Sustained edge localized mode (ELM) mitigation has been achieved on MAST and AUG using resonant magnetic perturbations (RMPs) with various toroidal mode numbers over a wide range of low to medium collisionality discharges. The ELM energy loss and peak heat loads at the divertor targets have been reduced. The ELM mitigation phase is typically associated with a drop in plasma density and overall stored energy. In one particular scenario on MAST, by carefully adjusting the fuelling it has been possible to counteract the drop in density and to produce plasmas with mitigated ELMs, reduced peak divertor heat flux and with minimal degradation in pedestal height and confined energy. While the applied resonant magnetic perturbation field can be a good indicator for the onset of ELM mitigation on MAST and AUG there are some cases where this is not the case and which clearly emphasize the need to take into account the plasma response to the applied perturbations. The plasma response calculations show that the increase in ELM frequency is correlated with the size of the edge peeling-tearing like response of the plasma and the distortions of the plasma boundary in the X-point region. In many cases the RMPs act to increase the frequency of type I ELMs, however, there are examples where the type I ELMs are suppressed and there is a transition to a small or type IV ELM-ing regime.

Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor.

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak e Xperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility ( P/ S ∼ 1.5 MW m ⁻²) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma–material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv:1409.3540)) that makes use of high-temperature superconductor technology—a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

Periods of edge localized mode (ELM)-free H-mode with increased pedestal pressure and width were observed in the DIII-D tokamak when density fluctuations localized to the region near the separatrix were present. Injection of a powder of 45  µm diameter lithium particles increased the duration of the enhanced pedestal phases to up to 350 ms, and also increased the likelihood of a transition to the enhanced phase. Lithium injection at a level sufficient for triggering the extended enhanced phases resulted in significant lithium in the plasma core, but carbon and other higher Z impurities as well as radiated power levels were reduced. Recycling of the working deuterium gas appeared unaffected by this level of lithium injection. The ion scale, k _θ ρ _s ∼ 0.1–0.2, density fluctuations propagated in the electron drift direction with f ∼ 80 kHz and occurred in bursts every ∼1 ms. The fluctuation bursts correlated with plasma loss resulting in a flattening of the pressure profile in a region near the separatrix. This localized flattening allowed higher overall pedestal pressure at the peeling–ballooning stability limit and higher pressure than expected under the EPED model due to reduction of the pressure gradient below the ‘ballooning critical profile’. Reduction of the ion pressure by lithium dilution may contribute to the long ELM-free periods.

The ITER project requires additional heating by two neutral beam injectors, each accelerating to 1 MV a 40 A beam of negative deuterium ions, to deliver to the plasma a power of about 17 MW for one hour. As these requirements have never been experimentally met, it was recognized as necessary to setup a test facility, PRIMA (Padova Research on ITER Megavolt Accelerator), in Italy, including a full-size negative ion source, SPIDER, and a prototype of the whole ITER injector, MITICA, aiming to develop the heating injectors to be installed in ITER. This realization is made with the main contribution of the European Union, through the Joint Undertaking for ITER (F4E), the ITER Organization and Consorzio RFX which hosts the Test Facility. The Japanese and the Indian ITER Domestic Agencies (JADA and INDA) participate in the PRIMA enterprise; European laboratories, such as IPP-Garching, KIT-Karlsruhe, CCFE-Culham, CEA-Cadarache and others are also cooperating. Presently, the assembly of SPIDER is on-going and the MITICA design is being completed. The paper gives a general overview of the test facility and of the status of development of the MITICA and SPIDER main components at this important stage of the overall development; then it focuses on the latest and most critical issues, regarding both physics and technology, describing the identified solutions.

