Table of contents

Over the past two years noteworthy experimental and theoretical progress has been made towards the top-level scientific question for the US programme on heavy-ion-fusion-science and high energy density physics: 'How can heavy-ion beams be compressed to the high intensity required to create high energy density matter and fusion conditions?' New results in transverse and longitudinal beam compression, high-brightness transport and beam acceleration will be reported. Central to this campaign is final beam compression. With a neutralizing plasma, we demonstrated transverse beam compression by an areal factor of over 100 and longitudinal compression by a factor of > 50. We also report on the first demonstration of simultaneous transverse and longitudinal beam compression in plasma. High beam brightness is key to high intensity on target, and detailed experimental and theoretical studies on the effect of secondary electrons on beam brightness degradation are reported. A new accelerator concept for near-term low-cost target heating experiments was invented, and the predicted beam dynamics validated experimentally. We show how these scientific campaigns have created new opportunities for interesting target experiments in the warm dense matter regime. Finally, we summarize progress towards heavy-ion fusion, including the demonstration of a compact driver-size high-brightness ion injector. For all components of our high intensity campaign, the new results have been obtained via tightly coupled efforts in experiments, simulations and theory.

Recent studies at ASDEX Upgrade aim to further characterize and understand the physics of the improved H-mode scenario. The main focus is on the influence of the ramp-up phase of the plasma current and heating on energy confinement and MHD-activity during the subsequent flat-top phase. Depending on the ramp-up scenario two different stationary plasmas can be generated, which show different equilibrated current profiles, although external control parameters are the same in the flat-top phase. The difference of the current profiles in the flat-top phase seems to be due to different MHD-modes. These MHD-modes set in during relaxation of the current profile, which itself depends on the ramp-up scenario. Also the stored energy is different in the two cases as is the peaking of the temperature profiles.

Three mechanisms seem to play a role in linking the observed changes in MHD-behaviour and current profile to the changes of the kinetic profiles: the increased transport due to the MHD-modes themselves, the variation of the ratio of magnetic shear s to safety factor q, which modifies the critical temperature gradient length for the onset of ion temperature gradients, and effects on the H-mode pedestal pressure.

The pseudosymmetry principle being used for hot plasma confinement magnetic systems design could provide the realization of all the ability of magnetic confinement for reactor construction. In particular, it enables hot plasma confinement at steady state operational mode with a 'tokamak' level of confinement time and a high average β value (plasma pressure to 'vacuum' magnetic field pressure ratio). High β operation provides the possibility of thinking of advanced fuels (D–³He, D–D) use.

It is proposed to check the main idea and the results of a preliminary theoretical study experimentally in the frame of the EPSILON project—closed mirror magnetic trap with extremely small rotational transform and with poloidal pseudosymmetry. The EPSILON system consists of two (or more) multimirror traps, closed with curvilinear equilibrium elements. Mirror field, MHD stability and plasma cleaning are provided by axially symmetric divertors with zero magnetic field at the separatrix displaced along the multimirror traps.

We develop a multiple interface variational model, comprising multiple Taylor-relaxed plasma regions separated by ideal MHD barriers. The magnetic field in each region is Beltrami, ∇ × B = μB, and the pressure constant. Between regions the pressure, field strength, and rotational transform may have step changes at the ideal barrier. A principle motivation is the development of a mathematically rigorous ideal MHD model to describe intrinsically 3D equilibria, with nonzero internal pressure, using robust KAM surfaces as the barriers. This article chiefly addresses whether the stability of two interface configurations with continuous rotational transform, but vanishing interface separation, is different from the stability of a single interface configuration with jump in the rotational transform. To make the problem analytically tractable, we derive the equilibria and stability of a multi-interface plasma in a periodic cylinder, generalizing the cylindrical treatment of Kaiser and Uecker (2004 Q. J. Mech. Appl. Math.57 1–17). For two interfaces with no jump in rotational transform, we show that one eigenmode has in-phase interface displacements, and an eigenvalue that converges to the single barrier case in the limit of vanishing interface width. The complementary eigenmode is out-of-phase, and highly unstable. Physically, the unstable eigenmode is driven by the parallel current, and caused by the high shear required to match the different rotational transform on each interface. In the limit that the interface separation vanishes, the shear and parallel current density become infinite, and the parallel current between the interfaces nonzero. Surfaces with out-of-phase displacements will then collide, unless the amplitude goes to zero as the interface separation goes to zero. These results suggest the hypothesis that KAM barriers with different irrational rotational transform on either side may be allowable without violating nonlinear stability.

The pellet pacing ELM control concept developed at the mid-sized tokamak ASDEX Upgrade is considered for the larger machines JET and ITER as well. By driving up the ELM frequency, the ELM size can be reduced and eventually suppressed below a dangerous level. An according pellet injection system for JET is under construction; the ITER design includes one as well. However, it is still to be proven whether the concept will also work for bigger machine sizes. A step forward in this mission was achieved by re-analysing previous JET experiments dedicated to pellet particle fuelling. Although the experiments were performed in a parameter regime unsuitable for pellet ELM pacing both with respect to pellet frequency and size they demonstrate that prompt triggering of ELMs takes place at JET as well. Prompt triggering, an indispensable premise for useful pacing, obviously can be achieved already with a pellet size sufficiently small to avoid unbearable side-effects by particle fuelling. An ELM released by the local perturbation imposed by the ablating pellet in the plasma edge region is detected when only a minor fraction (less than 1%) of the fuelling size pellets (containing about 4 × 10²¹ D) was ablated and deposited. It is thus very likely that JET will allow for investigations on the operational technique and the underlying physics of pellet pacing once the new system becomes operational.

