Variational autoencoder (VAE)-based representation learning algorithms are explored for their capability to disentangle tokamak size dependence from other dependencies in a dataset of thousands of observed pedestal electron density and temperature profiles from JET and ASDEX Upgrade tokamaks. Representation learning aims to establish a useful representation that characterizes the dataset. In the context of magnetic confinement fusion devices, a useful representation could be considered to map the high-dimensional observations to a manifold that represents the actual degrees of freedom of the plasma scenario. A desired property for these representations is organization of the information into disentangled variables, enabling interpretation of the latent variables as representations of semantically meaningful characteristics of the data. The representation learning algorithms in this work are based on VAE that encodes the pedestal profile information into a reduced dimensionality latent space and learns to reconstruct the full profile information given the latent representation. Attaching an auxiliary regression objective for the machine control parameter configuration, broadly following the architecture of the domain invariant variational autoencoder (DIVA), the model learns to associate device control parameters with the latent representation. With this multimachine dataset, the representation does encode density scaling with device size that is qualitatively consistent with Greenwald density limit scaling. However, if the major radius of the device is given through a common regression objective with the other machine control parameters, the latent state of the representation struggles to clearly disentangle the device size from changes of the other machine control parameters. When separating the device size as an independent latent variable with dedicated regression objectives, similar to separation of domain and class labels in the original DIVA publication, the latent space becomes well organized as a function of the device size.

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).

A review is given on the highlights of a scatter-shot approach of developing machine-learning methods and artificial neural networks based fast predictors for the application to fusion exhaust. The aim is to enable and facilitate optimized and improved modeling allowing more flexible integration of physics models in the light of extrapolations towards future fusion devices. The project encompasses various research objectives: (a) developments of surrogate model predictors for power & particle exhaust in fusion power plants; (b) assessments of surrogate models for time-dependent phenomena in the plasma-edge; (c) feasibility studies of micro-macro model discovery for plasma-facing components surface morphology & durability; and (d) enhancements of pedestal models & databases through interpolators and generators exploiting uncertainty quantification. Presented results demonstrate useful applications for machine-learning and artificial intelligence in fusion exhaust modeling schemes, enabling an unprecedented combination of both fast and accurate simulation.

The path to fusion in the United States requires partnership between public and private sector. While the private sector provides the vigor to take some of the major steps necessary, there is a depth of expertise and capability in the public sector that is vital to resolving feasible approaches. As an open national user facility, DIII-D provides a crucial testbed to develop the required new technologies and approaches in relevant conditions. It has unparalleled potential to meet this challenge, thanks to its extreme flexibility and world leading diagnostics. This provides a basis to rapidly develop solutions that project to future reactors with confidence. The program has thus been redeveloped to enable public and private sector engagement and testing of new concepts. A new technology program has been launched to resolve plasma interacting technologies. With modest heating upgrades, the facility can confront the crucial "Integrated Tokamak Exhaust and Performance" gap, to resolve core, exhaust and technology solutions together. The device is also being redeveloped as a training facility, with dedicated student run time, a mentorship program, and open access to all opportunity roles, part of wider efforts to diversify and open pathways through inclusion, access, and equity. This exciting agenda is enabling scientists and technology researchers to pioneer the solutions needed for a Fusion Pilot Plant (FPP) and ITER this decade. As a national user facility, DIII D has singular potential to provide the tools, teams, and insight necessary, to do its part in moving the United States rapidly toward the commercialization of 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.

