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Tokamak à configuration variable (TCV), recently celebrating 30 years of near-continual operation, continues in its missions to advance outstanding key physics and operational scenario issues for ITER and the design of future power plants such as DEMO. The main machine heating systems and operational changes are first described. Then follow five sections: plasma scenarios. ITER Base-Line (IBL) discharges, triangularity studies together with X3 heating and N2 seeding. Edge localised mode suppression, with a high radiation region near the X-point is reported with N _2 injection with and without divertor baffles in a snowflake configuration. Negative triangularity (NT) discharges attained record, albeit transient, β _N ∼ 3 with lower turbulence, higher low-Z impurity transport, vertical stability and density limits and core transport better than the IBL. Positive triangularity L-Mode linear and saturated ohmic confinement confinement saturation, often-correlated with intrinsic toroidal rotation reversals, was probed for D, H and He working gases. H-mode confinement and pedestal studies were extended to low collisionality with electron cyclotron heating obtaining steady state electron iternal transport barrier with neutral beam heating (NBH), and NBH driven H-mode configurations with off-axis co-electron cyclotron current drive. Fast particle physics. The physics of disruptions, runaway electrons and fast ions (FIs) was developed using near-full current conversion at disruption with recombination thresholds characterised for impurity species (Ne, Ar, Kr). Different flushing gases (D2, H2) and pathways to trigger a benign disruption were explored. The 55 kV NBH II generated a rich Alfvénic spectrum modulating the FI fas ion loss detector signal. NT configurations showed less toroidal Alfvén excitation activity preferentially affecting higher FI pitch angles. Scrape-off layer and edge physics. gas puff imaging systems characterised turbulent plasma ejection for several advanced divertor configurations, including NT. Combined diagnostic array divertor state analysis in detachment conditions was compared to modelling revealing an importance for molecular processes. Divertor physics. Internal gas baffles diversified to include shorter/longer structures on the high and/or low field side to probe compressive efficiency. Divertor studies concentrated upon mitigating target power, facilitating detachment and increasing the radiated power fraction employing alternative divertor geometries, optimised X-point radiator regimes and long-legged configurations. Smaller-than-expected improvements with total flux expansion were better modelled when including parallel flows. Peak outer target heat flux reduction was achieved (>50%) for high flux-expansion geometries, maintaining core performance ( H _98 > 1). A reduction in target heat loads and facilitated detachment access at lower core densities is reported. Real-time control. TCV’s real-time control upgrades employed MIMO gas injector control of stable, robust, partial detachment and plasma β feedback control avoiding neoclassical tearing modes with plasma confinement changes. Machine-learning enhancements include trajectory tracking disruption proximity and avoidance as well as a first-of-its-kind reinforcement learning-based controller for the plasma equilibrium trained entirely on a free-boundary simulator. Finally, a short description of TCV’s immediate future plans will be given.

Electron–cyclotron waves are a tool commonly used in tokamaks, in particular to drive current. Their ability to drive current in a very localized manner renders them an optimal tool for MHD mode mitigation. However, such applications require high accuracy and good control of the power deposition location to efficiently target the magnetic islands. It has been indirectly observed that the suprathermal electron distribution, resulting from the wave absorption, is broader than what is expected from experimentally-constrained forward drift-kinetic modeling. The present paper explores the possibility that beam scattering through the turbulent edge of the plasma may explain this observed discrepancy. In particular, full-wave studies exhibit three beam broadening regimes, from superdiffusive to diffusive, with an intermediate regime characterized by a Lorentzian beam profile with a slightly increased full-width at half maximum with respect to the quiet plasma case. In the tokamak à configuration variable, dedicated plasma scenarios have been developed to test this hypothesis. A realistic worst-case fluctuation scenario falls into this intermediate beam broadening regime. By comparing the experimental hard x-ray emission from suprathermal electron Bremmstrahlung with the emission calculated by coupling a full-wave model to a Fokker–Planck solver, it is shown that, in the tested cases, the beam broadening is not sufficient to explain the aforementioned discrepancy between simulation and experiment and that another mechanism must play the main role in broadening the suprathermal electron distribution.

