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DOE OSTI2018

catsHTM: A Tool for Fast Accessing and Cross-matching Large Astronomical Catalogs

Fast access to large catalogs is required for some astronomical applications. Here we introduce the catsHTM tool, consisting of several large catalogs reformatted into HDF5-based file format, which can be downloaded and used locally. To allow fast access, the catalogs are partitioned into hierarchical triangular meshes and stored in HDF5 files. Several tools are provided to perform efficient cone searches at resolutions spanning from a few arc-seconds to degrees, within a few milliseconds time. The first released version includes the following catalogs (by alphabetical order): 2MASS, 2MASS extended sources, AKARI, APASS, Cosmos, DECaLS/DR5, FIRST, GAIA/DR1, GAIA/DR2, GALEX/DR6Plus7, HSC/v2, IPHAS/DR2, NED redshifts, NVSS, Pan-STARRS1/DR1, PTF photometric catalog, ROSAT faint source, SDSS sources, SDSS/DR14 spectroscopy, SkyMapper, Spitzer/SAGE, Spitzer/IRAC galactic center, UCAC4, UKIDSS/DR10, VST/ATLAS/DR3, VST/KiDS/DR3, WISE and XMM. We provide Python code that allows to perform cone searches, as well as MATLAB code for performing cone searches, catalog cross-matching, general searches, as well as load and create these catalogs.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2017

Overview of the JET results in support to ITER

Here, 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.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

The effect of plasma shape and neutral beam mix on the rotation threshold for RMP-ELM suppression

Varying the neutral beam injection (NBI) mix reveals a clear pedestal-top rotation threshold for edge localized mode (ELM) suppression by resonant magnetic perturbations. Guided by expectations for the RMP penetration mechanism, the rotation threshold is found to correspond to a critical radius for the ExB rotation zero-crossing. No such critical radius is observed for the electron perpendicular rotation zero-crossing. Varying the amount and ratio of power in different NBI source geometries (termed the NBI mix) also reveals that the rotation threshold can be crossed at widely varying total injected NBI torques. Computing the local torque density at the edge, the rotation threshold is found to be crossed when the local edge NBI torque is negative in nearly all discharges. Reducing the upper triangularity (delta_up ) from the ITER-similar value of 0.3 to 0.1 significantly reduces the pedestal height and width. This in turn: 1) Increases the rotation threshold and yields a more outward critical ExB rotation zero-crossing location. 2) Decreases the density threshold, consistent with a comparable collisionality range at lower pedestal temperatures. 3) Increases the input torque requirement, due to observed lower confinement and smaller intrinsic torque. These findings represent an important step along the road to predicting ELM suppression access conditions in future tokamaks such as ITER, where the toroidal rotation is expected to be low and consequently the rotation zero-crossing far from the pedestal-top.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Central ion cyclotron emission in the DIII-D tokamak

Collective ion cyclotron emission (ICE) at the ion cyclotron frequency and its harmonics is a potential passive diagnostic of the fast-ion distribution in fusion reactors. ICE is observed in most plasmas in the DIII-D tokamak and is most strongly excited by the fast ions from neutral beam injection. The conventional outboard-edge ICE is detected in H-mode plasmas. However, weaker centrally-localized ICE is measured in L-mode plasmas, including those with negative triangularity shapes. Similar ICE spectra are found with both ICE diagnostics systems, the dedicated magnetic probes and the instrumented antenna straps. Many differences in the behavior of this central ICE are generated by varying the deuterium beam injection angle into deuterium plasmas. The co-current 'near-perpendicular' beam excites the most central ICE from the co-current beams, with this emission detected from the fundamental to the fifth ICE harmonic. However, the counter-current 'near-tangential' beam destabilizes the highest amounts of centrally-localized ICE. This emission is spectrally broader than that driven by the co-current beams and is observed up to its seventh ICE harmonic. The central ICE excited by this co-current beam correlated strongly on the local electron density and related parameters (plasma current and neutron rate) and increased with deeper fast-ion loss boundaries towards the magnetic axis. This was also the case with second harmonic ICE driven by the counter-current beam but not with its stronger third harmonic emission. The central ICE harmonics destabilized by both beams are observed to have different temporal dynamics. The central ICE amplitude responds rapidly to transient MHD events; it dropped and recovered in less than a millisecond at each sawtooth event. ICE frequency splitting is triggered by both the co-current and counter-current 'near-tangential' beams. The data presented provide opportunities to test and validate models of excitation of ICE by energetic ions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Turbulence driven widening of the near-SOL power width in ASDEX Upgrade H-Mode discharges