The heating of tungsten monoblocks at the ITER divertor vertical targets is calculated using the heat flux predicted by three-dimensional ion orbit modelling. The monoblocks are beveled to a depth of 0.5 mm in the toroidal direction to provide magnetic shadowing of the poloidal leading edges within the range of specified assembly tolerances, but this increases the magnetic field incidence angle resulting in a reduction of toroidal wetted fraction and concentration of the local heat flux to the unshadowed surfaces. This shaping solution successfully protects the leading edges from inter-ELM heat loads, but at the expense of (1) temperatures on the main loaded surface that could exceed the tungsten recrystallization temperature in the nominal partially detached regime, and (2) melting and loss of margin against critical heat flux during transient loss of detachment control. During ELMs, the risk of monoblock edge melting is found to be greater than the risk of full surface melting on the plasma-wetted zone. Full surface and edge melting will be triggered by uncontrolled ELMs in the burning plasma phase of ITER operation if current models of the likely ELM ion impact energies at the divertor targets are correct. During uncontrolled ELMs in pre-nuclear deuterium or helium plasmas at half the nominal plasma current and magnetic field, full surface melting should be avoided, but edge melting is predicted.

Nonlinear dynamics of single toroidal number Alfvén eigenmodes destabilised by the the resonant interaction with fast ions is investigated, in tokamak equilibria, by means of Hamiltonian mapping techniques. The results obtained by two different simulation codes, XHMGC and HAGIS, are presented for n  =  2 Beta induced Alfvén eigenmodes and, respectively n  =  6 toroidal Alfvén eigenmodes. Simulations of the bump-on-tail instability performed by a 1-dimensional code, PIC1DP, are also analysed for comparison. As a general feature, modes saturate as the resonant-particle distribution function is flattened over the whole region where mode-particle power transfer can take place in the linear phase. Such region is limited by the narrowest of resonance width and mode width. In the former case, mode amplitude at saturation exhibits a quadratic scaling with the linear growth rate; in the latter case, the scaling is linear. These results are explained in terms of the approximate analytic solution of a nonlinear pendulum model. They are also used to prove that the radial width of the single poloidal harmonic sets an upper limit to the radial displacement of circulating fast ions produced by a single-toroidal-number gap mode in the large n limit, irrespectively of the possible existence of a large global mode structure formed by many harmonics.

Although the ultimate goal of most current fusion research is to build an economically attractive power plant, the present status of physics and technology does not provide the performance necessary to achieve this goal. Therefore, in order to model how such plants may operate and what their output might be, extrapolations must be made from existing experimental data and technology. However, the expected performance of a plant built to the operating point specifications can only ever be a ‘best guess’. Extrapolations far beyond the current operating regimes are necessarily uncertain, and some important interactions, for example the coupling of conducted power from the scape-off layer to the divertor surface, lack reliable predictive models. This means both that the demands on plant systems at the target operating point can vary significantly from the nominal value, and that the overall plant performance may potentially fall short of design targets.

In this contribution we discuss tools and techniques that have been developed to assess the robustness of the operating points for the EU-DEMO tokamak-based demonstration power plant, and the consequences for its design. The aim is to make explicit the design choices and areas where improved modelling and DEMO-relevant experiments will have the greatest impact on confidence in a successful DEMO design.

The helium cooled lithium lead (HCLL) blanket concept was originally developed in CEA at the beginning of 2000: it is one of the two European blanket concepts to be tested in ITER in the form of a test blanket module (TBM) and one of the four blanket concepts currently being considered for the DEMOnstration reactor that will follow ITER. The TBM is a highly optimized component for the ITER environment that will provide crucial information for the development of the DEMO blanket, but its design needs to be adapted to the DEMO reactor. With respect to the TBM design, reduction of the steel content in the breeding zone (BZ) is sought in order to maximize tritium breeding reactions. Different options are being studied, with the potential of reaching tritium breeding ratio (TBR) values up to 1.21. At the same time, the design of the back supporting structure (BSS), which is a DEMO specific component that has to support the blanket modules inside the vacuum vessel (VV), is ongoing with the aim of maximizing the shielding power and minimizing pumping power. This implies a re-engineering of the modules’ attachment system. Design changes however, will have an impact on the manufacturing and assembly sequences that are being developed for the HCLL-TBM. Due to the differences in joint configurations, thicknesses to be welded, heat dissipation and the various technical constraints related to the accessibility of the welding tools and implementation of non-destructive examination (NDE), the manufacturing procedure should be adapted and optimized for DEMO design. Laser welding instead of TIG could be an option to reduce distortions. The time-of-flight diffraction (TOFD) technique is being investigated for NDE. Finally, essential information expected from the HCLL-TBM program that will be needed to finalize the DEMO design is discussed.