Experiments on ASDEX Upgrade (AUG) in 2021 and 2022 have addressed a number of critical issues for ITER and EU DEMO. A major objective of the AUG programme is to shed light on the underlying physics of confinement, stability, and plasma exhaust in order to allow reliable extrapolation of results obtained on present day machines to these reactor-grade devices. Concerning pedestal physics, the mitigation of edge localised modes (ELMs) using resonant magnetic perturbations (RMPs) was found to be consistent with a reduction of the linear peeling-ballooning stability threshold due to the helical deformation of the plasma. Conversely, ELM suppression by RMPs is ascribed to an increased pedestal transport that keeps the plasma away from this boundary. Candidates for this increased transport are locally enhanced turbulence and a locked magnetic island in the pedestal. The enhanced D-alpha (EDA) and quasi-continuous exhaust (QCE) regimes have been established as promising ELM-free scenarios. Here, the pressure gradient at the foot of the H-mode pedestal is reduced by a quasi-coherent mode, consistent with violation of the high-n ballooning mode stability limit there. This is suggestive that the EDA and QCE regimes have a common underlying physics origin. In the area of transport physics, full radius models for both L- and H-modes have been developed. These models predict energy confinement in AUG better than the commonly used global scaling laws, representing a large step towards the goal of predictive capability. A new momentum transport analysis framework has been developed that provides access to the intrinsic torque in the plasma core. In the field of exhaust, the X-Point Radiator (XPR), a cold and dense plasma region on closed flux surfaces close to the X-point, was described by an analytical model that provides an understanding of its formation as well as its stability, i.e., the conditions under which it transitions into a deleterious MARFE with the potential to result in a disruptive termination. With the XPR close to the divertor target, a new detached divertor concept, the compact radiative divertor, was developed. Here, the exhaust power is radiated before reaching the target, allowing close proximity of the X-point to the target. No limitations by the shallow field line angle due to the large flux expansion were observed, and sufficient compression of neutral density was demonstrated. With respect to the pumping of non-recycling impurities, the divertor enrichment was found to mainly depend on the ionisation energy of the impurity under consideration. In the area of MHD physics, analysis of the hot plasma core motion in sawtooth crashes showed good agreement with nonlinear 2-fluid simulations. This indicates that the fast reconnection observed in these events is adequately described including the pressure gradient and the electron inertia in the parallel Ohm’s law. Concerning disruption physics, a shattered pellet injection system was installed in collaboration with the ITER International Organisation. Thanks to the ability to vary the shard size distribution independently of the injection velocity, as well as its impurity admixture, it was possible to tailor the current quench rate, which is an important requirement for future large devices such as ITER. Progress was also made modelling the force reduction of VDEs induced by massive gas injection on AUG. The H-mode density limit was characterised in terms of safe operational space with a newly developed active feedback control method that allowed the stability boundary to be probed several times within a single discharge without inducing a disruptive termination. Regarding integrated operation scenarios, the role of density peaking in the confinement of the ITER baseline scenario (high plasma current) was clarified. The usual energy confinement scaling ITER98( p,y ) does not capture this effect, but the more recent H20 scaling does, highlighting again the importance of developing adequate physics based models. Advanced tokamak scenarios, aiming at large non-inductive current fraction due to non-standard profiles of the safety factor in combination with high normalised plasma pressure were studied with a focus on their access conditions. A method to guide the approach of the targeted safety factor profiles was developed, and the conditions for achieving good confinement were clarified. Based on this, two types of advanced scenarios (‘hybrid’ and ‘elevated’ q -profile) were established on AUG and characterised concerning their plasma performance.

Experimental measurements of the turbulence amplitude utilizing a short pulse reflectometry method are presented. Two discharges with shaped plasma possessing opposite signs of triangularity are considered and a higher turbulence amplitude is found in the positive triangularity case. To confirm this result, a synthetic short pulse reflectometry diagnostic is developed. Local gyrokinetic modeling with the GENE code is carried out to produce turbulence relevant to the experimental conditions. This turbulence is then used as an input for the full-wave CUWA code to produce synthetic short pulse reflectometry signals. By matching synthetic and experimental reflectometry data, the difference between turbulence amplitudes in the two cases is confirmed. Additionally, the capability of the diagnostic to also measure the frequency spectrum of the turbulence is demonstrated.