The radial electric field in known to be one of the drivers of parallel ion flow in the SOL. It contributes to the ion Pfirsch–Schlüter flow and also determines the 'return parallel flow' that can arise to compensate poloidal E × B drift. It was established recently that 2D fluid codes EDGE2D and SOLPS underestimate the predicted E_r in the SOL compared to experimentally measured values. The underestimate is likely to be related with kinetic effects of parallel transport by heat-carrying electrons. The present work demonstrates that this underestimate can be responsible for the large discrepancy between measured and simulated parallel ion flows in the SOL observed in a number of experiments. Were the radial electric field modelled correctly by the codes, a significant increase in the predicted Mach number of the parallel ion flow could be expected. For the part of the ion flow that depends on the toroidal field direction, this would greatly reduce, or even possibly eliminate, the difference between the experiment and model, as concluded from the comparison between measured and simulated radial electric fields and Mach numbers of the parallel ion flow in JET and ASDEX Upgrade.

Results of active control of neoclassical tearing modes (NTMs) by electron cyclotron current drive (ECCD) in JT-60U are described. Growth of an NTM with poloidal mode number m = 3 and toroidal mode number n = 2 has been suppressed by ECCD inside the sawtooth inversion radius in the co-direction, showing the possibility of the coexistence of sawtooth oscillations and a small-amplitude m/n = 3/2 NTM without large confinement degradation. Stabilization of an m/n = 2/1 NTM by ECCD at the mode rational surface has been demonstrated with a small ratio of the current density driven by the electron cyclotron (EC) wave to the local bootstrap current density (∼ 0.5). In addition, dependence of the stabilization effect on ECCD location has been investigated in detail. It has been found that an m/n = 2/1 NTM can be completely stabilized with the misalignment of the ECCD location less than about half of the full island width, and that the m/n = 2/1 NTM is destabilized with the misalignment comparable to the full island width. Time-dependent, self-consistent simulation of magnetic island evolution using the TOPICS code has shown that the stabilization and destabilization of an m/n = 2/1 NTM are well reproduced with the same set of coefficients of the modified Rutherford equation. The TOPICS simulation has also clarified that EC wave power required for complete stabilization can be significantly reduced by narrowing the ECCD deposition width.

The modified RFX is a very flexible device used for a variety of control schemes for MHD instabilities and for advanced reverse field pinch scenarios. Relative to the previous machine, RFX-mod has a thin Cu shell with vertical field penetration time τ_S, lowered from 450 to 50 ms and shell/plasma proximity from b/a = 1.24–1.1. Toroidal equilibrium is feedback-controlled and new power supplies provide better B_ϕ control. Newly designed graphite tiles protect the vessel from localized power deposition. A mesh of 192 external saddle coils, supervised by a digital feedback system, controls radial fields due to field errors and MHD modes. The paper presents an overview of the very encouraging results obtained using both new and 'standard' advanced operational modes in the current range 0.3–1 MA. A dramatic improvement of plasma performance was obtained by using the saddle coils to cancel all the radial field components, an operation mode dubbed virtual shell (VS). The toroidal voltage was lowered by more than 25% and the pulse length was tripled, up to 7 times the τ_S. Steady-state RFP pulses are now limited only by the applied volt-seconds. The improved magnetic boundary also has an effect on the tearing modes underlying the sustainment of the RFP, whose core amplitude is more than halved. The VS combined with new schemes for the active rotation of the MHD dynamo modes has allowed us to obtain reliable and well-controlled long RFP pulses in the MA current range. This results in a 100% increase in the particle and energy confinement time relative to the previous experiment and opens the possibility of exploring the machine performance in the 2 MA design range.

Coaxial helicity injection (CHI) has been successfully used in the National Spherical Torus Experiment for a demonstration of closed flux current generation without the use of the central solenoid. The favourable properties of the spherical torus (ST) arise from its very small aspect ratio. However, small aspect ratio devices have very restricted space for a substantial central solenoid. Thus methods for initiating the plasma current without relying on induction from a central solenoid are essential for the viability of the ST concept. CHI is a promising candidate for solenoid-free plasma startup in a ST. The method has now produced closed flux current up to 160 kA verifying the high current capability of this method in a large ST built with conventional tokamak components.

A new scenario to control transport of 'cold' α-particles in a helical fusion device by changing poloidal field coil current during plasma discharge is proposed here. A way to enhance radial transport of α-particles in the intermediate energy range is considered which relies on specific features of helical magnetic field. To remove helically trapped, cold α-particles, a β-induced change in the B/B₀ modulation along the particle trajectory is employed. To remove passing particles resonant effects connected with the poloidal field coil currents are used.

The following results are presented from the development and application of TEMPEST, a fully nonlinear (full-f) five-dimensional (3d2v) gyrokinetic continuum edge-plasma code. (1) As a test of the interaction of collisions and parallel streaming, TEMPEST is compared with published analytic and numerical results for endloss of particles confined by combined electrostatic and magnetic wells. Good agreement is found over a wide range of collisionality, confining potential and mirror ratio, and the required velocity space resolution is modest. (2) In a large-aspect-ratio circular geometry, excellent agreement is found for a neoclassical equilibrium with parallel ion flow in the banana regime with zero temperature gradient and radial electric field. (3) The four-dimensional (2d2v) version of the code produces the first self-consistent simulation results of collisionless damping of geodesic acoustic modes and zonal flow (Rosenbluth–Hinton residual) with Boltzmann electrons using a full-f code. The electric field is also found to agree with the standard neoclassical expression for steep density and ion temperature gradients in the plateau regime. In divertor geometry, it is found that the endloss of particles and energy induces parallel flow stronger than the core neoclassical predictions in the SOL.

