Tokamak a 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, beta(N) similar to 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 Alfvenic spectrum modulating the FI fas ion loss detector signal. NT configurations showed less toroidal Alfven 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 beta 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.

In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements-new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors-as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D-T operations is presented in (King et al Nucl. Fusion submitted).

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.

A set of dedicated H-mode discharges with constant heating power combining Neutral Beam Injection and Electron Cyclotron Resonance Heating have been executed at the ASDEX Upgrade tokamak using a high triangularity magnetic geometry in order to investigate the impact of filamentary transport to divertor and non-divertor components. The evolution of upstream scrape-off layer (SOL) profiles have been correlated with dedicated separatrix quantities, mostly with the turbulence control parameter alpha t (Eich and Manz 2021 Nucl. Fusion 61 086016) describing the turbulence level at the separatrix. With increasing alpha t , a broadening of the upstream density profiles in the near-SOL together with the formation of a density shoulder in the far-SOL have been observed. This phenomenon is associated with an enhanced filamentary transport dominating the radial turbulent transport in the far-SOL and confirmed by means of the cooling water calorimetry on non-divertor components. The probe measurements conducted with the ball-pen probe-head mounted on the midplane manipulator and a retarding-field analyzer close to the limiter surface indicate that the key mechanism increasing the radial filamentary transport to the first wall is an increase of the particle flux Gamma r,fil , caused primarily by the packing fraction fPF,fil and the filament density ne,fil . At the same time, the electron temperature Te and ion temperature Ti measured close to the limiter surface show only small variations above alpha t > 0.5. Both the filamentary heat flux and the gross erosion derived from the first wall probe measurements reach a magnitude that should be considered in the design of future fusion reactors.

The plasma electron density is measured with multiple diagnostics in the scrape-off layer of the ASDEX Upgrade tokamak. The recorded values are typically in good agreement in the case of exponentially decaying profiles at low densities, but they become inconsistent as soon as a density shoulder is formed, and up to a factor of 3 deviation is experienced at high densities. An idea is put forward; namely, that the local plasma-wall proximity has an influence on the local density and thus the measurement results depend on the local limiter-plasma and limiter-diagnostic distances at the same time. The effect of local wall clearance dependence is demonstrated by creating a plasma having an extremely high scrape-off layer density (n(e )> 3 & sdot;10(19)m(-3)) after the H-L back transition and varying the position of the separatrix in different directions. As a possible physical process, neutral recycling at the plasma facing components is identified. Similar behaviour is also observed in the QCE regime, where the appearance of small ELMs barely impacts the scrape-off layer profiles. There are also indications that the local wall clearance does have an impact on the density profile not only in the scrape-off layer, but also in the confined region.

Recent experiments at the ASDEX Upgrade tokamak have provided the first ever measurements from the imaging heavy-ion beam probe. In this work, we show that the developed simulation framework can reproduce qualitatively the measurement's observed shape and position. Quantitatively, we demonstrate that the model reproduces, within the experimental uncertainties, the observed signal levels. A detailed explanation of the synthetic model is presented, along with the calibration of the optical setup that reproduces the measurements.

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.

Over previous campaigns, an intense experimental program on advanced tokamak (AT) scenarios, has been carried out at the ASDEX Upgrade tokamak with full-tungsten wall. These discharges have been executed shortly after the boronization of the first wall to reduce the density and the impurity influx. The confinement level of such AT discharges was found to vary considerably, even when discharges with similar, if not identical, engineering parameters were carried out. This work investigates the causes of such confinement variations. Among all plasma quantities analyzed, confinement quality of AT scenarios correlates best with divertor neutral pressure, highlighting the key role of edge and scrape-off layer physics in determining global plasma confinement. In particular, it is found that the main cause of confinement degradation is the reduction of pedestal stability, which is in turn caused by the outward shift of the maximum density gradient position typically observed when the divertor neutral pressure increases. Owing to the low density of AT discharges under analysis, the movement of the maximum density gradient position can be caused entirely by changes in deuterium outgassing from the wall, which is strongly influenced by the boron layer deposited on the plasma-facing components and by the deuterium wall inventory. Finally, the predictive capability of confinement quality with the integrated model IMEP [Luda et al., Nucl. Fusion 60, 036023 (2020)] is tested on these discharges and shows promising results.