The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q ⩾ 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant β phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent ‘pedestal-to-divertor’ integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T _e and radiation which need control. High- β _P scenarios with q _min > 2, q _95 –7.9, β _N > 4, β _T –3.3% and H _98y2 > 1.5 are sustained with high density ( $\bar n$ = 7E19 m ^−3 , f _G –1) for 6 τ _E , improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors.

Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.

Superconducting (SC) tokamak JT-60SA plays an essential role in fusion research and development by supporting and complementing the ITER project, providing directions to the DEMO design activity and fostering next generation scientists and engineers. Since the short circuit incident at the terminal joints of equilibrium field coil #1 during the integrated commissioning (IC) in March 2021, both EU and JA implementing agencies (IAs) have examined how to ensure safe operation of JT-60SA by mitigating the risk of possible discharge occurrence inside the cryostat. Based on the experience of the global Paschen tests, the IAs have established a strategy of risk mitigation measures, which is a combination of (i) reinforcement of insulation, (ii) avoiding unnecessary voltage application to the coil systems and (iii) immediate de-energization of the coils when deteriorated vacuum conditions are detected. Thanks to the considerable efforts of the Integrated Project Team members, the IC restarted in May 2023. After confirmation of the SC state of the coil systems (TF, EF and CS), the coil energization test and the plasma operation phase 1 (OP-1) started. The first plasma was successfully achieved on 23 October 2023 with a limited value of voltage and current applied to the coils. The plasma configuration control was also confirmed with low plasma current and low auxiliary heating power conditions. Based on the IO–F4E–QST collaboration, activities of JT-60SA have been shared with the IO and provided an important lesson for ITER assembly and commissioning, and will provide an outstanding contribution to fusion research at large. After OP-1, maintenance & enhancement phase 1 (M/E-1) starts from January 2024, in which in-vessel components are installed, and heating and diagnostic systems are extensively upgraded to allow a high power heating experiment planned in OP-2. In order to make the best use of JT-60SA, a newly organized JT-60SA experiment team will refine the research plan for the future high heating power operation phase.

Based on a gyrokinetic analysis of and extrapolation from TCV discharges with large negative and positive triangularity δ , the potential of extreme $|\delta|$ in reducing turbulent transport is assessed. Linearly, both positive and negative δ can exert a stabilizing influence, with substantial sensitivity to the radial wavenumber k _x . Nonlinear fluxes are reduced at extreme δ in a trapped-electron-mode regime, whereas low-amplitude ion-temperature-gradient turbulence is boosted by large negative δ . Focusing on the former case, nonlinear fluxes exceed quasilinear ones at negative δ , a trend that reverses as δ > 0. A change in saturation efficiency is the cause of these features: the zonal-flow residual is boosted at δ > 0, reducing fluxes compared with the linear drive as δ is increased, and a shift towards larger zonal-flow scales occurs with increasing δ due to finite- k _x modes weakening with δ .

Experiments, gyrokinetic simulations and transport predictions were performed to investigate if a negative triangularity (NT) L-mode option for the Divertor Tokamak Test (DTT) full-power scenario would perform similarly to the positive triangularity (PT) H-mode reference scenario, avoiding the harmful edge localized modes (ELMs). The simulations show that a beneficial effect of NT coming from the edge/scrape-off layer (SOL) region $\rho_{\mathrm{tor}} \gt 0.9$ is needed to allow the actual NT L-mode option to perform like the PT H-mode. Dedicated experiments at TCV and AUG, with DTT-like shapes, show an optimistic picture. In TCV, experiments indicate that even with the relatively small triangularity of the DTT NT scenario, a large beneficial effect of NT comes from the plasma edge and SOL, allowing NT L-modes to outperform PT L-modes with the same power input, reaching the same central pressures as PT H-modes with twice as much applied heating power. For AUG, NT plasmas go into H-mode more easily than for TCV, but always present much smaller pedestals compared with PT plasmas with the same input power, showing a much weaker or absent ELM activity. However, NT has a smaller beneficial effect for AUG than for TCV, with NT pulses outperforming PT pulses with the same input power only for an ECRH-only case with relatively low input power. For the considered AUG cases, PT pulses perform better than NT ones at higher ECRH power or with mixed NBI and ECRH power. Based on this analysis, the NT option is a viable alternative for the DTT full power scenario, providing high performance plasmas with reduced or absent ELMs.

