Dirk Stieglitz

After a major hardware extension ASDEX Upgrade has now established and characterized a variety of alternative divertor configurations (ADCs) with heating powers reaching up to 20 MW and plasma currents up to 1 MA. These high performance conditions in configurations with arbitrarily small field line incidence angles were made possible due to the high precision in the tile alignment of 300 μ m as well as a careful optimization of the error fields caused by the current feeds. The formation of an X-point radiator (XPR) was observed in a low-field-side snowflake minus (LFS SF − ) configuration at a very low impurity concentration determined by the intrinsic sources only, i.e. without additional seeding of impurities. Compared to a typical lower single-null (SN) configuration the edge localized modes (ELMs) were found to be substantially smaller in size and higher in frequency and therefore barely detectable. While the primary strike line is fully detached the secondary one in the far-SOL shows heat fluxes of a similar magnitude as the primary strike line in a previous upper SN phase. These fluxes might be reduced by installing configuration-optimized baffles and/or by increasing the plasma current and thereby reducing λ q . In fact significantly smaller far-SOL heat fluxes were found when increasing the plasma current from 800 to 1000 kA. According to the divertor Langmuir probes the peak heat flux is then by a factor of two smaller than the ones in the SN reference.

Recent experiments conducted during the 2025 campaign in ASDEX Upgrade have pushed the limits of negative triangularity shaping, achieving top triangularity values of δ top ∼ − 0.55 while keeping bottom triangularity around 0 in diverted plasmas. Guided by ideal-MHD stability calculations using the linear MHD BALOO solver and supported by TCV results, these experiments have demonstrated a transition to an edge that is more stable against peeling-ballooning modes, leading to ELM-free plasmas. Those ELM-free plasmas also exhibit a dithering behavior similar to limit-cycle oscillations. Linear MHD modeling showed that the achieved shapes remained marginal with respect to the second stability region for ballooning modes suggesting that with minor further shaping, a robust NT edge could be achieved. The associated energy confinement is generally reduced, with clear signs of strong power degradation. In addition, a high fraction of fast-ion energy is observed, particularly in high-power/low density discharges. An approximately 15% improvement in normalized energy confinement is obtained when seeding nitrogen. As this effect comes from the edge, it highlights a path towards pedestal tailoring for further confinement optimization and supporting negative triangularity as a potentially viable reactor-relevant operational scenario in full-metal-wall devices.

This paper presents the new thermographic inversion code DELVER (Divertor Energy Load Versatile EstimatoR). Its task is to calculate the heat flux distribution on plasma-facing components from the evolution of surface temperature derived from infrared camera measurements. For operational campaign 2 of Wendelstein 7-X, water-cooled high-heat flux divertors made of layers of different materials have been installed into the plasma vessel. The previously employed explicit thermographic inversion code, THEODOR, was deemed insufficient to model these accurately due to the missing capability of defining multiple material layers. This motivated the development of DELVER, which improves upon explicit THEODOR in further aspects by handling 1D, 2D, and 3D modeling with flexible boundary conditions, an implicit solver, arbitrary functional temperature dependency of the orthotropic thermal properties of the materials, and a non-equidistant orthogonal calculation grid. The two codes are being compared using a simple analytical model, the finite element solver ANSYS®, and finally, experimental heat flux calculated from the calorimetry and heat fluxes calculated from infrared measurements. In all test cases, good agreement between DELVER simulations and expected results is achieved.

The purpose of the 5th International Atomic Energy Agency technical meeting on fusion data processing, validation and analysis (FDPVA) (Ghent University, Ghent, Belgium, 12–15 June 2023) was to provide a platform during which a set of topics relevant to FDPVA were discussed with the view of meeting the needs of next step fusion devices such as ITER. The validation and analysis of experimental data obtained from diagnostics used to characterize fusion plasmas are crucial for a knowledge-based understanding of the physical processes governing the dynamics of these plasmas. This paper presents the recent progress and achievements in the domain of plasma diagnostics data analysis and synthetic diagnostics reported at the meeting, including concept description of new devices; fusion databases; integrated data analysis; inverse problems; uncertainty propagation, verification and validation; probabilistic methods and machine learning. The relevant results underline trends observed in the current major fusion confinement devices.

In this work, we explore the pedestal properties of negative triangularity discharges with upper triangularity of δ u ≈ − 0.35 and with both ion ∇ B drift directions. In all cases, the discharges undergo a transition to H-mode with accompanying edge-localized modes (ELMs) that are not explained by the peeling-ballooning stability analysis alone. A variation in the ion ∇ B drift direction is observed. A lower pressure gradient, shallower radial electric field well and increased temperature fluctuations are measured in the favorable case. In addition, irregular ELMs are present. This difference is more pronounced in electron cyclotron resonance heating (ECRH) plasmas compared to plasmas with combined neutral beam injection and ECRH. A comparative analysis between two discharges, featuring similar plasma parameters but varied shaping, suggests that an interplay between other parameters, such as magnetic shear, the timing of the auxiliary heating and shaping, might play a strong role in low- to high-confinement transitions. While the low shaping discharge maintained L-mode, the high shaping discharge entered H-mode with ELMs, contrary to expectations.

