the helicity-injection gap drops, allowing resistive magnetic reconnection even though the total injected current is almost constant. Where appropriate, the simulations are compared with Transient Coaxial Helicity Injection experiments in the NSTX spherical tokamak, which have demonstrated the formation of a promising candidate for non-inductive startup plasmas [Raman et al., Phys. Rev. Lett. 90, 075005 (2003)].

Overview of physics results from NSTX

In the last two experimental campaigns, the low aspect ratio NSTX has explored physics issues critical to both toroidal confinement physics and ITER. Experiments have made extensive use of lithium coatings for wall conditioning, correction of non-axisymmetric field errors and control of n = 1 resistive wall modes (RWMs) to produce high-performance neutral-beam heated discharges extending to 1.7 s in duration with non-inductive current fractions up to 0.7. The RWM control coils have been used to trigger repetitive ELMs with high reliability, and they have also contributed to an improved understanding of both neoclassical tearing mode and RWM stabilization physics, including the interplay between rotation and kinetic effects on stability. High harmonic fast wave (HHFW) heating has produced plasmas with central electron temperatures exceeding 6 keV. The HHFW heating was used to show that there was a 20–40% higher power threshold for the L–H transition for helium than for deuterium plasmas. A new diagnostic showed a depletion of the fast-ion density profile over a broad spatial region as a result of toroidicity-induced Alfvén eigenmodes (TAEs) and energetic-particle modes (EPMs) bursts. In addition, it was observed that other modes (e.g. global Alfvén eigenmodes) can trigger TAE and EPM bursts, suggesting that fast ions are redistributed by high-frequency AEs. The momentum pinch velocity determined by a perturbative technique decreased as the collisionality was reduced, although the pinch to diffusion ratio, Vpinch/χϕ, remained approximately constant. The mechanisms of deuterium retention by graphite and lithium-coated graphite plasma-facing components have been investigated. To reduce divertor heat flux, a novel divertor configuration, the ‘snowflake’ divertor, was tested in NSTX and many beneficial aspects were found. A reduction in the required central solenoid flux has been realized in NSTX when discharges initiated by coaxial helicity injection were ramped in current using induction. The resulting plasmas have characteristics needed to meet the objectives of the non-inductive start-up and ramp-up program of NSTX.

Noninductive plasma initiation and startup in the DIII-D tokamak

Noninductive plasma startup with second harmonic electron cyclotron heating has been studied in DIII-D. Plasma currents up to 33 kA have been obtained in this phase. The maximum current obtained was primarily limited by the experimental time and does not necessarily represent the highest achievable current. The dominant physical mechanism for the observed current is the Pfirsch–Schlüter current on open field lines. Closed flux surfaces have been observed, but in a rather limited region, compared with previous experiments in spherical tokamaks. This method of plasma initiation and initial current ramp might be used in future burning plasma devices in conjunction with other current drive techniques to provide a fully noninductive startup.

Transient process of a spherical tokamak plasma startup by electron cyclotron waves

Non-inductive startup by a 2.45 GHz microwave through electron cyclotron resonance heating (ECRH) has been studied in the SUNIST spherical tokamak (R/a: 0.3 m/0.23 m; BT0: < 0.15 T). Two types of discharges, with quite different transient characteristics, were found. The transient processes of discharges are experimentally investigated by scanning the radial position of the ECR layer, the vertical magnetic field and the microwave power. Time domain analysis of microwave reflections and Hα emissions reveal a process dominated by the combination of ionization, loss along the open field lines and the gradient B drift. It is shown that the transient character is insensitive to the experimental parameters except the pressure of filled gas. This result confirms that the control of gas puffing is of great importance for starting up a spherical tokamak by ECRH using low-frequency microwaves.

Electron Density Measurements of Non-Inductive Start-Up Plasmas in the TST-2 Spherical Tokamak

A new 50GHz microwave interferometer was installed, and non-inductively sustained spherical tokamak plasmas in the TST-2 device were measured. The line integrated electron densities (nels) on five chords were compared with visible CCD camera image and equilibrium analysis. It is concluded that the high-density region has a C-like shape along the outboard boundary. This shape suggests that the electron density is not constant on the magnetic flux surface. c

Plasma Current Sustainment by RF Power in ECH Start-up Plasma in the TST-2 Spherical Tokamak

In the TST-2 spherical tokamak, a non-inductive start-up plasma produced by electron cyclotron heating (ECH) at 2.45 GHz (3-5 kW) is sustained by RF wave heating at 21 MHz (10-24 kW). A current jump, which occurs at a certain time when only ECH is used, can be induced earlier by the RF heating, and a constant plasma current is sustained even after the ECH power is turned off. The plasma current is sustained in both hydrogen and deuterium plasmas, as long as the RF power exceeds a threshold level, which is lower for deuterium plasma. The maximum sustainment time by the the RF power alone was 40 ms, limited by the RF pulse length. According to CCD camera images of the total plasma light, the major radii are almost the same, but the RF-sustained plasma has a more peaked profile, compared with the distribution of an ECH-sustained plasma.