Runaway electrons (REs) are a concern for tokamak fusion reactors from discharge startup to termination. A sudden localized loss of a multi-megaampere RE beam can inflict severe damage to the first wall. Should a disruption occur, the existence of a RE seed may play a significant role in the formation of a RE beam and the magnitude of its current. The application of central electron cyclotron resonance heating (ECRH) in the Tokamak à Configuration Variable (TCV) reduces an existing RE seed population by up to three orders of magnitude within only a few hundred milliseconds. Applying ECRH before a disruption can also prevent the formation of a post-disruption RE beam in TCV where it would otherwise be expected. The RE expulsion rate and consequent RE current reduction are found to increase with applied ECRH power. Whereas central ECRH is effective in expelling REs, off-axis ECRH has a comparatively limited effect. A simple 0-D model for the evolution of the RE population is presented that explains how the effective ECRH-induced RE expulsion results from the combined effects of increased electron temperature and enhanced RE transport.

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international C oordination on I nternational C hallenges on L ong duration OP eration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed.

Local gyrokinetic simulations are used to model turbulent transport for the first time in a representative high-performance plasma discharge projected for the new JT-60SA tokamak. The discharge features a double-null separatrix, 41 MW of combined neutral beam heating and electron cyclotron heating, and a high predicted ratio of the normalized plasma kinetic to magnetic pressure β . When considering input parameters computed from reduced transport models, gyrokinetic simulations predict a turbulent heat flux well below the injected 41 MW. Increasing the background gradients, on the other hand, can trigger a non-zonal transition (NZT), causing heat fluxes to no longer saturate. Furthermore, when considering fast ions in the simulations, a high-frequency mode is destabilized that substantially impacts the turbulence. The NZT is avoided by reducing the electron pressure by 10% below its nominal value, and the fast-ion resonance is removed by reducing the fast-ion temperature. The thus-obtained simulation features broadband frequency spectra and density and temperature fluctuation levels $\delta n/n \approx 1\%$ –2%, $\delta T/T \approx 1\%$ –6% that should be measurable with fluctuation diagnostics planned for JT-60SA. The temperature profile is fixed by the critical main-ion temperature gradient as a consequence of the high stiffness; heat fluxes increase by a factor of ten when increasing the main ion temperature gradient by 17%. Despite large gradients, it is demonstrated that, due to the large β , retaining compressional magnetic field fluctuations and in particular, the contribution of the pressure gradient in the $\nabla B$ drifts, is crucial to achieving non-zero heat fluxes.

An analytical 1D model based on the WKB approximation is used to study the properties of a short pulse reflectometry diagnostic in a magnetic fusion device. Expressions linking the pulse delay with the parameters of the turbulence near the cut-off layer are derived for both ordinary and extraordinary polarizations of the probing beam. These results are used to develop a method for measuring the turbulence amplitude and the radial correlation length. The analytical conclusions and the proposed method are validated using full-wave numerical modelling. The latter is also used to study the limitations of the method and potential experimental effects not included in the reduced model.

Through a diverse research programme, the Tokamak à Configuration Variable (TCV) addresses physics issues and develops tools for ITER and for the longer term goals of nuclear fusion, relying especially on its extreme plasma shaping and electron cyclotron resonance heating (ECRH) launching flexibility and preparing for an ECRH and NBI power upgrade. Localized edge heating was unexpectedly found to decrease the period and relative energy loss of edge localized modes (ELMs). Successful ELM pacing has been demonstrated by following individual ELM detection with an ECRH power cut before turning the power back up to trigger the next ELM, the duration of the cut determining the ELM period. Negative triangularity was also seen to reduce the ELM energy release. H-mode studies have focused on the L–H threshold dependence on the main ion species and on the divertor leg length. Both L- and H-modes have been explored in the snowflake configuration with emphasis on edge measurements, revealing that the heat flux to the strike points on the secondary separatrix increases as the X-points approach each other, well before they coalesce. In L-mode, a systematic scan of the auxiliary power deposition profile, with no effect on confinement, has ruled it out as the cause of confinement degradation. An ECRH power absorption observer based on transmitted stray radiation was validated for eventual polarization control. A new profile control methodology was introduced, relying on real-time modelling to supplement diagnostic information; the RAPTOR current transport code in particular has been employed for joint control of the internal inductance and central temperature. An internal inductance controller using the ohmic transformer has also been demonstrated. Fundamental investigations of neoclassical tearing mode (NTM) seed island formation by sawtooth crashes and of NTM destabilization in the absence of a sawtooth trigger were carried out. Both stabilizing and destabilizing agents (electron cyclotron current drive on or inside the q = 1 surface, respectively) were used to pace sawtooth oscillations, permitting precise control of their period. Locking of the sawtooth period to a pre-defined ECRH modulation period was also demonstrated. Sawtooth control has permitted nearly failsafe NTM prevention when combined with backup NTM stabilization by ECRH. [ABSTRACT FROM AUTHOR]