The highly radiating nitrogen-seeded H-mode plasmas in unfavorable B_T has been characterized in the ASDEX Upgrade tokamak (AUG). Three levels of nitrogen puffing rates have been injected into a fully detached H-mode plasma, which is run in lower single null configuration with the ion B × ∇ B drift away from the X-point. A cold ( ∼ 1 − 2 eV) highly radiating ( ∼ 13.0 MW m − 3 ) region forms close to the X-point immediately after nitrogen seeding, as evidenced by measurements of the divertor Thomson scattering (DTS) and the two-dimensional bolometry reconstructions. In addition, the radiator moves further upwards above the X-point along the separatrix at the high-field side (HFS) with increasing nitrogen puffing rates, as evidenced by the Absolute eXtended UltraViolet (AXUV) measurements. The formation of the highly radiating regime is closely correlated to the modifications of the divertor plasma conditions. Along the line of sight of the DTS measurement, the electron temperature reduces down to a few eV, which initials near the X-point and further extends to the HFS scrape-off layer (SOL) simultaneously with the upward movement of the radiator, however, the electron temperature sustains ( ∼ 30 − 50 eV) at the low-field side (LFS) SOL with slightly decreased electron density. The highly radiating regime shows LFS/HFS divertor asymmetry, in contrast to that for H-mode plasmas in favorable B_T , suggesting that the drifts play an important role for the formation of the highly radiating X-point regime at AUG. The neutral particle flux increases significantly (factor of ∼ 2 ) in the private flux region, while it increases slightly ( < 20 % ) in the main chamber, thus suggesting an enhanced sub-divertor neutral compression with the formation of the highly radiating regime. Furthermore, a degradation of the pedestal electron density was observed with an enhancement of the electron temperature further inside the pedestal, and complete divertor detachment was achieved by nitrogen seeding with sustained plasma confinement. Finally, particle sources and flow patterns of deuterium and nitrogen ions have been analyzed by SOLPS-ITER modelling, confirming that the E × B drift plays an important role for the formation of the highly radiating regime in unfavorable B_T at AUG.

The temperature decay length ( λ T e ) in the near scrape-off layer (SOL) reflects the ratio between the transport parallel to the magnetic field lines via Spitzer–Härm electron conduction and the perpendicular mechanisms involving neoclassical and anomalous transport. The implementation of the thermal helium beam diagnostic on ASDEX Upgrade has enabled an excellent spatiotemporal study of the structure of the electron temperature profile ( T e ) around the separatrix and the derivation of the near SOL decay lengths. From the analysis of the T e profile structure of attached H- and L-mode discharges, a self-consistent correspondence between the extrema of the electron temperature curvature profile ( ∂ R 2 T e ) and the position of the separatrix ( R s e p ) is revealed. A 1.5 D model for the power balance from closed to open field lines, including the escaping heat flux from the confined region and the parallel losses to the divertor, as well as results from the plasma edge code GRILLIX, support these experimental results. The evaluation shows that the temperature decay length is not constant over the plasma edge and SOL region, so that the absolute value of λ T e strongly depends on the analyzed region. Independent of the exact location and in agreement with edge Thomson scattering evaluations, the decay lengths show the known inverse relation to the plasma current ( I p ).

In the scrape-off layer (SOL) of fusion plasmas, radial particle and energy transport is mainly carried by plasma filaments. The plasma parameters of such filaments can be measured by the thermal helium beam diagnostic (THB). By means of an extended collisional radiative model (CRM) used for the evaluation of the diagnostic data, the light intensity response of neutral helium transitions from filaments is studied, which depends on the parameters of the filament and the background. The electron density of the filament is found to be the dominant parameter impacting the line intensities. By applying a numerical reconstruction algorithm based on the CRM, electron temperatures and densities as well as the sizes of experimentally measured filaments are obtained. This method allows for the first time measurements of the temporal decay rates of filament density and temperature in the co-moving frame of the filament and its radial propagation velocity.

This letter reports on the efforts carried out at the ASDEX Upgrade tokamak to integrate I-mode plasmas with pellet fueling and to increase the I-mode Greenwald fraction f G W , two important requirements for any DEMO operational scenario. For the first time, stationary I-mode plasmas have been achieved with pellet fueling and the core Greenwald fraction has been increased up to 0.8. Larger f G W were not achieved due to technical constraints rather than to a physics-based limit. Pellet-fueled I-mode plasmas exhibit enhanced core and edge density, while core and edge temperature are reduced. Nonetheless, edge normalized gradients remain I-mode-like, namely shallow for the density and steep for the temperature. The I-mode energy confinement time is found to obey two distinct density dependencies: for f G W < 0.4 it rises with increasing f G W , while for f G W > 0.4 the energy confinement time plateaus with increasing f G W , or even decreases with f G W for the pellet-fueled plasmas. Similarities with Ohmic and L-mode energy confinement time dependency on density are discussed.