Divertor heat flux data from infra red (IR) from various tokamaks in H-Mode regime show that the power width, $\lambda_{q}$, scales approximately like 1.6 $(R/a) ρ_{s,pol}$. However, the IR based scaling comes with the restriction that only discharges were considered with attached divertor conditions. Here we set up a new data base covering low edge density data but also including high density plasmas reaching the H-mode density limit. We use Thomson-scattering to measure the electron temperature decay length which will set the near-SOL power width through parallel heat conduction. We present scaling results for the density, temperature and pressure separatrix fall-off lengths. As the principal result we derive a generalized power width scaling which reads as $\lambda_{q}\propto\,\rho_{s,pol}\cdot(1+2.1\alpha_{t}^{1.7})$ where $\alpha_{t}$ describes a normalized collisionality ($\alpha_{t} = 3\cdot\,10^{-18}\,R\,q^2\,n\,Z_{eff}T^{-2}$). The parameter $\alpha_{t}$ describes the relative importance of the interchange effect on drift-wave turbulence as proposed by Scott and is found to be closely linked to the diamagnetic parameter αd introduced by Rogers, Drake and Zeiler (RDZ). The edge operational space for ASDEX Upgrade displays a boundary qualitatively in agreement with the region labelled as density limit in the edge plasma phase space proposed by RDZ. The data base shows in the limit of low edge densities ($\alpha_{t}\simeq\,0.2$) accurate agreement to the IR based power width scaling. At elevated separatrix densities ($\alpha_{t}\simeq\,1$) the power width is broadened by a factor of up to about three accompanied by a reduction of the $\text{H}_{98,y2}$ confinement factor. Finally, plasmas with higher triangularity show higher values for $\text{H}_{98,y2}$ in the same range of normalized collisionality.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Analytical formula for pellet fuel source density in toroidal plasma configurations based on an areal deposition model

An analytical formula for the fuel source density η F (Ψ) over a magnetic flux surface Ψ following pellet injection is derived. During pellet motion the ablated and ionized fuel will be initially deposited nearby the pellet, prior to subsequent homogenization within the magnetic flux surfaces. Here in our research, it is assumed that the deposition area projected normal to the magnetic field is on the order of the cross-sectional area of the ionized ablation column whose axis is aligned with the magnetic field. Previous pellet deposition codes [1] employed a point source deposition approximation, which overpredicts the surface-averaged density, particularly when the obliquity between the pellet trajectory and the flux surface normal approaches 90 degrees. A key finding of this study is that as the obliquity approaches 90 degrees, grazing trajectory, the areal model predicts a well behaved η F and removes the unphysical divergence stemming from the point model. To illustrate the difference exhibited between areal and point models for near grazing trajectories, we considered vertical pellet injection into a CFETR-like tokamak with a negative triangularity plasma cross section using a pellet ablation/deposition module.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Gyrokinetic simulation of low-frequency Alfvénic modes in DIII-D tokamak

Global gyrokinetic simulations find that a beta-induced Alfvén eigenmode (BAE) and a Low frequency mode (LFM) coexist in the DIII-D tokamak experiments. The LFM can be excited without fast ions and has a frequency inside the gap of the beta-induced Alfvén-acoustic eigenmode (BAAE). The mode structure of the LFM from simulations and many of its parametric dependencies are consistent with experimental observations. For the BAE, the simulated mode structure is consistent with experiment but the frequency is lower than experiment. An antenna scan shows that the LFM is NOT the conventioanl BAAE; rather, it is an interchange-like electromagnetic mode excited by non-resonant drive of pressure gradients. The compressible magnetic perturbations significantly increase the growth rates of the BAE and LFM. On the other hand, effects of trapped electrons and equilibrium current have modest effects on the BAE and LFM. Finally, radial transfer of wave energy is different between the BAE and LFM, resulting in their respective triangular and elliptic mode shapes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Toroidal modeling of runaway avalanche in DIII-D discharges