Fast ion prompt loss induced by resonant magnetic perturbations (RMPs) is simulated by solving Hamiltonian equations strictly in the guiding center coordinate system. Full orbit simulations show that the prompt loss rate can increase significantly in resonant regions when RMPs are added. Furthermore, the prompt loss rate is larger in the low-field side than in the high-field side in tokamak plasmas. Detailed analyses show that a number of trapped ions which lie near the center of the trapped region can be lost, because of the enhancement of radial orbit drifts induced by the resonance between RMPs and the unperturbed orbit. Meanwhile, orbit conversion from counter-passing orbit to trapped orbit occurs near the trapped-passing boundary in the low-field side, while it occurs near the co-counter boundary in the high-field side, both of which play an important role in prompt loss. Simulations also demonstrate a periodicity for orbit drifts, and the mechanism of drift periodicity results from the resonance between RMP and the equilibrium magnetic field.

Neoclassical transport processes are important to the understanding of plasma confinement physics in doubly periodic magnetized toroidal plasmas, especially, after the impact of the momentum confinement on the particle and energy confinement is recognized. Real doubly periodic tori in general are non-axisymmetric, with symmetric tori as a special case. An eight-moment approach to transport theory with plasma density N, plasma pressure p, mass flow velocity V and heat flow q as independent variables is adopted. Transport processes are dictated by the solutions of the momentum and heat flux balance equations. For toroidal plasma confinement devices, the first order (in the gyro-radius ordering) plasma flows are on the magnetic surface to guarantee good plasma confinement and are thus two-dimensional. Two linearly independent components of the momentum equation are required to determine the flows completely. Once this two-dimensional flow is relaxed, i.e. the momentum equation reaches a steady state, plasmas become ambipolar, and all the transport fluxes are determined through the flux–force relation. The flux–force relation is derived both from the kinetic definitions for the transport fluxes and from the manipulation of the momentum and heat flux balance equations to illustrate the nature of the transport fluxes by examining their corresponding driven forces and their roles in the momentum and heat flux balance equations. Steady-state plasma flows are determined by the components of the stress and heat stress tensors in the momentum and heat flux balance equations. This approach emphasizes the pivotal role of the momentum equation in the transport processes and is particularly useful in modelling plasma flows in experiments. The methodology for neoclassical transport theory is applied to fluctuation-driven transport fluxes in the quasilinear theory to unify these two theories. Experimental observations in tokamaks and stellarators for the physics discussed are presented.

The physics design of the accelerator for the heating neutral beamline on ITER is now finished and this paper describes the considerations and choices which constitute the basis of this design. Equal acceleration gaps of 88 mm have been chosen to improve the voltage holding capability while keeping the beam divergence low. Kerbs (metallic plates around groups of apertures, attached to the downstream surface of the grids) are used to compensate for the beamlet–beamlet interaction and to point the beamlets in the right direction. A novel magnetic configuration is employed to compensate for the beamlet deflection caused by the electron suppression magnets in the extraction grid. A combination of long-range and short-range magnetic fields is used to reduce electron leakage between the grids and limit the transmitted electron power to below 800 kW.

The physics of non-axisymmetry is a far more important topic in the theory of toroidal fusion plasmas than might be expected. (1) Even a small toroidal asymmetry in the magnetic field strength, δ ≡ ∂ln  B/∂ φ ∼ 10 ⁻⁴, can cause an unacceptable degradation in performance. (2) Nevertheless, asymmetries—even large asymmetries δ ∼ 1—can give beneficial plasma control and circumvent issues, such as magnetic-configuration maintenance and plasma disruptions, that make axisymmetric fusion devices problematic. Viewed from prospectives that are adequate for designing and studying axisymmetric plasmas, the physics of non-axisymmetric plasmas appears dauntingly difficult. Remarkably, Maxwell's equations provide such strong constraints on the physics of toroidal fusion plasmas that even a black-box model of a plasma answers many important questions. Kinetic theory and non-equilibrium thermodynamics provide further, but more nuanced, constraints. This paper is organized so these constraints can be used as a basis for the innovations and for the extrapolations that are required to go from existing experiments to fusion systems. Outlines are given of a number of calculations that would be of great importance to ITER and to the overall fusion program and that could be carried out now with limited resources.