The effects of negative triangularity () on confinement and fluctuations in plasmas covering a large range of parameters were investigated on the tokamak à configuration variable (TCV). The conditions explored in this paper include discharges where neutral beam (NB) heating was employed to obtain an electron-ion temperature ratio across a large fraction of the plasma profile. This significantly extended the range of negative plasmas studied on TCV towards conditions more relevant to future reactor-like tokamaks. Negative triangularity was found to improve confinement over the full range of collisionality studied () and (). The amplitude of radiative temperature fluctuations, measured using a correlation electron cyclotron emission diagnostic over the range , was found to be reduced, in negative with respect to positive plasmas, for all combinations of parameters explored. This was, in particular, verified for a pair of positive and negative plasmas with comparable density and under different conditions of NB heating. Linear gyrokinetic simulations found the dominant turbulence regime, in the strongly NB heated discharges, to be a mixture of trapped electron modes (TEMs) and ion temperature gradient driven modes. This is in contrast to ohmic or electron cyclotron heated discharges, for which the dominant turbulence regime was found to be pure TEM. Negative triangularity was found to lead to partial stabilization of the most unstable modes for low wavenumbers in both turbulence regimes. These findings demonstrate that negative triangularity could provide significant confinement improvement over a large range of parameters, that include conditions closer to future reactor-like machines (, low collisionality). [ABSTRACT FROM AUTHOR]

The novel plasma position and shape controller of the TCV tokamak aids in the precise control of complex configurations such as snowflake (SF) divertors. The ability of the controller design to simultaneously control the position, shape, divertor leg and X-points is extended to various SF plasma configurations. The unique feature of ordering the controlled variables from the most easily to the least easily controlled, while respecting the hardware limits on the poloidal-field coil currents, is exploited in particular to provide reliable control of SF equilibria with closely spaced X-points, approaching the exact snowflake. The off-line optimisation of the control parameters and their experimental implementation on snowflake plasma configuration is reported. The performance of the new controller is compared with the legacy TCV analogue controller for a SF plasma discharge. The new controller is successful in providing a marked improvement over the legacy TCV controller with respect to , the main parameter of interest for this study. [ABSTRACT FROM AUTHOR]

The research program of the TCV tokamak ranges from conventional to advanced-tokamak scenarios and alternative divertor configurations, to exploratory plasmas driven by theoretical insight, exploiting the device’s unique shaping capabilities. Disruption avoidance by real-time locked mode prevention or unlocking with electron-cyclotron resonance heating (ECRH) was thoroughly documented, using magnetic and radiation triggers. Runaway generation with high-Z noble-gas injection and runaway dissipation by subsequent Ne or Ar injection were studied for model validation. The new 1 MW neutral beam injector has expanded the parameter range, now encompassing ELMy H-modes in an ITER-like shape and nearly non-inductive H-mode discharges sustained by electron cyclotron and neutral beam current drive. In the H-mode, the pedestal pressure increases modestly with nitrogen seeding while fueling moves the density pedestal outwards, but the plasma stored energy is largely uncorrelated to either seeding or fueling. High fueling at high triangularity is key to accessing the attractive small edge-localized mode (type-II) regime. Turbulence is reduced in the core at negative triangularity, consistent with increased confinement and in accord with global gyrokinetic simulations. The geodesic acoustic mode, possibly coupled with avalanche events, has been linked with particle flow to the wall in diverted plasmas. Detachment, scrape-off layer transport, and turbulence were studied in L- and H-modes in both standard and alternative configurations (snowflake, super-X, and beyond). The detachment process is caused by power ‘starvation’ reducing the ionization source, with volume recombination playing only a minor role. Partial detachment in the H-mode is obtained with impurity seeding and has shown little dependence on flux expansion in standard single-null geometry. In the attached L-mode phase, increasing the outer connection length reduces the in–out heat-flow asymmetry. A doublet plasma, featuring an internal X-point, was achieved successfully, and a transport barrier was observed in the mantle just outside the internal separatrix. In the near future variable-configuration baffles and possibly divertor pumping will be introduced to investigate the effect of divertor closure on exhaust and performance, and 3.5 MW ECRH and 1 MW neutral beam injection heating will be added. [ABSTRACT FROM AUTHOR]