Abstract A toroidal modeling tool is developed to study the runaway electron (RE) avalanche production process in tokamak plasmas, by coupling the Rosenbluth–Putvinski avalanche model (Rosenbluth and Putvinski 1997 Nucl. Fusion 37 1355) with an n = 0 magneto-hydrodynamic (MHD) solver. Initial value numerical simulations are carried out for two DIII-D discharges with different plasma shapes (one near circular, and the other with high elongation). It is found that, assuming the same level of about 1% seed current level, the Rosenbluth–Putvinski model somewhat underestimates the RE plateau current for the circular-shaped plasma, as compared with that measured in DIII-D experiments. For an elongated, higher current plasma, simulations find strong runaway current avalanche production despite the lack of measured plateau RE current in experiments. A possible reason for this discrepancy is a lack of additional RE dissipation physics in the present two-dimensional model. Systematic scans of the plasma boundary shape, at fixed pre-disruption plasma current, find that the plasma elongation helps to reduce the RE avalanche production, confirming recent results obtained with an analytic model (Fülöp et al 2020 J. Plasma Phys. 86 474860101). The effect of the plasma triangularity (either positive or negative), on the other hand, has a minor effect. On the physics side, the avalanche process involves two competing mechanisms associated with the electric field. On the one hand, a stronger electric field produces a higher instantaneous avalanche growth rate. On the other hand, a fast growing RE current quickly reduces the fraction of the conduction current together with the electric field, and hence a faster dissipation of the poloidal flux. As a final result of these two dynamic processes, the runaway plateau current is not always the largest with the strongest initial electric field. These results lay the foundation for future self-consistent inclusion of the MHD dynamics and the RE amplification process.

Physics↗

DOE OSTI2021

Exploration of RMP ELM control on ITER-similar shape (ISS) in KSTAR

Recent control advances at KSTAR enabled us to not only establish but also to sustain the ITER-similar shape (ISS) in a stationary manner, matching the ITER-like dimensionless shape parameters in lower single null configuration with vastly contrasting upper/lower triangularities. Here, the kinetic parameters of the ISS are different from the typical KSTAR configurations seen in ELM-RMP suppression experiments. We experimentally demonstrate the ISS-compatible resonant magnetic perturbation (RMP) control in KSTAR for the first time, using the n = 2, +90 degrees phasing RMP.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

DIII-D research advancing the physics basis for optimizing the tokamak approach to fusion energy

DIII-D physics research addresses critical challenges for the operation of ITER and the next generation of fusion energy devices. This is done through a focus on innovations to provide solutions for high performance long pulse operation, coupled with fundamental plasma physics understanding and model validation, to drive scenario development by integrating high performance core and boundary plasmas. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power, and in pressure broadening for Alfven eigenmode control from a co-/counter-I p steerable off-axis neutral beam, all improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. Fundamental studies into the modes that drive the evolution of the pedestal pressure profile and electron vs ion heat flux validate predictive models of pedestal recovery after ELMs. Understanding the physics mechanisms of ELM control and density pumpout by 3D magnetic perturbation fields leads to confident predictions for ITER and future devices. Validated modeling of high-Z shattered pellet injection for disruption mitigation, runaway electron dissipation, and techniques for disruption prediction and avoidance including machine learning, give confidence in handling disruptivity for future devices. For the non-nuclear phase of ITER, two actuators are identified to lower the L–H threshold power in hydrogen plasmas. With this physics understanding and suite of capabilities, a high poloidal beta optimized-core scenario with an internal transport barrier that projects nearly to Q = 10 in ITER at ~8 MA was coupled to a detached divertor, and a near super H-mode optimized-pedestal scenario with co-I p beam injection was coupled to a radiative divertor. The hybrid core scenario was achieved directly, without the need for anomalous current diffusion, using off-axis current drive actuators. Also, a controller to assess proximity to stability limits and regulate β N in the ITER baseline scenario, based on plasma response to probing 3D fields, was demonstrated. Finally, innovative tokamak operation using a negative triangularity shape showed many attractive features for future pilot plant operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

The dependence of confinement on the isotope mass in the core and the edge of AUG and JET-ILW H-mode plasmas