A machine learning approach has been implemented to measure the electron temperature directly from the emission spectra of a tokamak plasma. This approach utilized a neural network (NN) trained on a dataset of 1865 time slices from operation of the DIII-D tokamak using extreme ultraviolet/vacuum ultraviolet emission spectroscopy matched with high-accuracy divertor Thomson scattering measurements of the electron temperature, Te. This NN is shown to be particularly good at predicting Te at low temperatures (Te < 10 eV) where the NN demonstrated a mean average error of less than 1 eV. Trained to detect plasma detachment in the tokamak divertor, a NN classifier was able to correctly identify detached states (Te < 5 eV) with a 99% accuracy (an F1 score of 0.96) at an acquisition rate 10× faster than the Thomson scattering measurement. The performance of the model is understood by examining a set of 4800 theoretical spectra generated using collisional radiative modeling that was also used to predict the performance of a low-cost spectrometer viewing nitrogen emission in the visible wavelengths. These results provide a proof-of-principle that low-cost spectrometers leveraged with machine learning can be used to boost the performance of more expensive diagnostics on fusion devices and be used independently as a fast and accurate Te measurement and detachment classifier.

A model of ion temperature anisotropy for 2D plasma transport in the scrape-off layer (SOL) of tokamaks is described and implemented in the UEDGE fluid transport code. Two ion energy equations are used to describe the evolution of the separate parallel and perpendicular ion temperatures. The temperature anisotropy generates viscous forces in both parallel and perpendicular directions that modify the parallel force balance equation and add an additional cross-magnetic-field drift velocity. Using the full set of UEDGE plasma and neutral equations (particle continuity, momentum, and energy), simulations are performed for both a 1D poloidal case and a 2D (radial and poloidal) single-null tokamak geometry case to highlight the 2D effects. The results show that ion parallel flows near the magnetic X-point in a comparatively low collisionality regime can be overestimated by the standard isotropic Braginskii model. The 2D ion temperature anisotropy varies substantially near the X-point and also near the divertor target plates, due to ionization sources. Moving radially outwards at the outer midplane, the anisotropy decreases between the core boundary and the magnetic separatrix and then it increases while moving across the SOL to the chamber wall.

The Super H-mode, a regime with high pedestal pressure and stored energy, is explored on DIII-D and combined with an ion transport barrier in the plasma core to increase performance. A significant improvement of ion temperatures and confinement is facilitated by favorable conditions such as high rotational shear and high ion pedestal temperatures. As a result of a rise in density and simultaneous decrease in rotation, the ion transport barrier disappears during the discharge evolution, leading to a transition from a very high confinement state at early times, to a reduced but still high confinement phase. Additionally, in many discharges, a global magnetohydrodynamic (MHD) event consistent with the coupling of a destabilized internal mode to an edge localized mode causes a large energy loss and leads to a reorganization of the plasma into a lower temperature, higher density state. Depending on the magnitude of the global MHD event, the plasma edge collisionality can increase significantly and shift the operational boundary from the peeling to the ballooning side, which can be understood as a drop out of the Super H-mode channel into standard H-mode. Hence, in Super H-mode discharges with ion transport barriers, both the improved pedestal height and rotational shear contribute to the high stored energy. At very low levels of rotation, the confinement factor for SH modes is still expected to exceed standard H-mode by 20%–30%. With their overall stationarity and high-performance levels, Super H-mode discharges provide an attractive regime for ITER and may enable a more compact design of future fusion power plants.

KInetic code for Plasma Periphery (KIPP) was used to assess the importance of kinetic effects of parallel ion transport in the scrape-off layer (SOL) and divertor of JET high radiative H-mode inter-edge localised mode plasma conditions with strong nitrogen (N2) injection, leading to partial detachment at divertor targets. Plasma parameter profiles along the magnetic field from one of the EDGE2D-EIRENE simulation cases were used as an input for KIPP runs. The profiles were maintained by particle and power sources. This work is a continuation of the previous study carried out for electrons (Chankin et al 2018 Plasma Phys. Control. Fusion 60 115011). In this modelling KIPP calculated ion distribution functions and ion parallel power fluxes. In the main SOL kinetic effects lead to a reduction of heat (conductive power) fluxes compared to Braginskii fluxes by factors 3−4 (‘heat flux limiting’). In the divertor, on the contrary, a strong ‘heat flux enhancement’, by up to two orders of magnitude above Braginskii’s, was found. Similar to cases for electrons, high ion heat flux enhancement factors, in particular near targets, are attributed to a non-local transport of super-thermal ions originating from positions along field lines with the highest ion temperature, resulting in the appearance of bump-on-tail features on ion heat flux density profiles. Despite ion heat flux enhancement factors at the target being much higher than for electrons, total power fluxes, ion plus electron, were dominated by ion and electron convection and electron conduction, with ion conductive fluxes playing a secondary role. This must be attributed to lower ion (than electron) velocities (factor ∼ me/mi reduction), which are not compensated by kinetic effects of lower ion upstream collisionality.