NSTX Plasma Start-Up Using Transient Coaxial Helicity Injection

Future toroidal magnetic confinement fusion plasma devices such as the Component Test Facility (CTF) require non-inductive toroidal current drive. A new method of non-inductive startup, referred to as transient coaxial helicity injection (Transient CHI), has been developed on the Helicity Injected Torus (HIT-II) experiment and the National Spherical Torus Experiment NSTX). In this method, plasma current is produced by discharging a capacitor bank between coaxial electrodes in the presence of toroidal and poloidal magnetic fields chosen such that the plasma rapidly expands into the chamber. When the injected current is rapidly decreased, magnetic reconnection occurs near the injection electrodes with the toroidal plasma current forming closed flux surfaces. In NSTX, transient CHI has demonstrated closed-flux current generation of up to 160 kA, without the use of a central solenoid. Detailed experimental measurements made on NSTX include fast time-scale visible imaging of the entire process.

A current profile model for magnetic analysis of the start-up phase of toroidal plasmas driven by electron cyclotron heating and current drive

An estimation model of plasma current density distribution for the start-up phase of toroidal plasmas generated by electron cyclotron heating (ECH) in the low aspect ratio torus experiment device is presented. The model assumes a power law parabolic current profile having seven fitting parameters. Its position, extent and broadness (or steepness) are fitted by adjusting these parameters to the observed magnetic flux signals. The adequacy of the model has been examined and confirmed by comparisons of the reconstructed current profiles and the resultant poloidal flux surfaces with the plasma images at visible light range at various stages of start-up discharges, including both the initial open field phase, the subsequent closed field phase, the current decay phase after ECH is turned off and also by a current-profile limiting experiment. This method may be useful for the study of non-inductive start-up experiments by ECH, where there is no appropriate MHD constraint on the current distribution as that in the full tokamak discharge plasmas.

Solenoid-free Plasma Start-up in HIT-II and NSTX using Transient CHI

A new method of non-inductive startup, referred to as transient coaxial helicity injection (CHI), has been successfully developed on the HIT-II experiment. In this method, a plasma current is rapidly produced by discharging a capacitor bank between coaxial electrodes in the presence of toroidal and poloidal magnetic fields. The initial poloidal field configuration is chosen such that the plasma rapidly expands into the chamber. When the injected current is rapidly decreased, magnetic reconnection occurs near the injection electrodes, with the toroidal plasma current forming closed flux surfaces. On HIT-II, CHI-started plasmas outperform discharges initiated by induction alone and consume fewer volt-seconds. The method has now been successfully applied on NSTX for an unambiguous proof-of-principle demonstration of closed-flux current generation without the use of the central solenoid.

Initial Experiments at High Normalized Current in the Pegasus Toroidal Experiment

Initial experiments in the Pegasus ST exhibited an operational limit of I p ∼ I tf at A ∼ 1.15. With new power supplies, discharge control has greatly increased. Equilibria and stability modeling of the high I p/I tf regime has produced stable equilibria with I p/I tf up to 2 while values of I p/I tf approaching 3 show a potential upper bound on stability. A new operating regime has been accessed in recent experiments with the addition of electrostatic plasma sources. Using the sources as a pre-ionization technique, I p/I tf values of 1.5 have been accessed at very low toroidal fields. Using the sources as a means for non-inductive startup, values of I p/I tf up to 2.3 are attained. The MHD activity in these discharges is characteristically different than ohmic discharges, suggesting the new current sources are modifying the current profile to allow stable discharge evolution at minimal toroidal field.

Progress towards high performance plasmas in the National Spherical Torus Experiment (NSTX)

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ∼40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfvén eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

Initial Plasma Startup Test on SUNIST Spherical Tokamak

The goal of the Sino-United Spherical Tokamak (SUNIST) at Tsinghua University is to extend the understanding of toroidal plasma physics at a low aspect ratio (R/a ≈ 1.3) and to demonstrate a maintainable target plasma by non-inductive startup. The SUNIST device is designed to operate with up to 13 kA of ohmic heating field current, and to 0.15 T of toroidal field at 10 kA of discharge current. All of the poloidal fields can provide 30 mVs of Volt-seconds transformer. Experimental results of plasma startup show that SUNIST has remarkable characteristics of high ramp rate (dIP/dt ≈ 50 MA/s), high normalized current IN of about 2.8 (IN = (IP/aBT), and high-efficiency (IP/IROD ≈ 0.4) production of plasma current while operating at a low toroidal field. Major disruption phenomena have not been observed from magnetic diagnostics of all testing shots. Initial discharges with 52 kA of plasma current (exceeding the designed value of 50 kA), 2 ms of pulse length and 50 MA/s of ramp rate have been achieved easily with pre-ionized filament.