In order to extend the enhanced D-Alpha H-mode to future devices, it is crucial to understand the properties of the main signature of this regime, the quasicoherent mode (QCM), that likely clamps the pressure gradient below the ideal magnetohydrodynamic limit. The turbulent character of the QCM is investigated with scanning probes in ASDEX Upgrade. Analysis reveals the multi-faced character of the mode that spans both the confined region (where the radial electric field is negative) and the near scrape-off layer (SOL) (where it is positive). Fluctuations of density and potential at the QCM frequency are more anti-correlated towards the confined region, which is a fingerprint of electromagnetic instabilities, while they become more correlated in the SOL, as expected for a drift-wave, inducing cross-field transport across the separatrix.

The imaging heavy ion beam probe (i-HIBP) diagnostic has been successfully commissioned at ASDEX Upgrade. The i-HIBP injects a primary neutral beam into the plasma, where it is ionized, leading to a fan of secondary (charged) beams. These are deflected by the magnetic field of the tokamak and collected by a scintillator detector, generating a strike-line light pattern that encodes information on the density, electrostatic potential, and magnetic field of the plasma edge. The first measurements have been made, demonstrating the proof-of-principle of this diagnostic technique. A primary beam of 85/87Rb has been used with energies ranging between 60 and 72 keV and extracted currents up to 1.5 mA. The first signals have been obtained in experiments covering a wide range of parameter spaces, with plasma currents (Ip) between 0.2 and 0.8 MA and on-axis toroidal magnetic field (Bt) between 1.9 and 2.7 T. Low densities appear to be critical for the performance of the diagnostic, as signals are typically observed only when the line integrated density is below 2.0–3.0 × 1019 m−2 in the central interferometer chord, depending on the plasma shape. The strike line moves as expected when Ip is ramped, indicating that current measurements are possible. Additionally, clear dynamics in the intensity of the strike line are often observed, which might be linked to changes in the edge profile structure. However, the signal-to-background ratio of the signals is hampered by stray light, and the image guide degradation is due to neutron irradiation. Finally, simulations have been carried out to investigate the sensitivity of the expected signals to plasma density and temperature. The results are in qualitative agreement with the experimental observations, suggesting that the diagnostic is almost insensitive to fluctuations in the temperature profile, while the signal level is highly determined by the density profile due to the beam attenuation.

Experimental validation is presented for a new type of microwave diagnostic, first introduced in the theoretical study in M. Usoltceva et al., Rev. Sci. Instrum. 93, 013502 (2022). A new term is adopted for this technique to highlight its difference from interferometry: intensity refractometry. The diagnostic allows measuring electron density, and in this work, it is applied at the edge of a tokamak. The implementation of this technique at ASDEX Upgrade, called Microwave Intensity refractometer in the Limiter Shadow (MILS), provides the first experimental proof of the diagnostic concept. Densities predicted by MILS are compared to several other diagnostics. The agreement and discrepancy in various radial regions of the density profile are analyzed and possible reasons are discussed. A wide density coverage is shown in the example discharges with densities from 2*10^17 m^-3 to 2*10^19 m^-3 at the limiter position. In these experiments, the radial location of the measurements varied from 5 cm in front of the limiter (up to 1 cm inside the separatrix was measured) to 3 cm in the limiter shadow. Experimental challenges of MILS operation and data processing are presented.