The area of energetic particle (EP) physics in fusion research has been actively and extensively researched in recent decades. The progress achieved in advancing and understanding EP physics has been substantial since the last comprehensive review on this topic by Heidbrink and Sadler (1994 Nucl. Fusion 34 535). That review coincided with the start of deuterium–tritium (DT) experiments on the Tokamak Fusion Test Reactor (TFTR) and full scale fusion alphas physics studies.

Fusion research in recent years has been influenced by EP physics in many ways including the limitations imposed by the ‘sea’ of Alfvén eigenmodes (AEs), in particular by the toroidicity-induced AE (TAE) modes and reversed shear AEs (RSAEs). In the present paper we attempt a broad review of the progress that has been made in EP physics in tokamaks and spherical tori since the first DT experiments on TFTR and JET (Joint European Torus), including stellarator/helical devices. Introductory discussions on the basic ingredients of EP physics, i.e., particle orbits in STs, fundamental diagnostic techniques of EPs and instabilities, wave particle resonances and others, are given to help understanding of the advanced topics of EP physics. At the end we cover important and interesting physics issues related to the burning plasma experiments such as ITER (International Thermonuclear Experimental Reactor).

Poloidal and toroidal rotation has been recognized to play an important role in heat transport and magnetohydrodynamic (MHD) stability in tokamaks and helical systems. It is well known that the E ×  B shear due to poloidal and toroidal flow suppresses turbulence in the plasma and contributes to the improvement of heat and particle transport, while toroidal rotation helps one to stabilize MHD instabilities such as resistive wall modes and neoclassical tearing mode. Therefore, understanding the role of momentum transport in determining plasma rotation is crucial in toroidal discharges, both in tokamaks and helical systems. In this review paper, the driving and damping mechanisms of poloidal and toroidal rotation are outlined. Driving torque due to neutral beam injection and radio-frequency waves, and damping due to parallel viscosity and neoclassical toroidal viscosity (NTV) are described. Regarding momentum transport, the radial flux of momentum has diffusive and non-diffusive (ND) terms, and experimental investigations of these are discussed. The magnitude of the diffusive term of momentum transport is expressed as a coefficient of viscous diffusivity. The ratio of the viscous diffusivity to the thermal diffusivity (Prandtl number) is one of the interesting parameters in plasma physics. It is typically close to unity, but sometimes can deviate significantly depending on the turbulent state. The ND terms have two categories: one is the so-called momentum pinch, whose magnitude is proportional to (or at least depends on) the velocity itself, and the other is an off-diagonal term in which the magnitude is proportional to (or at least depends on) the temperature or/and pressure gradient, independent of the velocity or its gradient. The former has no sign dependence; rotation due to the momentum pinch does not depend on the sign of the rotation itself, whether it is parallel to the plasma current (co-direction) or anti-parallel to the plasma current (counter-direction). In contrast, the latter has a sign dependence; the rotation due to the off-diagonal residual term is either in the co- or counter-direction depending on the turbulence state, but not on the sign of the rotation itself. This residual term can also act as a momentum source for intrinsic rotation. The experimental results of investigations of these ND terms are described. Finally the current understanding of the mechanisms behind the ND terms in momentum transport, and predictions of intrinsic rotation driven by these terms are reviewed.

The effects of electrode biasing (EB) on the m/ n  =  2/1 tearing mode have been experimentally studied in J-TEXT tokamak discharges, where m and n are the poloidal and toroidal mode numbers. It is found that for a negative bias voltage, the mode amplitude is reduced, and the mode frequency is increased accompanied by the increased toroidal plasma rotation speed in the counter- I _p direction. For a positive bias voltage, the mode frequency is decreased together with the change of the rotation velocity towards the co- I _p direction, and the mode amplitude is increased. Statistic results show that the variations in the toroidal rotation speed, the 2/1 mode frequency and its amplitude linearly depend on the bias voltage. The threshold voltages for complete suppression and locking of the mode are found. The experimental results suggest that applied electrode biasing is a possible method for the avoidance of mode locking and disruption.