Experiments in ASDEX Upgrade (AUG) and JET with the ITER-like wall (JET-ILW) are performed to separate the pedestal and core contributions to confinement in H-modes with different main ion masses. A strong isotope mass dependence in the pedestal is found which is enhanced at high gas puffing. This is because the ELM type changes when going from D to H for matched engineering parameters, which is likely due to differences in the inter ELM transport with isotope mass. The pedestal can be matched in H and D plasmas by varying only the triangularity and keeping the engineering parameters relevant for core transport the same. With matched pedestals Astra/TGLF (Sat1geo) core transport simulations predict the experimental profiles equally well for H and D. These core transport simulations show a negligible mass dependence and no gyro-Bohm scaling is observed. However, to match the experimental observations at medium β it is required to take the fast-ion dilution and rotation into account. This is not enough for high β plasmas where for the first time a profile match between H and D plasmas was achieved experimentally. Under these conditions quasilinear modelling with TGLF over predicts the transport in the core of H and D plasmas alike.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Experimental study on the role of the target electron temperature as a key parameter linking recycling to plasma performance in JET-ILW*

Changes in global and edge plasma parameters (H98(y,2), dimensionless collisionality ν*, core density peaking, separatrix density ne,sep) with variations in the D2 fueling rate and divertor configuration are unified into a single trend when mapped to $\langle$Te,ot$\rangle$, the spatially averaged spectroscopically derived outer target electron temperature. Dedicated JET with the ITER-like wall (JET-ILW) experiments in combination with an extended JET-ILW database of unseeded low-triangularity H-mode plasmas spanning a wide range of D2 fueling rates, Ip, Bt and heating power have demonstrated the importance of $\langle$Te,ot$\rangle$ as a key physics parameter linking the recycling particle source and detachment with plasma performance. The remarkably robust H98(y,2) trend with $\langle$Te,ot$\rangle$ is connected to a strong inverse correlation between $\langle$Te,ot$\rangle$, ne,sep and ν*, thus directly linking changes in the divertor recycling moderated by $\langle$Te,ot$\rangle$ with the previously established relationship between ν*, core density peaking and core pressure resulting in a degradation in core plasma performance with decreasing $\langle$Te,ot$\rangle$ (increasing ν*). Additionally, a strong inverse correlation between the separatrix to pedestal density ratio, ne,sep/ne,ped, and $\langle$Te,ot$\rangle$ is also established, with the rise in ne,sep/ne,ped saturating at $\langle$Te,ot$\rangle$ > 10 eV. A strong reduction in H98(y,2) is observed as $\langle$Te,ot$\rangle$ is driven from 30 to 10 eV via additional D2 gas fueling, while the divertor remains attached. Consequently, the pronounced performance degradation in attached divertor conditions has implications for impurity seeding radiative divertor scenarios, in which H98(y,2) is already low (~0.7) before impurities are injected into the plasma since moderate gas fueling rates are required to promote high divertor neutral pressure. A favorable pedestal pressure, pe,ped, dependence on Ip has also been observed, with an overall increase in pe,ped at Ip = 3.4 MA as $\langle$Te,ot$\rangle$ is driven down from attached to high-recycling divertor conditions. In contrast, pe,ped is reduced with decreasing $\langle$Te,ot$\rangle$ in the lower Ip branches. Further work is needed to (i) clarify the potential role of edge opacity on the observed favorable pedestal pressure Ip scaling; as well as to (ii) project the global and edge plasma performance trends with $\langle$Te,ot$\rangle$ to reactor-scale devices to improve predictive capability of the coupling between recycling and confined plasma fueling in what are foreseen to be more opaque edge plasma conditions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Optimizing the HSX stellarator for microinstability by coil-current adjustments

Abstract The optimization of helically symmetric experiment (HSX) for reduced microinstability has been achieved by examining a large set of configurations within a neighborhood of the standard operating configuration. This entailed generating a database of more than 10 6 magnetic-field configurations for HSX by varying the currents in external coils. Using a set of volume-averaged metrics and gyrokinetic simulations, this database has helped to identify a set of configurations that can be used to regulate trapped-electron-mode stability in HSX. This set of configurations is also found to correlate flux-surface elongation and triangularity with an increase in magnetic-well depth, an increase in rotational transform, and low neoclassical heat-flux relative to the standard quasi-helically-symmetric configuration. These results demonstrate sensitivity of plasma behavior in response to changes in a 3D magnetic field to both neoclassical and gyrokinetic models, and the experimental potential in HSX to explore turbulence optimization. This perturbative optimization approach is not unique to HSX, and can readily be deployed on existing fusion devices to identify novel magnetic-fields to be used in turbulence-optimization experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Stability and transport of gyrokinetic critical pedestals