Gas puff modulation experiments are performed at ASDEX Upgrade in L-mode, EDA H-mode and quasi-continuous exhaust discharges. Plasma density and temperatures are measured and their temporal development is analyzed simultaneously, revealing that both heat and particle transport are strongly influenced by the modulation. As a consequence, the particle transport coefficients are underdetermined. In the transport modelling, the pedestal cannot be treated as a single region, but the pedestal foot must be allowed to increase its transport with gas puff modulation independently. The analysis of the temporal behaviours of the heat and particle diffusivities shows that they are strongly correlated. Considering the heat diffusivity as a proxy for the particle diffusivity, allows interpretation of the density evolution: a pinch is not required for any of the discharges. An analysis with the gyrokinetic turbulence code GENE identifies dominant instabilities and reproduces several experimentally found trends. Despite all uncertainties concerning particle transport, one can expect a future reactor featuring a weak edge density gradient even with purely diffusive transport.

Gas puff modulation experiments are performed at ASDEX Upgrade in L-mode plasmas. We model the discharge with the ASTRA transport code in order to determine transport coefficients outside of a normalized radius of ρ pol = 0.95. The experimental data is consistent with a range of particle diffusivities and pinch velocities of the order of D = (0.20 ± 0.13) m2 s−1 and v = (−1 ± 2) m s−1, respectively. The electron temperature response caused by the gas modulation permits to estimate also that heat diffusivity χ e increases almost linearly when collisionality rises due to fuelling. The fuelling particle flux is amplified by recycling, overcompensating losses.

An overview of recent results obtained at the tokamak ASDEX Upgrade (AUG) is given. A work flow for predictive profile modelling of AUG discharges was established which is able to reproduce experimental H-mode plasma profiles based on engineering parameters only. In the plasma center, theoretical predictions on plasma current redistribution by a dynamo effect were confirmed experimentally. For core transport, the stabilizing effect of fast ion distributions on turbulent transport is shown to be important to explain the core isotope effect and improves the description of hollow low-Z impurity profiles. The L–H power threshold of hydrogen plasmas is not affected by small helium admixtures and it increases continuously from the deuterium to the hydrogen level when the hydrogen concentration is raised from 0 to 100%. One focus of recent campaigns was the search for a fusion relevant integrated plasma scenario without large edge localised modes (ELMs). Results from six different ELM-free confinement regimes are compared with respect to reactor relevance: ELM suppression by magnetic perturbation coils could be attributed to toroidally asymmetric turbulent fluctuations in the vicinity of the separatrix. Stable improved confinement mode plasma phases with a detached inner divertor were obtained using a feedback control of the plasma β. The enhanced D α H-mode regime was extended to higher heating power by feedback controlled radiative cooling with argon. The quasi-coherent exhaust regime was developed into an integrated scenario at high heating power and energy confinement, with a detached divertor and without large ELMs. Small ELMs close to the separatrix lead to peeling-ballooning stability and quasi continuous power exhaust. Helium beam density fluctuation measurements confirm that transport close to the separatrix is important to achieve the different ELM-free regimes. Based on separatrix plasma parameters and interchange-drift-Alfvén turbulence, an analytic model was derived that reproduces the experimentally found important operational boundaries of the density limit and between L- and H-mode confinement. Feedback control for the X-point radiator (XPR) position was established as an important element for divertor detachment control. Stable and detached ELM-free phases with H-mode confinement quality were obtained when the XPR was moved 10 cm above the X-point. Investigations of the plasma in the future flexible snow-flake divertor of AUG by means of first SOLPS-ITER simulations with drifts activated predict beneficial detachment properties and the activation of an additional strike point by the drifts.

The L-mode density limit in ASDEX Upgrade is investigated with focus on the scrape-off layer. We observe a chain of events that involves the collapse of the E×B outer shear layer (just outside the separatrix), widening of the electron pressure fall-off length, outer divertor detachment, density shoulder formation, reduction and flattening of the parallel Mach number, and enhanced turbulent cross-field transport. The latter effect was associated with higher amplitude density fluctuations at low frequencies, where the turbulent cross-phase between density and velocity fluctuations pointing radially outward was favourable.