A toroidally symmetric tungsten ring inserted in the lower outer divertor of DIII-D was exposed to 25 repeated, attached L-mode shots in reverse- configuration. Radial profiles of the W gross erosion flux inferred in situ from spectroscopic measurements of the WI line (400.9?nm) during these experiments are well reproduced by ERO-D3D simulations of carbon and tungsten impurity erosion, transport and redeposition in the outer divertor region. Tungsten gross erosion is mainly induced by physical sputtering of tungsten by carbon impurities. The outward radial transport of carbon impurities in the outer divertor is shown to be mainly governed by drifts in the sheath region. In addition, the erosion and redeposition of carbon on tungsten, induced by the implantation of carbon into tungsten modeled with the homogeneous mixed material model, increases the effective flux of carbon impurities onto the tungsten ring (carbon recycling on tungsten). The dynamics of carbon implantation in tungsten is shown to be consistent with the plasma shot duration in DIII-D. Moreover, it is shown that the localized deposition of tungsten measured experimentally in the outboard region away from the tungsten ring is caused by the long-range radial transport of tungsten impurities in the outer divertor region induced by the interplay between poloidal and radial drifts. Such experimental measurements might provide direct quantitative estimations of tungsten net erosion. The modeling and analysis of carbon and tungsten erosion and redeposition presented in this paper demonstrates that various physical mechanisms and their synergistic effects need to be taken into account to accurately describe erosion, transport and redeposition of impurities in tokamak divertors.

The structure of the edge plasma in a magnetic confinement system has a strong impact on the overall plasma performance. We uncover for the first time a magnetic-field-direction dependent density shelf, i.e., local flattening of the density radial profile near the magnetic separatrix, in high confinement plasmas with low edge collisionality in the DIII-D tokamak. The density shelf is correlated with a doubly peaked density profile near the divertor target plate, which tends to occur for operation with the ion B×∇B drift direction away from the X-point, as currently employed for DIII-D advanced tokamak scenarios. This double-peaked divertor plasma profile is connected via the E×B drifts, arising from a strong radial electric field induced by the radial electron temperature gradient near the divertor target. The drifts lead to the reversal of the poloidal flow above the divertor target, resulting in the formation of the density shelf. The edge density shelf can be further enhanced at higher heating power, preventing large, periodic bursts of the plasma, i.e., edge-localized modes, in the edge region, consistent with ideal magnetohydrodynamics calculations.

The radial width of the exhaust heat flux flowing in the SOL of DIII-D is found to expand at high input power and plasma density, consistent with MHD ballooning stability limits. At low heating power, ~3 MW, the SOL width remains constant and consistent with established empirical scaling dependent only on the midplane poloidal field. At high heating power, ~ 13 MW a higher separatrix density, and resulting higher separatrix pressure is required for divertor detachment. The separatrix pressure gradient at the separatrix continues to increase with density until it begins to saturate at levels ~50% above the calculated ideal MHD ballooning limit. Examination of the separate contributions to the pressure gradient from electrons and ions reveals the ion pressure gradient to saturate more strongly than the electron pressure gradient. Potential analysis issues leading to the measured pressure gradient exceeding the ballooning limit are discussed. At high density, particularly for detached divertor plasmas, the SOL width for temperature and density expand modestly, ~30–50%. The divertor plasma density profile in detachment also reflects this trend, expanding in the radial direction a factor of 2–3. Despite the SOL width expansion at the highest power and density no degradation of the pedestal and resulting core confinement is observed with the additional density at high power. These results imply a more favorable scaling for divertor heat flux control in future reactor-scale tokamaks than predicted by existing empirical scaling.

