Highlights of 2017

Divertor detachment is explored on the TCV tokamak in alternative magnetic geometries. Starting from typical TCV single-null shapes, the poloidal flux expansion at the outer strikepoint is varied by a factor of 10 to investigate the X-divertor characteristics, and the total flux expansion is varied by 70$\%$ to study the properties of the super-X divertor. The effect of an additional X-point near the target is investigated in X-point target divertors. Detachment of the outer target is studied in these plasmas during Ohmic density ramps and with the ion $\nabla$B drift away from the primary X-point. The detachment threshold, depth of detachment, and the stability of the radiation location are investigated using target measurements from the wall-embedded Langmuir probes and two-dimensional CIII line emissivity profiles across the divertor region, obtained from inverted, toroidally-integrated camera data. It is found that increasing poloidal flux expansion results in a deeper detachment for a given line-averaged density and a reduction in the radiation location sensitivity to core density, while no large effect on the detachment threshold is observed. The total flux expansion, contrary to expectations, does not show a significant influence on any detachment characteristics in these experiments. In X-point target geometries, no evidence is found for a reduced detachment threshold despite a 2–3 fold increase in connection length. A reduced radiation location sensitivity to core plasma density in the vicinity of the target X-point is suggested by the measurements.

Demonstrating the production of net electricity and operating with a closed fuel-cycle remain unarguably the crucial steps towards the exploitation of fusion power. These are the aims of a demonstration fusion reactor (DEMO) proposed to be built after ITER. This paper briefly describes the DEMO design options that are being considered in Europe for the current conceptual design studies as part of the Roadmap to Fusion Electricity Horizon 2020. These are not intended to represent fixed and exclusive design choices but rather 'proxies' of possible plant design options to be used to identify generic design/material issues that need to be resolved in future fusion reactor systems. The materials nuclear design requirements and the effects of radiation damage are briefly analysed with emphasis on a pulsed 'low extrapolation' system, which is being used for the initial design integration studies, based as far as possible on mature technologies and reliable regimes of operation (to be extrapolated from the ITER experience), and on the use of materials suitable for the expected level of neutron fluence. The main technical issues arising from the plasma and nuclear loads and the effects of radiation damage particularly on the structural and heat sink materials of the vessel and in-vessel components are critically discussed. The need to establish realistic target performance and a development schedule for near-term electricity production tends to favour more conservative technology choices. The readiness of the technical (physics and technology) assumptions that are being made is expected to be an important factor for the selection of the technical features of the device.

Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling materials, thereby minimizing the release of tritium under normal operation conditions. Finally, solutions for the unique bonding requirements of dissimilar material used in a fusion reactor are demonstrated by describing the current status and prospects of functionally graded materials.

The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L–H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H  =  1 at β_N ~ 1.8 and n/n_GW ~ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D–T campaign and 14 MeV neutron calibration strategy are reviewed.

The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.

Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day devices. A two prong approach is adopted. On the one hand, scenarios with tolerable transient heat and particle loads, including active edge localised mode (ELM) control are developed. On the other hand, divertor solutions including advanced magnetic configurations are studied. Considerable progress has been made on both approaches, in particular in the fields of: ELM control with resonant magnetic perturbations (RMP), small ELM regimes, detachment onset and control, as well as filamentary scrape-off-layer transport. For example full ELM suppression has now been achieved on AUG at low collisionality with n  =  2 RMP maintaining good confinement ${{H}_{\text{H}\left(98,\text{y}2\right)}}\approx 0.95$. Advances have been made with respect to detachment onset and control. Studies in advanced divertor configurations (Snowflake, Super-X and X-point target divertor) shed new light on SOL physics. Cross field filamentary transport has been characterised in a wide parameter regime on AUG, MAST and TCV progressing the theoretical and experimental understanding crucial for predicting first wall loads in ITER and DEMO. Conditions in the SOL also play a crucial role for ELM stability and access to small ELM regimes.

IFMIF, the International Fusion Materials Irradiation Facility, is presently in its engineering validation and engineering design activities (EVEDA) phase under the Broader Approach Agreement.

The engineering design activity (EDA) phase was successfully accomplished within the allocated time.

The engineering validation activity (EVA) phase has focused on validating the Accelerator Facility (AF), the Target Facility and the Test Facility (TF) by constructing prototypes. The ELTL at JAEA^c, Oarai successfully demonstrated the long-term stability of a Li flow under the IFMIF's nominal operational conditions keeping the specified free-surface fluctuations below  ±1 mm in a continuous manner for 25 d. A full-scale prototype of the high flux test module (HFTM) was successfully tested in the HELOKA loop (KIT, Karlsruhe), where it was demonstrated that the irradiation temperature can be set individually and kept uniform. LIPAc, designed and constructed in European labs under the coordination of F4E, presently under installation and commissioning in the Rokkasho Fusion Institute, aims at validating the concept of IFMIF accelerators with a D⁺ beam of 125 mA continuous wave (CW) and 9 MeV. The commissioning phases of the H⁺/D⁺ beams at 100 keV are progressing and should be concluded in 2017; in turn, the commissioning of the 5 MeV beam is due to start during 2017. The D⁺ beam through the superconducting cavities is expected to be achieved within the Broader Approach Agreement time frame with the superconducting cryomodule being assembled in Rokkasho.