Abstract A gyrokinetic threshold model for pedestal width–height scaling prediction is applied to multiple devices. A shaping and aspect ratio scan is performed on National Spherical Torus Experiment (NSTX) equilibria, finding Δ ped = 0.92 A 1.04 κ − 1.24 0.38 δ β θ , ped 1.05 for the wide-pedestal branch with pedestal width Δ ped , aspect ratio A , elongation κ , triangularity δ , and normalized pedestal height β θ , ped . The width–transport scaling is found to vary significantly if the pedestal height is varied either with a fixed density or fixed temperature, showing how fueling and heating sources affect the pedestal density and temperature profiles for the kinetic-ballooning-mode (KBM) limited profiles. For an NSTX equilibrium, at fixed density, the wide branch is Δ ped = 0.028 ( q e / Γ e − 1.7 ) 1.5 ∼ η e 1.5 and at fixed temperature Δ ped = 0.31 ( q e / Γ e − 4.7 ) 0.85 ∼ η e 0.85 , where q e and Γ e are turbulent electron heat and particle fluxes and η e = ∇ ln ⁡ T e / ∇ ln ⁡ n e for an electron temperature T e and density n e . Pedestals close to the KBM limit are shown to have modified turbulent transport coefficients compared to the strongly driven KBMs. The role of flow shear is studied as a width–height scaling constraint and pedestal saturation mechanism for a standard and lithiated wide pedestal discharge. Finally, the stability, transport, and flow shear constraints are combined and examined for an NSTX experiment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Predicted thresholds for RMP ELM suppression access in double-null configurations

Modeling of DIII-D plasmas spanning shapes from single to double-null (DN) reveals new insights into the nature of resonant magnetic perturbation (RMP) conditions necessary for edge-localized mode (ELM) suppression. The suppression of ELMs with RMPs has proven difficult in DN configurations, where no device has thus far reported any hints of suppression. Modeling using the GPEC code finds a reduced high-field side response closer to DN shaping. The resulting synthetic diagnostic measurements are consistent with what has been observed in experiments on DIII-D, validating the plasma response model in this regime. While common metrics for suppression do not illustrate a clear distinction between single-null (SN) and DN cases, the pedestal top resonant field does show a ∼20% decrease at DN shaping in modeling. Field penetration is assessed using linear tearing theory, which demonstrates a lack of sufficient pedestal top resonant flux in the DN shape, requiring at least 1.5 × greater RMP coil currents than what was used in experiment. Analysis from drift kinetic simulations further indicate up to 2 × larger critical island widths are required at the pedestal top for tearing mode growth compared to SN cases. Effective island widths inferred from 3D ideal MHD are also analyzed, where maximum widths in lab coordinates indicate a threshold of ∼18–24× the ion gyroradius for sufficient profile flattening for ELM suppression. These results suggest that ELM suppression may be possible in DN with sufficiently large RMP coil amplitude. Future prospects of achieving RMP ELM suppression in the DN configuration may involve going to lower triangularity, which is also highlighted in this work.

RMP ELM suppression↗

DOE OSTI2026

Overview of the ASDEX Upgrade results

After a 26-month vent ASDEX Upgrade (AUG) went back in operation with a newly designed upper W-divertor suitable for alternative divertor configurations (featuring in-vessel coils and cryo-pump). Parameter scans and an extensive set of measurements were obtained and their interpretation is ongoing. Prompted by the ITER wall change, dedicated experiments on non-boronized plasma startup were contrasted to that employing asymmetric and more symmetric boronizations. The asymmetric boronization proved to be as beneficial as the more symmetric one, which is in contrast to previous model calculations assuming perfect sticking of boron (measurements suggest sticking ≈ 0.3). In the startup phase also the impurity influxes at the outboard limiters were investigated contrasting the unboronized case featuring cold edges (low- Z radiation) to the boronized case, in which the lifetime of the boron layers could be estimated. Pedestal stability investigations revealed that the quasi-continuous exhaust (QCE) regime is obtained when ballooning modes are active in the vicinity of the separatrix and the global peeling-ballooning stability is high enough. Thus, at high enough shaping and high gas flux both can be achieved and QCE is a consequence. The closely related enhanced D-alpha (EDA) mode is not clearly distinguishable from QCE, e.g. the quasi-coherent mode characteristic for EDA also shows up in QCE. In QCE the impurity transport is behaving benign as could be measured for Ne with a novel analysis method making use of a comprehensive set of CXRS measurements. For high radiative fractions the regime of the X-point radiator (XPR) is accessible at AUG and the understanding of its access conditions and behaviour is further developed. Due to the localized radiative cooling at the X-point the XPR can be well diagnosed and thus controlled. For negative triangularity shapes, further experiments at increased shaping resulted in strongly heated L-mode plasmas avoiding ELMs. Two integrated modelling approaches towards ITER suggested that core W-accumulation will be no issue for ITER and that the fusion yield in ITER may be Q = 12 (i.e. ITPA20-IL scaling is too pessimistic). Further, investigations of the ITER ramp-down in AUG provide insights into maintaining position control. Various aspects of shattered pellet injection were investigated in AUG and one of the results show that with increasing Ne fraction the radiation during the current quench increases and the current decay becomes faster.