The average tungsten concentrations in the pedestal region (cW) predicted by the Monte Carlo code DIVIMP and the coupled multi-fluid plasma/kinetic neutral code EDGE2D-EIRENE are found to agree within a factor of 2 for a range of JET-ILW L-mode and H-mode plasma conditions. Under attached divertor conditions with cW exceeding 10−6, the cW predicted by DIVIMP is consistently ~50% higher than by EDGE2D-EIRENE. In colder plasma scenarios with cW <10-6, stochastic variations exceed the systematic disagreement between the two codes. The average tungsten charge predicted by EDGE2D-EIRENE in the upstream scrape-off layer is lower by 40–50% due to the bundling of the 74 tungsten ion charge states into 6 fluid species, which explains the reduced tungsten accumulation in the main plasma compared to the DIVIMP predictions.

Tungsten sputtering rates and density profiles predicted using the edge plasma codes EDGE2D-EIRENE and DIVIMP are found to agree within a factor of 4 with measurements of neutral and singly-ionized W spectral line emission in the JET low-field side (LFS) divertor, and within a factor of 2 with SXR, VUV, and bolometric calculations of the W density in the main plasma. The edge plasma W predictions are extended to the core plasma using JINTRAC integrated core-edge modelling. Prompt redeposition of W is identified as the primary reason for the discrepancy between predicted and measured W emission in the divertor. The studied plasmas include attached divertor conditions in L-mode and type-I ELMy H-mode plasmas typical for JET.To more accurately reproduce the spectroscopically inferred W sputtering rates in EDGE2D-EIRENE, imposing the experimentally observed Be concentration of order 0.5% in the divertor is necessary. However, the W density in the main plasma is predicted to be insensitive to whether or not W is sputtered by Be at the divertor targets. Instead, the majority of the predicted core W originated in L-mode from sputtering due to fast D charge-exchange atoms at the W-coated tiles above the LFS divertor, and in H-mode due to D and W ions at the targets during ELMs.

Simulations for DIII-D high confinement mode plasmas with the multifluid code UEDGE show a strong role of poloidal E x B drifts on divertor heat transport, challenging the paradigm of conduction-limited scrape-off layer (SOL) transport. While simulations with reduced drift magnitude are well aligned with the assumption that electron heat conduction dominates the SOL heat transport, simulations with drifts predict that the poloidal convective E x B heat transport dominates over electron heat conduction in both attached and detached conditions. As poloidal E x B flow propagates across magnetic field lines, poloidal transport with shallow magnetic pitch angles can reach values that are of the same order as would be provided by sonic flows parallel to the field lines. These flows can lead to strong convection-dominated divertor heat transport, increasing the poloidal volume of radiative power front, consistent with previous measurements at DIII-D. Due to these convective flows, the Lengyel integral approach, assuming zero convective fraction, is expected to provide a pessimistic estimate for the radiative capability of impurities in the divertor. For the DIII-D simulations shown here, the Lengyel integral approach underestimates the radiated power by a factor of 6, indicating that, for reliable DIII-D divertor power exhaust predictions, full two-dimensional (2D) calculations, including drifts, would be necessary.