We report on the resolution of a significant discrepancy between published continuum-code simulations and subsequent global particle-in-cell (PIC) simulations of electron-temperature-gradient (ETG) turbulence. Our investigations, using gyrokinetic δf-PIC- and continuum-code simulations and analytical theory, strongly support the conclusion from the earlier continuum-code simulations that ETG turbulence can drive the electron thermal conductivity χ_e large enough to be significant in some tokamaks. A successful ETG-turbulence benchmark between δf-PIC and continuum codes for ETG turbulence has also been completed. Scans in the magnetic shear show an abrupt transition to a high-χ_e state as the shear is increased from the benchmark value of s = 0.1 to above s = 0.4. When nonadiabatic ions are used, this abrupt transition is absent, and χ_e reaches values consistent with transport analyses of DIII-D, JET, JT60-U and NSTX discharges. The balances of zonal-flow driving and damping terms in late-time quasi-steady phase of ITG turbulence have been unfolded using a new run-time gyrokinetic-simulation diagnostic. The zonal flow level is set by a balance of large driving and damping terms which almost cancel each other. The driving is found to be mostly by the Reynolds stress, while the dissipation is mostly by the linear (transit-time) damping terms. It is also shown that useful zonal-flow-balance information can be obtained with spatially localized samples at as few as four poloidal locations. Real-geometry simulations have been undertaken, using the nonlinear δf-PIC gyrokinetic code SUMMIT/PG3EQ_NC, of the DIII-D 'Cyclone' shot #81499 and of shot #118561, which had broad-wavenumber-range density fluctuation measurements. Real geometry is found to have a significant effect on the transport rates, even though the effect on the linear growth rates is often modest.

Alternatives to the usual picture of advanced tokamak (AT) discharges are those that form when anomalous thermal conductivity and/or resistivity alter plasma current and pressure profiles to achieve stationary characteristics through self-organizing mechanisms where a measure of desired AT features is maintained without external current-profile control. Regimes exhibiting these characteristics are those where the safety factor (q) evolves to a stationary profile with the on-axis and minimum q∼ 1. Operating scenarios with fusion performance exceeding H-mode at the same plasma current and where the inductively driven current density achieves a stationary configuration with either small or nonexisting sawteeth should enhance the performance of ITER and future burning plasma experiments. We present simulation results of anomalous current-profile formation and evolution using theory-based hyper-resistive models. These simulations are stimulated by experimental observations with which we compare and contrast the simulated evolution. We find that the hyper-resistivity is sufficiently strong to modify the current-profile evolution to achieve conditions consistent with experimental observations. Modelling these anomalous effects is important for developing a capability to scale current experiments to future burning plasmas.

The majority of optical diagnostics presently foreseen for ITER will implement in-vessel metallic mirrors as plasma-viewing components. Mirrors are used for the observation of the plasma radiation in a very wide wavelength range: from about 1 nm up to a few mm. In the hostile ITER environment, mirrors are subject to erosion, deposition, particle implantation and other adverse effects which will change their optical properties, affecting the entire performance of the respective diagnostic systems. The Specialists Working Group (SWG) on first mirrors was established under the wings of the International Tokamak Physics Activity (ITPA) Topical Group (TG) on Diagnostics to coordinate and guide the investigations on diagnostic mirrors towards the development of optimal, robust and durable solutions for ITER diagnostic systems. The results of tests of various ITER-candidate mirror materials, performed in Tore-Supra, TEXTOR, DIII-D, TCV, T-10, TRIAM-1M and LHD under various plasma conditions, as well as an overview of laboratory investigations of mirror performance and mirror cleaning techniques are presented in the paper. The current tasks in the R&D of diagnostic mirrors will be addressed.

Recent progress in the RF low activation structural materials R&D road map towards DEMO via the FBR tests (BOR-60, BN-600, BN-800) and the TBM tests in ITER is overviewed. The properties of the RAFMS RUSFER-EK-181 (Fe–12Cr–2W–Ta–V–B–C) and the V–4Ti–4Cr alloys are presented. The next important steps include further studies on the influence of high dose and high-temperature irradiation on the properties of base structural materials and joints. Activation, transmutation and radiation damage of the materials in BN-600 and DEMO-RF (Kurchatov Institute project) neutron spectra are calculated. The results of the application of the internal friction (ultrasonic) non-destructive method to research the DBTT are in good agreement with the results of the destructive impact method. The important influence of boron on the heat resistance of materials and the He concentration level under irradiation are calculated. The new special regimes of the heat treatments of the alloys are suggested to widen the temperature windows of the applications. The results of the BOR-60 examinations of RUSFER-EK-181 (irradiation temperature 320–340 °C and doses up to 15 dpa) are presented. The BN-600 projects for the high dose and high-temperature irradiation tests of manufactured alloys are presented.