The role of the pedestal position on the pedestal performance has been investigated in AUG, JET-ILW and TCV. When the pedestal is peeling-ballooning (PB) limited, the three machines show a similar behaviour. The outward shift of the pedestal density relative to the pedestal temperature can lead to the outward shift of the pedestal pressure which, in turns, reduces the PB stability, degrades the pedestal confinement and reduces the pedestal width. Once the experimental density position is considered, the EPED model is able to correctly predict the pedestal height. An estimate of the impact of the density position on a ITER baseline scenario shows that the maximum reduction in the pedestal height is 10% while the reduction in the fusion power is between 10% and 40% depending on the assumptions for the core transport model used. In other plasmas, where the pedestal density is shifted even more outwards relative to the pedestal temperature, the pedestal does not seem PB limited and a different behaviour is observed. The outward shift of the density is still empirically correlated with the pedestal degradation but no change in the pressure position is observed and the PB model is not able to correctly predict the pedestal height. On the other hand, the outward shift of the density leads to a significant increase of ηe and ηi (where ηe,i is the ratio of density to temperature scale lengths, ) which leads to the increase of the growth rate of microinstabilities (mainly ETG and ITG) by 50%. This suggests that, in these plasmas, the increase in the turbulent transport due to the outward shift of the density might play an important role in the decrease of the pedestal performance. [ABSTRACT FROM AUTHOR]

The effects of plasma shaping, in particular triangularity (δ), on temperature fluctuations in the tokamak à configuration variable have been studied using the correlation electron cyclotron emission system. It has been found that, in ohmic discharges with comparable density profiles, the relative fluctuation level measured at the edge is significantly reduced in plasmas with negative triangularity with respect to positive triangularity ones. Additionally, the critical temperature gradients for the onset of turbulence are observed to increase in the negative triangularity plasmas. An estimation of the correlation length of the fluctuating structure shows smaller structure size in the negative triangularity cases, which are known to be associated with improved confinement. Together, these observations suggest that changing triangularity from positive to negative strongly influences the nature of the turbulent fluctuations excited in the plasma. [ABSTRACT FROM AUTHOR]

JT-60SA envisions electron cyclotron wall conditioning (ECWC), as wall conditioning method in the presence of the toroidal field to control fuel and impurity recycling and to improve plasma performance and reproducibility. This paper reports on Helium ECWC experiments on TCV in support of JT-60SA operation. Nearly sixty Helium conditioning discharges have been successfully produced in TCV, at a toroidal field BT  =  1.3 or 1.54 T, with gyrotrons at 82.7 GHz in X2 mode, mimicking ECWC operation in JT-60SA at the second harmonic of the EC wave. Discharge parameters were tuned in order to (i) minimize the time for the onset of ECWC plasmas, thus minimizing absorption of stray radiation by in-vessel components, (ii) improve discharge homogeneity by extending the discharge vertically and radially, and wall coverage, in particular of inboard surfaces where JT-60SA plasmas will be initiated, (iii) assess the efficiency of He-ECWC to deplete carbon walls from fuel. An optimized combination of vertical and radial magnetic fields, with amplitudes typically 0.1 to 0.6% of that of BT, has been determined, which resulted in lowest breakdown time, improved wall coverage and enhanced fuel removal. A standard ohmic D2-plasma could be then sustained, whereas it would not have been possible without He-ECWC. [ABSTRACT FROM AUTHOR]