UEDGE simulations highlight the role of cross-field drifts on the onset of detached conditions, and new calibrated divertor vacuum ultra violet (VUV) spectroscopy is used to challenge the predictions of radiative constituents in these simulations. UEDGE simulations for DIII-D H-mode plasmas with the open divertor with the ion ∇B-drift towards the X-point show a bifurcated onset of the low field side (LFS) divertor detachment, consistent with experimentally observed step-like detachment onset (Jaervinen A.E. et al 2018 Phys. Rev. Lett. 121 075001). The divertor plasma in the simulations exhibits hysteresis in upstream separatrix density between attached and detached solution branches. Reducing the drift magnitude by a factor of 3 eliminates the step-like detachment onset in the simulations, confirming the strong role of drifts in the bifurcated detachment onset. When measured local plasma densities and temperatures are within proximity of predicted values in the simulations, there is no shortfall of the local emission of the dominant resonant radiating lines. However, the simulations systematically predict a factor of two lower total integrated radiated power than measured by the bolometer with the difference lost through radial heat flow out of the computational domain. Even though there is no shortfall in the emission of the dominant lines, a shortfall of total radiated power can be caused by underpredicted spatial extent of the radiation front, indicating a potential upstream or divertor transport physics origin for the radiation shortfall, or shortfall of radiated power in the spectrum between the dominant lines. In addition to the underpredicted spatial extent, in detached conditions, the simulations overpredict the peak radiation and dominant carbon lines near the X-point, which can be alleviated by manually increasing divertor diffusivity in the simulations, highlighting the ad hoc cross-field transport as one of the key limitations of the predictive capability of these divertor fluid codes.

Access to Super H-mode is demonstrated for moderately shaped plasmas in agreement with EPED [Snyder et al., Phys. Plasmas 16, 056118 (2009)] predictions. In particular, Super H-mode is realized in a DIII-D shape that is accessible to the JET tokamak. The reduced triangularity of the JET-compatible shape compared to previous Super H-mode plasma shapes does not prevent deep ascension into the so-called Super H-mode “channel.” Operationally, access is enabled and optimized by delaying the neutral beam power injection and, thus, protracting the L–H transition. In highly shaped DIII-D plasmas, the injection of nitrogen sufficient for the establishment of a radiative divertor is shown to be possible during Super H-mode without pedestal degradation. Due to its increased stored energy and radiative divertor integration capabilities, Super H-mode is a promising candidate as operating regime for JET, ITER, and future fusion reactors.

Radiative divertor experiments and 2D fluid simulations show strong impact of cross-field drifts on the low field side (LFS) divertor target heat flux and volumetric radiation profiles when evolving to detached conditions in DIII-D high confinement mode (H-mode) plasmas with forward and reversed toroidal field configurations. In both field configurations, the peak heat flux is reduced by about a factor of 2 in detachment by D-2-injection and by factor of 3 - 4 in detachment by N-2-injection. Operating with the B x del B-drift towards the X-point (fwd. B-T), the LFS divertor radiation front is observed to shift step-like from the target to near to the X-point at the onset of detachment. In contrast, operating with the B x del B-drift away from the X-point (rev. B-T), the radiation front is observed to remain closer to the target plate and to be radially shifted towards the far SOL. These phenomena occur with detachment induced both with N-2- and D-2-injection. The step-like detachment onset is consistent with recently published theory of the role of poloidal E x B-drift in the private flux region in driving highly non-linear detachment onset in fwd. B-T [22].

The contributions of deuterium molecular emission to the total deuterium radiation was assessed in DIII-D ohmically-confined plasmas in high-recycling divertor conditions. Radial profiles of the deuterium Ly-α line intensity across the low-field side divertor leg were obtained with the recently installed divertor Survey Poor Resolution, Extended Spectrometer [1]. A high-resolution spectrometer was used to measure the poloidal profiles of the deuterium Balmer-α and the deuterium Fulcher-α band intensity in the visible wavelength range. The scrape-off layer plasma and neutral distributions were simulated using the edge fluid EDGE2D-EIRENE [2], and the numerical solutions constrained utilizing Thomson scattering and Langmuir probe measurements at the low- field side midplane and the divertor target plate. The studies show that for these conditions molecular emission plays a negligible role in the total radiative power balance of the low-field side divertor, but molecular processes are important when evaluating deuterium Balmer-α line intensity for code-experiment validation.