The realisation of a fusion-relevant neutron source is a necessary step for the successful development of fusion. The ongoing success of the IFMIF/EVEDA involves ruling out concerns about potential technical showstoppers which were raised in the past. Thus, a situation has emerged where soon steps towards constructing a Li(d,xn) fusion-relevant neutron source could be taken, which is also justified in the light of costs which are marginal to those of a fusion plant.

After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3  ×  10¹⁹ m⁻³, central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre.

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 $\tt{EU\ DEMO1\ 2015}$ 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 ($\Delta {{\kappa}_{95}}=10 \%$ corresponds to $\Delta {{P}_{\text{el},\text{net}}}=125 \%$).

In confined plasma magnetic fusion devices significant amounts of the hydrogen isotopes used for the fusion reaction can be stored in the plasma-facing materials by implantation. The desorption of this retained hydrogen was seen to follow a t^α law with α  ≈  −0.7 in tokamaks. For a pulsed fusion reactor this outgassing can define the inter-pulse waiting time. This work presents new experimental data on the dynamic outgassing in ITER grade tungsten exposed under the well-defined conditions of PSI-2 to pure and mixed D₂ plasmas.

A peak ion flux of 10²² D⁺ m⁻² s is applied for up to 6 h at sample temperatures of up to 900 K. Pure D₂ and mixed D₂  +  He, D₂  +  N₂ and D₂  +  He  +  N₂ plasmas are applied to the sample at 68 V bias. The D₂, He, N outgassing at 293 K and 580 k are observed via in-vacuo quadrupole mass spectrometry covering the range of 40 s–200 000 s after exposure.

The outgassing decay follows a single power law with exponents α  =  −0.7  to  −1.1 at 293 K, but at 580 K a drop from α  =  −0.25 to  −2.35 is found. For DEMO a pump-down time to 0.5 mPa in the order of 1–5 h can be expected. The outgassing is in all cases dominated by D₂.

The impact of lithium (Li) on plasma performance was investigated at the ASDEX Upgrade tokamak, which features a full tungsten wall. Li pellets containing 1.6  ×  10²⁰ Li atoms were launched with a speed of 600 m s⁻¹ to achieve deep penetration into the plasma and minimize the impact on the first wall. Homogeneous transient Li concentrations in the plasma of up to 15% were established. The Li sustainment time in the plasma decreased with an increasing heating power from 150 to 40 ms. Due to the pellet rate being restricted to 2 Hz, no Li pile-up could take place. No significant positive impact on plasma properties, as reported from other tokamak devices, could be found; the Li pellets rather caused a small reduction in plasma energy, mainly due to enhanced radiation. Due to pellet injection, a short-lived Li layer was formed on the plasma-facing components, which lasted a few discharges and led to moderately beneficial effects during plasma start-up. Most pellets were found to trigger type-I ELMs, either by their direct local perturbation or indirectly by the altered edge conditions; however, reliability was less than 100%.

The ion temperature gradient (ITG) modes in transport barriers (TBs) of tokamak plasmas are numerically studied with a code solving gyrokinetic integral eigenvalue equations in toroidal configurations. It is found that multiple ITG modes with conventional and unconventional ballooning mode structures can be excited simultaneously in TBs with steep gradients of ion temperature and density. The characteristics of the modes, including the dependence of the mode frequencies, growth rate and structure on plasma parameters, are systematically investigated. Unconventional modes with large mode-number $l$ (where $l$ denotes a certain parity and peak number in ballooning space) dominate in the large ${{k}_{\theta}}{{\rho}_{s}}$ region (${{k}_{\theta}}{{\rho}_{s}}\geqslant 1.2$), while the conventional mode with $l=0$ dominates in the medium ${{k}_{\theta}}{{\rho}_{s}}$ region ($0.4\leqslant {{k}_{\theta}}{{\rho}_{s}}<1.2$), and unconventional modes with small mode-number $l$ dominate in the small ${{k}_{\theta}}{{\rho}_{s}}$ region (${{k}_{\theta}}{{\rho}_{s}}<0.4$). Thus, the ${{k}_{\theta}}{{\rho}_{s}}$ spectra of these conventional and unconventional modes at steep gradients are qualitatively different from those of the conventional ITG modes at small or medium gradients, in which the growth rate of the only ITG mode with $l=0$ reaches maximum at the medium value ${{k}_{\theta}}{{\rho}_{s}}=0.6$. Through scanning ion temperature gradient ${{\varepsilon}_{T\text{i}}}$ and density gradient ${{\varepsilon}_{n}}$ separately, it is proven that the synergetic effect of ${{\varepsilon}_{T\text{i}}}$ and ${{\varepsilon}_{n}}$, rather than ${{\varepsilon}_{T\text{i}}}$ alone, drives the unconventional ITG modes in TBs. Moreover, it is found that the critical value of ${{\varepsilon}_{n}}$ for driving the unconventional ITG modes with large l number increases with increasing ${{k}_{\theta}}{{\rho}_{s}}$. In addition, the effects of magnetic shear on conventional and unconventional ITG modes in the high confinement regime (H-mode) are analyzed in detail, and compared with equivalent effects on conventional modes in the low and intermediate gradient regimes (L- and I- modes). Finally, the effects of the poloidal wave number and gradients of ion temperature and density on radial transport are analyzed based on quasi-linear mixing length estimations.