This paper presents a survey of the experiments on runaway electrons (RE) carried out recently in frames of EUROFusion Consortium in different tokamaks: COMPASS, ASDEX-Upgrade, TCV and JET. Massive gas injection (MGI) has been used in different scenarios for RE generation in small and medium-sized tokamaks to elaborate the most efficient and reliable ones for future RE experiments. New data on RE generated at disruptions in COMPASS and ASDEX-Upgrade was collected and added to the JET database. Different accessible parameters of disruptions, such as current quench rate, conversion rate of plasma current into runaways, etc have been analysed for each tokamak and compared to JET data. It was shown, that tokamaks with larger geometrical sizes provide the wider limits for spatial and temporal variation of plasma parameters during disruptions, thus extending the parameter space for RE generation. The second part of experiments was dedicated to study of RE generation in stationary discharges in COMPASS, TCV and JET. Injection of Ne/Ar have been used to mock-up the JET MGI runaway suppression experiments. Secondary RE avalanching was identified and quantified for the first time in the TCV tokamak in RE generating discharges after massive Ne injection. Simulations of the primary RE generation and secondary avalanching dynamics in stationary discharges has demonstrated that RE current fraction created via avalanching could achieve up to 70–75% of the total plasma current in TCV. Relaxations which are reminiscent the phenomena associated to the kinetic instability driven by RE have been detected in RE discharges in TCV. Macroscopic parameters of RE dominating discharges in TCV before and after onset of the instability fit well to the empirical instability criterion, which was established in the early tokamaks and examined by results of recent numerical simulations. [ABSTRACT FROM AUTHOR]

A novel plasma position and shape controller has been developed for the highly flexible shaping poloidal-field coil set of the TCV tokamak, to aid in the precise control of advanced configurations such as negative-triangularity plasmas, snowflake and super-X divertors, and doublets. This work follows and relies on the deployment of a new, sub-ms, real-time magnetic equilibrium-reconstruction algorithm. The controller formulation ensures flexibility through an ordering of controlled variables from the most easily to the least easily controlled, while respecting the hardware limits on the poloidal-field coil currents. A rigid, linearised plasma response model for the TCV tokamak is used for the verification and determination of the control parameters. The controller has been applied successfully to a variety of TCV plasma discharges. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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 . 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. [ABSTRACT FROM AUTHOR]

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised. [ABSTRACT FROM AUTHOR]

A key feature of the new digital plasma control system installed on the TCV (Tokamak à Configuration Variable) tokamak is its possibility to rapidly design, test and deploy real-time algorithms. It accommodates hundreds of diagnostic inputs and actuator outputs, and offers the possibility to design advanced control algorithms with better knowledge of the plasma state and to coherently control all TCV actuators, including poloidal field coils, gas valves, the gyrotron powers and launcher angles of the electron cyclotron heating and current drive system together with diagnostic triggering signals. It encompasses plasma control applications ranging from basic experiments of coil current and density control to advanced experiments of magnetohydrodynamics (MHD) and plasma profile control. The system consists of multiple nodes, each of which may have a local analog to digital (ADC) and/or digital to analog (DAC) card; all nodes are connected to a reflective memory (RFM), providing a deterministic method of sharing memory between them. Recently, a generalized plasma position and shape controller based on the real-time (RT) Grad-Shafranov solver RTLIUQE was developed and implemented, providing the basis for future high performance plasma operation with advanced plasma configurations. The controller design is based on an isoflux control scheme and utilizes singular value decomposition (SVD), to respect the limits on poloidal field coils currents by limiting the controlled parameters to the set that can be more easily controlled. The controller is capable in principle of providing improved equilibrium control especially for unconventional plasma scenarios, for e.g. reliable control of ‘snowflake’ equilibria with closely spaced x-points, i.e. the ‘exact’ snowflake, and the development of negative triangularity plasmas in H-mode. An addition of a new node on the digital control system has enhanced the real time computational capacity and hosts the real-time transport code RAPTOR (rapid plasma transport simulator), an advanced density profile reconstruction algorithm including real-time fringe jump correction, as well as a plasma state monitoring, supervision and actuator management algorithm. In future, more signals from existing TCV diagnostics, including multiview pinhole x-ray diagnostics, Thomson scattering, visible image processing and magnetic signals for MHD mode analysis will be added to expand the capabilities of the digital control system. [ABSTRACT FROM AUTHOR]