The confinement degradation and transport of energetic ions due to Alfvén eigenmodes (AEs) destabilized by negative-ion-based neutral beam injection are quantitatively evaluated for the first time. AEs were observed in JT-60U weak shear plasmas with frequencies sweeping up to 30–50 kHz on a time scale of a few hundred milliseconds as the safety factor at the plasma centre, q(0), decreases from 1.75 to 1.5. Then as q(0) decreases from 1.5 to 1.25 in about 600 ms, the AE frequencies are saturated and the AEs are identified as the TAEs. The measured total neutron emission rate (Sn) in the presence of these AEs is compared with that predicted by classical theory. As a result, the confinement degradation of energetic ions is confirmed and the reduction rate in Sn is largest when AEs are destabilized and are located at the large fast ion pressure gradient region in the plasma core. In order to investigate the energetic ion transport due to these AEs, line-integrated neutron emission profiles measured with a large neutron collimator array are compared with the classical confinement calculations. It is found that energetic ions are transported from the plasma core region to the outer region due to these AEs. Furthermore, changes in the energy distribution of charge exchange energetic neutral particle fluxes suggest that the radial energetic ion transport is due to the resonance interaction between the energetic ions and the AEs.

The characteristics of momentum transport and plasma rotation in the toroidal direction are studied, using near-perpendicular neutral beam injection (PERP-NBI), co tangential and counter (CTR) tangential NBI in JT-60U. Diffusive and non-diffusive terms of momentum transport are evaluated from the transient analysis by using the momentum source modulation. Fast ion losses due to the toroidal field ripple, which locally induces the edge CTR rotation, are used as a novel momentum source. Parameter dependence of these transport coefficients i.e. the toroidal momentum diffusivity (χ_ϕ) and the convection velocity (V_conv), and the relation between momentum and heat diffusivities (χ_i) are investigated in L-mode plasmas systematically. The toroidal momentum diffusivity increases with increasing heating power and decreases with increasing plasma current. The relation of χ_ϕ and χ_i to some non-dimensional parameters is investigated. A clear dependence of χ_ϕ/χ_i on normalized plasma pressure (β_N) is observed. It is also found that toroidal rotation velocity profiles in the case with and without external torque input can be almost reproduced by χ_ϕ and V_conv estimated from the transient momentum transport analysis at low β (β_N < 0.4).

In order to investigate the surface temperature effects on plasma fuel recycling and impurity release from the plasma facing components, plasma discharges have been performed under selected plasma–wall interaction (PWI) conditions in the high-field superconducting tokamak, TRIAM-1M. By moving a water-cooled molybdenum movable limiter (ML) beyond the last closed flux surface, as defined by poloidal limiters, the surface temperature profile on it is varied. Hot spots have been observed on the ML surface in such conditions. The release behaviour of fuel as well as impurity particles from the ML surface has been studied as a function of hot spot temperature (T_hot) by means of wide range spectroscopy (200–1600 nm). A critical T_hot is found to be ∼2100 K above which the emission of both hydrogen and impurity particles enhances significantly. This is indicative of some thermally activated process playing an important role in PWIs between the limiter and the edge plasma. With the rise in hot spot temperature localized PWI at the ML is found to dominate the global recycling even when external fuelling is stopped.

The gyrokinetic full f simulation code ELMFIRE has been used to perform the first global self-consistent computational study of the dynamics of the radial electric field in the presence of turbulence in an experimental toroidal configuration. Parameters have been taken from the FT-2 tokamak. Appreciable differences have been found in comparison with the neo-classical analytic estimate of radial electric field, specially in the outer region with higher level of turbulence. Similar differences have previously been observed in experiments. The analysis included in this work does not support the hypothesis of the neo-classical non-ambipolar transport from finite Larmor-radius effects as a source of discrepancies, but qualitatively supports the theory of eddy viscosity producing those deviations from standard neo-classical theory.

Characteristics of ETG-driven turbulence dominated by zonal flows (ZFs) and nonlinearly generated large scale structures with low toroidal/poloidal wave numbers are investigated by gyrofluid simulations in slab geometry. Main results found in this research are as follows. (i) The electron transport analysis in various parameter regions shows that the electron heat flux was not necessarily enhanced as the electron temperature gradient increases. (ii) The ZFs change the characteristics of turbulence from a 'homogeneous' structure to an 'inhomogeneous' one, in which micro-scale vortices and nonlinearly generated macro-scale vortices coexist in different radial zones, exhibiting a two-scale nature in turbulence. (iii) The fractal dimension is simultaneously reduced in any radial region with an increase in the ratio of ZF energy to that of total fluctuations. Namely, macro-scale vortices, which are the origin of low fractal dimension and play a role in saturating ZFs, survive in the system without causing large thermal transport by adjusting the phase relation.

The long-lived saturated zonal flow (ZF) structures, spontaneously generated by drift-wave (DW) turbulence, are viewed as generators of nonlinear equilibria. Here, we derive a nonlinear evolution equation for the zonal response, valid for arbitrary wavelengths, whose time-asymptotic behaviour corresponds to the nonlinear equilibrium. Its corresponding stability, meanwhile, determines the nature of the ZF instability and the nonlinear up-shift of turbulent transport thresholds. On a shorter-time scale, the temporal evolution of the zonal response describes the DW–ZF generation and the regulation of the DW intensity by the ZFs. As an application, we present a kinetic stability analysis of zonal structures in the ion temperature gradient driven DW turbulence case as well as the corresponding collisionless zonal dissipation rates.