ASDEX Upgrade↗

DOE OSTI2026

Overview of the MAST Upgrade physics programme: testing novel concepts at low aspect ratio to inform future devices

The research programme performed on the Mega Amp Spherical Tokamak (MAST) Upgrade experiment has made significant advances in developing the physics understanding of low aspect ratio tokamaks in support of the operation of ITER and design of fusion powerplants. High performance plasma scenarios have been developed to facilitate a broad programme of experiments, in which confinement is constrained by the presence of m/n = 2/1 modes that cause substantial losses of fast ions. The onset of these modes coincides with the q = 2 surface residing in a local minimum in the toroidal current density profile. The maximum electron temperature at the pedestal top, T e,ped is limited with gas fuelling to ∼350 eV to maintain regular ELMs; higher T e,ped results in a transition to a non-stationary ELM-free regime. The operational space of spherical tokamaks has been expanded into small and ELM-free regimes. Strong shaping of the last closed flux surface can induce a transition from large to small ELMs, and ELM suppression with resonant magnetic perturbations has been observed for the first time in a low aspect ratio tokamak. Negative triangularity shaping has induced a transition from ELMy H-mode to a high-performance L-mode regime for the first time in a low aspect ratio tokamak. In studies of fast ion confinement, losses of fast particles due to Global Alfvén Eigenmodes have been identified. Interactions between fast ions generated by off-axis neutral beam injection and thermal neutrals can result in significant losses of fast ions. Experiments with on- and off-axis neutral beam injection exhibit a flux pumping mechanism, where the central safety factor is held to ∼1 in the absence of sawteeth. In studies of pedestal physics, it has been found that elevated main chamber neutral pressures result in an increase in the electron density and reduction in the temperature at the pedestal top. Advances in understanding plasma exhaust include the integration of a high-performance plasma core with detached outer divertors in the X-point target configuration. A newly commissioned lower divertor cryopump reduces the lower divertor neutral pressure by up to 50%, with minimal effect on the main chamber or upper divertor. New measurements and SOLPS-ITER simulations emphasise the importance of plasma–neutral interactions on divertor detachment in the conditions accessible in experiments. Real-time control of the ionisation front location in both divertor chambers independently has been demonstrated in double null experiments, enabled by the tightly baffled divertor chambers.

MAST Upgrade↗

DOE OSTI2026

Progress and innovations in the TCV tokamak research programme

Research on the Tokamak à Configuration Variable addresses a wide range of key questions relevant to ITER and future fusion power plants. Over the past two years, highly productive experimental campaigns have led to major advances across several areas: the ITER baseline scenario; pedestal properties in low-collisionality, peeling-limited conditions; and the development of high- β N , non-inductive regimes. Alternative high-confinement scenarios have likewise received significant attention, with remarkable progress in quasi-continuous exhaust operation, X-point radiator plasmas, and negative triangularity configurations. Substantial achievements were also made in the mitigation or benign termination of runaway electron beams, in elucidating fast-ion loss mechanisms, and in improving exhaust behaviour in both conventional and alternative divertor geometries. These experimental results have been strongly supported by advances in modelling and their direct application to the experiment, ranging from gyrokinetic simulations of core and pedestal turbulence to fluid-based studies of scrape-off layer and divertor physics in diverse geometries. Plasma control has taken on an increasingly important role, with model-based and data-driven approaches now closely intertwined with physics studies. This article provides a overview of these recent activities, together with a brief outlook on forthcoming upgrades and next steps.

EPFL↗