TCV is acquiring a new 1 MW neutral beam and 2 MW additional third-harmonic electron cyclotron resonance heating (ECRH) to expand its operational range. Its existing shaping and ECRH launching versatility was amply exploited in an eclectic 2013 campaign. A new sub-ms real-time equilibrium reconstruction code was used in ECRH control of NTMs and in a prototype shape controller. The detection of visible light from the plasma boundary was also successfully used in a position-control algorithm. A new bang-bang controller improved stability against vertical displacements. The RAPTOR real-time transport simulator was employed to control the current density profile using electron cyclotron current drive. Shot-by-shot internal inductance optimization was demonstrated by iterative learning control of the current reference trace. Systematic studies of suprathermal electrons and ions in the presence of ECRH were performed. The L–H threshold power was measured to be ∼50–75% higher in both H and He than D, to increase with the length of the outer separatrix, and to be independent of the current ramp rate. Core turbulence was found to decrease from positive to negative edge triangularity deep into the core. The geodesic acoustic mode was studied with multiple diagnostics, and its axisymmetry was confirmed by a full toroidal mapping of its magnetic component. A new theory predicting a toroidal rotation component at the plasma edge, driven by inhomogeneous transport and geodesic curvature, was tested successfully. A new high-confinement mode (IN-mode) was found with an edge barrier in density but not in temperature. The edge gradients were found to govern the scaling of confinement with current, power, density and triangularity. The dynamical interplay of confinement and magnetohydrodynamic modes leading to the density limit in TCV was documented. The heat flux profile decay lengths and heat load profile on the wall were documented in limited plasmas. In the snowflake (SF) divertor configuration the heat flux profiles were documented on all four strike points. SF simulations with the EMC3-EIRENE code, including the physics of the secondary separatrix, underestimate the flux to the secondary strike points, possibly resulting from steady-state E × B drifts. With neon injection, radiation in a SF was 15% higher than in a conventional divertor. The novel triple-null and X-divertor configurations were also achieved in TCV. [ABSTRACT FROM AUTHOR]

For a comprehensive comparison with theoretical models and advanced numerical turbulence simulations, a large spectrum of fluctuation parameters was measured on the devices ASDEX Upgrade, TCV, and Tore-Supra. Radial profiles of scale-resolved turbulence levels in H-mode discharges are measured and compared with GENE simulations in the transition range from ion-temperature-gradient to trapped-electron-mode turbulence. Correlation reflectometry is used to study the microscopic structure of turbulence and GAMs in discharges where poloidal flow damping was varied by means of variations of the shape of the poloidal plasma cross-section and collisionality. Full-wave codes and synthetic diagnostics are applied for the interpretation of the data. [ABSTRACT FROM AUTHOR]

The design of a snowflake (SF) equilibrium requires a strong effort on the poloidal field (PF) currents in terms of MAturns and mechanical loads. This has limited the maximum plasma current in SF configurations on Tokamak à Configuration Variable (TCV) to values well below the intrinsic magnetohydrodynamic limits. In this paper the definition of optimized SF configurations in TCV and their experimental tests are illustrated. The PF current optimization procedure proposed in Albanese et al (2014 Plasma Phys. Control. Fusion56 035008) is adapted and applied to a SF scenario in TCV where the PF currents were close to their operational limits with the aim of reducing the total MAturns in view of higher values of the plasma current. This procedure optimizes the PF currents while fulfilling the machine technological constraints for a given bound on the tolerable plasma shape changes. The method exploits the linearized relation between the plasma–wall gaps and the PF currents. In the investigated TCV scenario the optimization procedure allowed a 20% increase of the plasma current while keeping the plasma shape alignment with respect to the nominal shape within a tolerance of 1 cm. The predicted optimization potential was confirmed in a TCV experiment. [ABSTRACT FROM AUTHOR]