The concept for a compact DEMO reactor named 'SlimCS' is presented. Distinctive features of the concept are low aspect ratio (A = 2.6) and use of a reduced-size centre solenoid (CS) which has the function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field coil system which contributes to reducing the weight and perhaps lessening the construction cost. Low-A has merits of vertical stability for high elongation (κ) and high normalized beta (β_N), which leads to a high power density with reasonable physics requirements. This is because high κ facilitates high n_GW (because of an increase in I_p), which allows efficient use of the capacity of high β_N. From an engineering aspect, low-A may ensure ease in designing blanket modules robust to electromagnetic forces acting on disruptions. Thus, a superconducting low-A tokamak reactor such as SlimCS can be a promising DEMO concept with physics and engineering advantages.

The massive injection of impurity gas into a plasma has been proved to reduce forces and localized thermal loads caused by disruptions in tokamaks. This mitigation system is routinely used on ASDEX Upgrade to shut down plasmas with a locked mode. The plasma response to impurity injection and the mechanism of reduction of the mechanical forces is discussed in the paper.

The role of zonal flows in the formation of transport barrier in helical plasmas is analysed using the transport code. A set of one-dimensional transport equations is analysed, including the effect of zonal flows. The turbulent transport coefficient is shown to be suppressed when the plasma state takes the strong positive radial electric field. This bifurcation of the turbulent transport is newly caused by the change of the damping rate of zonal flows. It is theoretically demonstrated that the damping rate of zonal flows plays an important role in global confinement in toroidal plasmas.

Beyond a certain heating power, measured and predicted distributions of neutral beam injection (NBI) driven currents deviate from each other even in the absence of MHD instabilities. The most reasonable explanation is a redistribution of fast NBI ions on a time scale smaller than the current redistribution time. The hypothesis of a redistribution of fast ions by background turbulence is discussed. Direct numerical simulation of fast test particles in a given field of electrostatic turbulence indicates that for reasonable parameters fast and thermal particle diffusion can indeed be similar. High quality plasma edge density profiles on ASDEX Upgrade and the recent extension of the reflectometry system allow for a direct comparison of observed TAE eigenfunctions with theoretical ones as obtained with the linear, gyrokinetic, global stability code LIGKA. These comparisons support the hypothesis of TAE-frequency crossing the continuum at the plasma edge in ASDEX Upgrade H-mode discharges. A new fast ion loss detector with 1 MHz time resolution allows frequency and phase resolved correlation between the observed losses and low frequency magnetic perturbations such as TAE modes and rotating magnetic islands. Whereas losses caused by TAE modes are known to be due to resonances in velocity space, by modelling the particle drift orbits we were able to explain losses caused by magnetic islands as due to island formation and stochasticity in the drift orbits.

Zonal flow behaviour and its effect on turbulent transport in tokamak plasmas with reversed magnetic shear are investigated by global fluid simulations of electrostatic ion temperature gradient driven turbulence. It is found that for high q (safety factor) turbulent transport is high even at a minimum-q region because oscillatory zonal flows called geodesic acoustic modes (GAMs) are dominant. The turbulent transport is reduced in the neighbourhood of a minimum-q surface when q is low enough to damp the GAMs. The difference in zonal flow behaviour causes a difference in turbulent transport and may trigger the formation of ion internal transport barriers in reversed shear tokamaks.

Density fluctuation measurements covering a wavenumber range relevant to ITG, TEM and ETG type instabilities (k_⊥ ρ_i ≈ 0.2→ 10) are compared during Ohmic and neutral beam injection (NBI). These short time duration NBI perturbations modified the background temperature as well as the density fluctuation behaviour but had little effect on the background density. It is found that both the measured and calculated responses to these perturbations varied with wavenumber supporting the need for broad wavenumber comparisons. The experimentally measured fluctuation changes and calculated growth rates (from the linear gyrokinetic code GKS) show qualitative similarities to each other at low and intermediate k while the comparison at high k is inconclusive.

The requirements of the DIII-D physics program have led to the development of many operational control results with direct relevance to ITER. These include new algorithms for robust and sustained stabilization of neoclassical tearing modes with electron cyclotron current drive, model-based controllers for stabilization of the resistive wall mode in the presence of ELMs, coupled linear–nonlinear algorithms to provide good dynamic axisymmetric control while avoiding coil current limits, and adaptation of the DIII-D plasma control system (PCS) to operate next-generation superconducting tokamaks. Development of integrated plasma control (IPC), a systematic approach to model-based design and controller verification, has enabled successful experimental application of high reliability control algorithms requiring a minimum of machine operations time for testing and tuning. The DIII-D PCS hardware and software and its versions adapted for other devices can be connected to IPC simulations to confirm control function prior to experimental use. This capability has been important in control system implementation for tokamaks under construction and is expected to be critical for ITER.

Intense electron cyclotron resonance heating (ECRH) and electron cyclotron current drive (ECCD) are employed on the Tokamak à Configuration Variable (TCV) both in second- and third-harmonic X-mode (X2 and X3). The plasma behaviour under such conditions is driven largely by the electron dynamics, motivating extensive studies of the heating and relaxation phenomena governing both the thermal and suprathermal electron populations. In particular, the dynamics of suprathermal electrons are intimately tied to the physics of X2 ECCD. ECRH is also a useful tool for manipulating the electron distribution function in both physical and velocity space. Fundamental studies of the energetic electron dynamics have been performed using periodic, low-duty-cycle bursts of ECRH, with negligible average power injection, and with electron cyclotron emission (ECE). The characteristic times of the dynamical evolution are clearly revealed. Suprathermal electrons have also been shown to affect the absorption of X3 radiation. Thermal electrons play a crucial role in high density plasmas where indirect ion heating can be achieved through ion–electron collisions. In recent experiments ≈ 1.35 MW of vertically launched X3 ECRH was coupled to a diverted ELMy H-mode plasma. In cases where ≥ 1.1 MW of ECRH power was coupled, the discharge was able to transition into a quasi-stationary ELM-free H-mode regime. These H-modes operated at β_N ≈ 2, and had high energy confinement, H_{IPB98(y, 2)} up to ≈ 1.6. Despite being purely electron heated and having no net particle source these discharges maintained a density peaking factor (n_e,o/⟨n_e⟩ ≈ 1.6). They also exhibited spontaneous toroidal momentum production in the co-current direction. The momentum production is due to a transport process as there is no external momentum input. This process supports little or no radial gradient of the toroidal velocity.