A large scale program to develop a conceptual design for a demonstration fusion power plant (DEMO) has been initiated in Europe. Central elements are the baseline design points, which are developed by system codes. The assessment of the credibility of these design points is often hampered by missing information. The main physics and technology content of the central European system codes have been published (Kovari et al 2014 Fusion Eng. Des. 89 3054–69, 2016 Fusion Eng. Des. 104 9–20, Reux et al 2015 Nucl. Fusion 55 073011). In addition, this publication discusses key input parameters for the pulsed and conservative design option and provides justifications for the parameter choices. In this context several DEMO physics gaps are identified, which need to be addressed in the future to reduce the uncertainty in predicting the performance of the device.

Also the sensitivities of net electric power and pulse duration to variations of the input parameters are investigated. The most extreme sensitivity is found for the elongation ( corresponds to ).

The I-mode is a plasma regime obtained when the usual L-H power threshold is high, e.g. with unfavourable ion direction. It is characterised by the development of a temperature pedestal while the density remains roughly as in the L-mode. This leads to a confinement improvement above the L-mode level which can sometimes reach H-mode values. This regime, already obtained in the ASDEX Upgrade tokamak about two decades ago, has been studied again since 2009 taking advantage of the development of new diagnostics and heating possibilities. The I-mode in ASDEX Upgrade has been achieved with different heating methods such as NBI, ECRH and ICRF. The I-mode properties, power threshold, pedestal characteristics and confinement, are independent of the heating method. The power required at the L-I transition exhibits an offset linear density dependence but, in contrast to the L-H threshold, depends weakly on the magnetic field. The L-I transition seems to be mainly determined by the edge pressure gradient and the comparison between ECRH and NBI induced L-I transitions suggests that the ion channel plays a key role. The I-mode often evolves gradually over a few confinement times until the transition to H-mode which offers a very interesting situation to study the transport reduction and its link with the pedestal formation. Exploratory discharges in which n  =  2 magnetic perturbations have been applied indicate that these can lead to an increase of the I-mode power threshold by flattening the edge pressure at fixed heating input power: more heating power is necessary to restore the required edge pressure gradient. Finally, the confinement properties of the I-mode are discussed in detail.

Toroidal computations are performed using the MARS-F code (Liu et al 2000 Phys. Plasmas 7 3681), in order to understand correlations between the plasma response and the observed mitigation of the edge localized modes (ELM) using resonant magnetic perturbation fields in ASDEX Upgrade. In particular, systematic numerical scans of the edge safety factor reveal that the amplitude of the resonant poloidal harmonic of the response radial magnetic field near the plasma edge, as well as the plasma radial displacement near the X-point, can serve as good indicators for predicting the optimal toroidal phasing between the upper and lower rows of coils in ASDEX Upgrade. The optimal coil phasing scales roughly linearly with the edge safety factor , for various choices of the toroidal mode number n  =  1–4 of the coil configuration. The optimal coil phasing is also predicted to vary with the upper triangularity of the plasma shape in ASDEX Upgrade. Furthermore, multiple resonance effects of the plasma response, with continuously varying , are computationally observed and investigated.

Fast ion pressure profiles flattened by multiple Alfvén eigenmodes (AEs) are investigated for various neutral beam deposition powers in a multi-phase simulation, which is a combination of classical simulation and hybrid simulation for energetic particles interacting with a magnetohydrodynamic fluid. Monotonic degradation of fast ion confinement and fast ion profile stiffness is found with increasing beam deposition power. The confinement degradation and profile stiffness are caused by a sudden increase in fast ion transport flux brought about by AEs for fast ion pressure gradients above a critical value. The critical pressure gradient and the corresponding beam deposition power depend on the radial location. The fast ion pressure gradient stays moderately above the critical value, and the profiles of the fast ion pressure and fast ion transport flux spread radially outward from the inner region, where the beam is injected. It is found that the square root of the MHD fluctuation energy is proportional to the beam deposition power. Analysis of the time evolutions of the fast ion energy flux profiles reveals that intermittent avalanches take place with contributions from the multiple eigenmodes. Surface of section plots demonstrate that the resonance overlap of multiple eigenmodes accounts for the sudden increase in fast ion transport with increasing beam power. The critical gradient and critical beam power for the profile stiffness are substantially higher than the marginal stability threshold.