The potential for damage, the magnitude of the extrapolation, and the importance of the atypical—incidents that occur once in a thousand shots—make theory and simulation essential for ensuring that relativistic runaway electrons will not prevent ITER from achieving its mission. Most of the theoretical literature on electron runaway assumes magnetic surfaces exist. ITER planning for the avoidance of halo and runaway currents is focused on massive-gas or shattered-pellet injection of impurities. In simulations of experiments, such injections lead to a rapid large-scale magnetic-surface breakup. Surface breakup, which is a magnetic reconnection, can occur on a quasi-ideal Alfvénic time scale when the resistance is sufficiently small. Nevertheless, the removal of the bulk of the poloidal flux, as in halo-current mitigation, is on a resistive time scale. The acceleration of electrons to relativistic energies requires the confinement of some tubes of magnetic flux within the plasma and a resistive time scale. The interpretation of experiments on existing tokamaks and their extrapolation to ITER should carefully distinguish confined versus unconfined magnetic field lines and quasi-ideal versus resistive evolution. The separation of quasi-ideal from resistive evolution is extremely challenging numerically, but is greatly simplified by constraints of Maxwell's equations, and in particular those associated with magnetic helicity. The physics of electron runaway along confined magnetic field lines is clarified by relations among the poloidal flux change required for an e-fold in the number of electrons, the energy distribution of the relativistic electrons, and the number of relativistic electron strikes that can be expected in a single disruption event.

In the last few years, long-pulse H-mode plasma discharges (with small edge-localized modes and normalized beta, β_N ~ 1) have been realized at the Experimental Advanced Superconducting Tokamak (EAST). This paper reports on high-β_N (>1.5) discharges in the 2015 EAST campaign. The characteristics of these H-mode plasmas have been presented in a database. Analysis of the experimental limit of β_N has revealed several main features of typical discharges. Firstly, efficient, stable high heating power is required. Secondly, control of impurity radiation (partly due to interaction between the plasma and the in-vessel components) is also a critical issue for the maintenance of high-β_N discharges. In addition an internal transport barrier (ITB) has recently been observed in EAST, introducing further improvement in confinement surpassing H-mode plasmas. ITB dynamics is another key issue for high-β_N plasmas in EAST. Each of these features is discussed in this paper. Study and improvement of these issues could be considered as the key to achieving long-pulse high-β_N operation with EAST.

During the first operational phase (referred to as OP1.1) of the new Wendelstein 7-X (W7-X) stellarator, five poloidal graphite limiters were mounted on the inboard side of the vacuum vessel, one in each of the five toroidal modules which form the W7-X vacuum vessel. Each limiter consisted of nine specially shaped graphite tiles, designed to conform to the last closed field line geometry in the bean-shaped section of the standard OP1.1 magnetic field configuration (Sunn Pedersen et al 2015 Nucl. Fusion 55 126001). We observed the limiters with multiple infrared and visible camera systems, as well as filtered photomultipliers. Power loads are calculated from infrared (IR) temperature measurements using THEODOR, and heating patterns (dual stripes) compare well with field line mapping and EMC3-EIRENE predictions. While the poloidal symmetry of the heat loads was excellent, the toroidal heating pattern showed up to a factor of 2×  variation, with peak heat loads on Limiter 1. The total power intercepted by the limiters was up to ~60% of the input ECRH heating power. Calorimetry using bulk tile heating (measured via post-shot IR thermography) on Limiter 3 showed a difference between short high power discharges, and longer lower power ones, with regards to the fraction of energy deposited on the limiters. Finally, fast heating transients, with frequency  >1 kHz were detected, and their visibility was enhanced by the presence of surface coatings which developed on the limiters by the end of the campaign.