Understanding the physics of the edge pedestal and edge localized modes (ELMs) is of great importance for ITER and the optimization of the tokamak concept. The peeling–ballooning model has quantitatively explained many observations, including ELM onset and pedestal constraints, in the standard H-mode regime. The ELITE code has been developed to efficiently evaluate peeling–ballooning stability for comparison with observation and predictions for future devices. We briefly review recent progress in the peeling–ballooning model, including experimental validation of ELM onset and pedestal height predictions, and nonlinear 3D simulations of ELM dynamics, which together lead to an emerging understanding of the physics of the onset and dynamics of ELMs in the standard intermediate to high collisionality regime. We also discuss new studies of the apparent power dependence of the pedestal, and studies of the impact of sheared toroidal flow. Recently, highly promising low collisionality regimes without ELMs have been discovered, including the quiescent H-mode (QH) and resonant magnetic perturbation (RMP) regimes. We present recent observations from the DIII-D tokamak of the density, shape and rotation dependence of QH discharges, and studies of the peeling–ballooning stability in this regime. We propose a model of the QH-mode in which the observed edge harmonic oscillation (EHO) is a saturated kink/peeling mode which is destabilized by current and rotation, and drives significant transport, allowing a near steady-state edge plasma. The model quantitatively predicts the observed density dependence and qualitatively predicts observed mode structure, rotation dependence and outer gap dependence. Low density RMP discharges are found to operate in a similar regime, but with the EHO replaced by an applied magnetic perturbation.

This paper summarizes results of dimensionless identity experiments in JT-60U and JET, aimed at the comparison of the H-mode pedestal and ELM behaviour in the two devices. Given their similar size, dimensionless matched plasmas are also similar in their dimensional parameters (in particular, the plasma minor radius a is the same in JET and JT-60U). Power and density scans were carried out at two values of I_p, providing a q scan (q₉₅ = 3.1 and 5.1) with fixed (and matched) toroidal field. Contrary to initial expectations, a dimensionless match between the two devices was quite difficult to achieve. In general, p_ped in JT-60U is lower than in JET and, at low q, the pedestal pressure of JT-60U with a Type I ELMy edge is matched in JET only in the Type III ELM regime. At q₉₅ = 5.1, a dimensionless match in ρ^*, ν^* and β_p,ped is obtained with Type I ELMs, but only with low power JET H-modes. These results motivated a closer investigation of experimental conditions in the two devices, to identify possible 'hidden' physics that prevents obtaining a good match of pedestal values over a large range of plasmas parameters. Ripple-induced ion losses of the medium bore plasma used in JT-60U for the similarity experiments are identified as the main difference with JET. The magnitude of the JT-60U ripple losses is sufficient to induce counter-toroidal rotation in co-injected plasma. The influence of ripple losses was demonstrated at q₉₅ = 5.1: reducing ripple losses by ≈2 (from 4.3 to 1.9 MW) by replacing positive with negative neutral beam injection at approximately constant P_in resulted in an increased p_ped in JT-60U, providing a good match to full power JET H-modes. At the same time, the counter-toroidal rotation decreased. Physics mechanisms relating ripple losses to pedestal performance are not yet identified, and the possible role of velocity shear in the pedestal stability, as well as the possible influence of ripple on thermal ion transport are briefly discussed. Toroidal rotation of the ITER reference inductive Q = 10 H-mode is predicted to be rather low, of the order of ∼1/10 of the frequency of typical JET H-modes. Nonetheless, fast ion ripple losses in that scenario are also predicted to be negligible (≲1%), and therefore plasma toroidal rotation slow-down or ripple-induced counter-rotation should not affect pedestal parameters and stability in ITER. Finally, the possible effect of ripple on thermal transport may deserve more attention in future experiments and modelling, since the ripple magnitude of ITER is intermediate between that of JET and JT-60U.

Recent experiments in ASDEX Upgrade (AUG) have experienced surprisingly high He plasma impurity concentrations. Such high He concentrations have not been observed with C walls, they were only observed since the increase in the W first-wall coverage of ASDEX Upgrade to 85%. The high He plasma concentration appears to be linked to the fraction of W surfaces open to plasma contact that are not covered by boronization layers and to the number of He glow discharges performed for wall conditioning prior to normal plasma operation. This points to the different retention and release properties of W and C for He. To elucidate these differences dedicated laboratory experiments have been performed. To study the retention, different types of W and C targets, including those used in ASDEX Upgrade, were implanted with ³He at 200 and 600 eV and the amount of retained He was determined through thermal effusion spectroscopy (TES) and ion beam analysis methods. These experiments showed that W can retain up to 10 times more He than C depending on the energy of the implanted ³He. The TES measurement showed that both W and C start to out gas He at ≈400 K which is not reached at the first-wall in AUG. Therefore the differences in the release of He from W and C surfaces due to particle bombardment was investigated. ³He implanted W and C substrates were exposed to ⁴He or H₂ plasmas and the loss of ³He was measured. For an H₂ plasma and 100 eV bias, three times more He is released from W than from C. For ⁴He and 100 eV bias between 2 and 10 times more ³He was released from W than from C depending on the type of W and C. The release rate of He from W and C through particle induced release was found to be equal. Therefore the stronger release of He from W can be explained by the higher retention of He in W.