The effect of the axisymmetric plasma shape on the non-axisymmetric plasma response to resonant magnetic perturbations is investigated in experiment and modeling for the DIII-D, EAST and ASDEX Upgrade tokamaks. Systematically modeling the effect of the triangularity, whilst keeping other equilibrium quantities largely unchanged, reveals that the plasma response is strongly suppressed at high triangularity compared to that at low triangularity. This is validated through targeted comparison with experiments at DIII-D, EAST, and ASDEX Upgrade. DIII-D and EAST magnetic measurements are used to validate simulations, while at ASDEX Upgrade the plasma edge displacement is measured. Both experiments and modeling find a reduced magnetic plasma response on the high-field side at high triangularity across devices. Multi-modal analysis of the simulation results extracts the mode structure and applied spectrum dependence of each mode. The amplitude of the dominant mode reveals similar trends with the edge resonance and radial displacement near the X-point, which suggests that the multi-mode plasma response provides another way to understand the edge localized mode (ELM) control physics. The plasma response is strongly reduced at high triangularity compared to that at low triangularity, which implies different ELM control effects as shaping is varied. These findings indicate that the plasma shape should be taken into consideration when designing an RMP-ELM control strategy in experiment, and that predictive plasma response calculations can be used to maximize access to RMP-ELM control in future devices by maximizing the coupling between coils and the plasma.

The impact of externally applied magnetic perturbations (MPs) on fast-ion losses has been investigated by means of the light ion beam probe (LIBP) technique in the ASDEX Upgrade tokamak. The LIBP technique allows to experimentally infer the fast-ion orbit displacement induced by MPs via first-orbit losses using scintillator based fast-ion loss detector (FILD) measurements. The fast-ion orbit displacement against different applied MP spectra has been studied. These shots were conducted in ELM mitigated H-mode plasmas. A rigid rotation of the MP coils was applied with a frequency of 1 Hz, with an n = 2 configuration and changing the differential phase between the upper and lower set of coils (ΔΦul) on a shot-to-shot basis. Beam sources Q7 (tangential) and Q8 (radial) were used to probe different fast-ion orbits with FILD1. The measured fast-ion orbit displacement ranges from 3 to 20 mm approximately, and no qualitative difference is observed between ions from beam sources Q7 and Q8. The minimum is found for a ΔΦul ∼ 50°, which is shifted with respect to the minimum of the plasma boundary displacement, found at ΔΦul ∼ 0°. A first attempt to validate the orbit following code ASCOT—including the plasma response calculated with the MARS-F code—against these experimental measurements is performed. While the dependence of the first-orbit fast-ion displacement with ΔΦul does not match the experimental measurements, these simulations do capture other features such as the order of magnitude of the orbit displacement and the importance of the toroidal spectrum of the applied perturbation.

The I-mode confinement regime can feature small edge temperature drops that can lead to an increase in the energy deposited onto the divertor targets. In this work, we show that these events are associated with a relaxation of both electron temperature and density edge profiles, with the largest drop found at the pedestal top position. The relative energy loss is about 1 %, and is thus lower than that of type-I ELMs for the same pedestal top collisionality. Stability analysis of edge profiles reveals that the operational points are far from the ideal peeling-ballooning boundary. Also, we show that these events appear close to the H-mode transition in the typical I-mode operational space in ASDEX Upgrade, and that no further enhancement of energy confinement is found when they occur. Moreover, scrape-off layer transport during these events is found to be very similar to type-I ELMs, with regard to timescales ( 800 mu s), filament propagation, toroidally asymmetric energy effluxes at the midplane and asymmetry between inner and outer divertor deposited energy. In particular, the latter reveals that more energy reaches the outer divertor target. Lastly, first measurements of the divertor peak energy fluence are reported, and projections to ARC-a reactor that could potentially operate in I-mode-are drawn.