A clean magnetic boundary, i.e. a vanishing radial magnetic field, is important for reversed field pinch operation. In order to achieve this target a large modification of the RFX device (a = 0.45 m, R = 2 m) into RFX-mod has been realized and 192 individually controlled saddle coils which provide a flexible control of magnetic boundary have been installed. Recent experiments have shown that this change has produced a significant improvement in plasma magnetohydrodynamic (MHD) behaviour.

This paper discusses 3D MHD numerical simulations aimed at interpreting these results. In particular it is shown that these simulations can predict the measured level of magnetic fluctuations in different operational scenarios. They can also predict the dominat modes and show in general an agreement with experiments. A specific comparison for a resistive wall mode branch is also done, showing differences between simulations and experiments which may indicate the presence of field errors.

Ferritic steel tiles (FSTs) have been installed to improve the energetic ion confinement by reducing a toroidal magnetic field ripple. Aiming at cost-effective installation, orbit-following calculations of energetic ions were carried out for a design of the installation of ferritic steel on the JT-60U by using the fully three dimensional magnetic field orbit-following Monte-Carlo (F3D OFMC) code, which had been developed for ferritic insert experiments on the JFT-2M and can treat the complex magnetic field structure produced by ferritic inserts. The installed FSTs add a non-linear magnetic field on magnetic sensors for plasma control and an equilibrium calculation. The code for real-time control has been modified to take into account the magnetic field by the FSTs. The plasma operation was successfully resumed after usual conditioning processes and real-time plasma control was successfully carried out. The heat load measurement indicates the improved confinement of energetic ions. These results are important for practical application of the ferritic steel, which is a leading candidate of a structural material on a DEMO reactor.

The National Spherical Torus Experiment (NSTX) offers an operational space characterized by high-beta (β_t = 39%, β_N > 7, ) and low aspect ratio (A > 1.27) to leverage the plasma parameter dependences of RWM stabilization and plasma rotation damping physics giving greater confidence for extrapolation to ITER. Significant new capability for RWM research has been added to the device with the commissioning of a set of six non-axisymmetric magnetic field coils, allowing generation of fields with dominant toroidal mode number, n, of 1–3. These coils have been used to study the dependence of resonant field amplification on applied field frequency and RWM stabilization physics by reducing the toroidal rotation profile below its steady-state value through non-resonant magnetic braking. Modification of plasma rotation profiles shows that rotation outside q = 2.5 is not required for passive RWM stability and there is large variation in the RWM critical rotation at the q = 2 surface, both of which are consistent with distributed dissipation models.

This paper reports on the recent studies of toroidal and poloidal momentum transport in JET. The ratio of the global energy confinement time to the momentum confinement is found to be close to τ_E/τ_ϕ = 1 except for the low density or low collisionality discharges where the ratio is τ_E/τ_ϕ = 2–3. On the other hand, local transport analysis of around 40 discharges shows that the ratio of the local effective momentum diffusivity to the ion heat diffusivity is χ_ϕ/χ_i ≈ 0.1–0.4 (averaged over the radial region r/a = 0.4–0.7) rather than unity, as expected from the global confinement times and used often in ITER predictions. The apparent discrepancy in the global and local momentum versus ion heat transport can be at least partly explained by the fact that momentum confinement within edge pedestal is worse than that of the ion heat and thus, momentum pedestal is weaker than that of ion temperature. In addition, while the ion temperature profile shows clearly strong profile stiffness, the toroidal velocity profile does not exhibit stiffness, as exemplified here during a giant ELM crash. Predictive transport simulations with the self-consistent modelling of toroidal velocity using the Weiland model and GLF23 also confirm that the ratio χ_ϕ/χ_i ≈ 0.4 reproduces the core toroidal velocity profiles well and similar accuracy with the ion temperature profiles. Concerning poloidal velocities on JET, the experimental measurements show that the carbon poloidal velocity can be an order of magnitude above the neo-classical estimate within the ITB. This significantly affects the calculated radial electric field and therefore, the E × B flow shear used for example in transport simulations. Both the Weiland model and GLF23 reproduce the onset, location and strength of the ITB well when the experimental poloidal velocity is used while they do not predict the formation of the ITB using the neo-classical poloidal velocity in time-dependent transport simulation. The most plausible explanation for the generation of the anomalous poloidal velocity is the turbulence driven flow through the Reynolds stress. Both CUTIE and TRB turbulence codes show the existence of an anomalous poloidal velocity, being significantly larger than the neo-classical values. And similarly to experiments, the poloidal velocity profiles peak in the vicinity of the ITB and seem to be dominantly caused by flow due to the Reynolds stress. However, it is important to note that both the codes treat the equilibrium in a simplified way and this affects the geodesic curvature effects and geodesic acoustic modes (GAMs). Therefore, the results should be considered as indicative, and most probably provide an upper bound of the mean poloidal velocity as results from other codes including GAM dynamics show that they often serve as a damping mechanism to flows.

Novel features of the rapid growth and the nonlinear dynamics of driven magnetic islands in rotating plasmas are studied by nonlinear magnetohydrodynamics simulations. The transition phase of the magnetic island evolution is found between the rapid growth phase and the Rutherford-like phase in the low resistivity regime. It is shown that the rapid growth of the magnetic island is associated with its deformation that leads to the secondary magnetic island formation around the original X-points. The suppressed state is characterized by the asymmetric deformation of the magnetic island caused by the non-monotonic torque profile within the island separatrix. In the regime of the high magnetic flux input rate, the critical magnetic island width shows weak dependence on viscosity while it noticeably depends on resistivity. It is conjectured that this is a result of the effect of the Alfven type interaction between the rotating plasma and the fast growing external magnetic perturbation.

In H₂ discharges of large helical device (LHD), H-mode-like transition has been obtained with edge localized mode (ELM)-like bursts in an outwardly shifted configuration of R_ax = 4.00 m with an extremely thick ergodic layer, where the ι/2π = 1 position is located in the middle of the ergodic layer. This H-mode-like transition can be triggered by changing P_NBI (<12 MW) from 3 beams to 2 beams in a density range (4–8) × 10¹³ cm⁻³. The spontaneous H-mode-like transition is also observed during a stable discharge with constant heating power and density, where a steep edge pressure gradient is formed at the ι/2π = 1 position. The ELM-like bursts occurred in the ergodic layer accompanied with density bursts. It is found that the density bursts triggered at the ι/2π = 1 position in the ergodic layer instantaneously propagate outward beyond the ergodic layer with positive density pulse, and a few ms later also begin to propagate inward from the ergodic layer with negative density pulse to recover the previous edge density profile. In the case of He discharges no H-mode-like transition and ELM-like bursts are observed. The frequency of the ELM-like bursts increases with heating power, e.g. 50 Hz (3 MW)–100 Hz (5 MW) at n_e = 4 × 10¹³ cm⁻³. The external resonant magnetic field to generate a static m/n = 1/1 island inside the ergodic layer is supplied to the ELM-like bursts. It is also found that the externally supplied magnetic field makes a clear response even in the ergodic layer. As a result, the frequency of the ELM-like bursts decreases, e.g. from 60 to 45 Hz. The present study strongly indicates that those phenomena observed in LHD are quite different from the ones which have been observed in tokamaks until now.

Non-inductive currents of electron cyclotron heated plasmas have been examined in the helical-axis heliotron device, Heliotron J. The bootstrap and EC currents were separated by comparing experiments with positive and negative magnetic field. The estimated bootstrap current was found to be affected by the magnetic field configuration. It increases with an increase in the bumpy component of the magnetic field spectrum, which agrees well with a neoclassical prediction calculated using the SPBSC code. The EC current driven by oblique launch with respect to the magnetic field strongly depends on the field configuration and the location of the EC power deposition. The EC current is enhanced when the EC power is deposited on the magnetic axis. The maximum EC current is I_EC = −4.6 kA and the current drive efficiency is η = n_eRI_p/P_EC = 8.4 × 10¹⁶ A W⁻¹ m⁻². The flow direction of the EC current depends on the magnetic field ripple structure where the EC power is deposited.

The dual-functional lithium lead-test blanket module (DFLL-TBM) system, which is designated to demonstrate the integrated technologies of both the He single coolant (SLL) blanket and the He-LiPb dual coolant (DLL) blanket, is proposed for testing in ITER to check and validate the feasibility of the Chinese LiPb blankets. Considering the conflict between the limited ITER resources for TBM testing and the requirement of various blanket concepts proposed by the parties, EAST, the superconducting tokamak being operated in China, can serve as a valuable pre-testing platform with some comparable design parameters to ITER for DFLL-TBM. In the contribution, neutronics calculations and activation of the TBMs in EAST are carried out and compared with those in the D–D and D–T phase of ITER. The results from EAST can make an analogy to ITER (to some extent).

Mechanisms determining the structure of edge temperature and density pedestals, which are associated with edge transport barrier (ETB) formation in tokamaks, are investigated on Alcator C-Mod. Experiments suggest a strong role for critical gradient behaviour in setting profile characteristics of edge plasma. The maximum pressure gradient scales as the square of plasma current in both H-modes without edge-localized modes, and in the near scrape-off layer in Ohmic discharges. In either case, the pressure gradient obtained, normalized to the square of plasma current, is a function of local collisionality, hinting that common physics may contribute to setting profile gradients in both confinement regimes. Varying the neutral fuelling source has little effect on density gradient scale lengths in the ETB and a relatively weak impact on the height of the density pedestal, even during aggressive deuterium puffing. Strong screening of neutrals in the ETB are observed, creating a challenge for fuelling H-modes in C-Mod beyond their 'natural' density. A simple pedestal fuelling model does not reproduce the typically clamped density gradients seen in experiment during H-mode puffing. These results suggest that a simple diffusive model for plasma transport is deficient, and that a critical gradient assumption for transport may be essential for pedestal modelling.

A new means of operating flux-core spheromaks with possibly increased stability, confinement and pulse length is analysed by a resistive magnetohydrodynamic (MHD) model. High amplification of the bias poloidal flux, required to minimize ohmic losses, is achieved by reducing the bias rapidly in a plasma formed at a lower amplification. The plasma separatrix is predicted to expand and incorporate the removed bias flux maintaining the total poloidal flux within the spheromak's flux-conserving wall. MHD energy on open magnetic field lines is reduced, reducing magnetic fluctuation levels. A means of experimental verification is suggested that may point the way to fusion-relevant spheromaks.