Arne Kallenbach

The electron density at the separatrix ( n e , sep ) plays a central role in balancing energy confinement, detachment achievement, and edge-localized mode suppression in tokamaks, thereby influencing core-edge integration. To study what determines this key parameter, a database of H-mode separatrix density measurements from the Alcator C-Mod, ASDEX Upgrade, and JET tokamaks has been assembled using a consistent analysis method across all devices. This dataset is used to assess the validity of a physics-based predictive model and to derive a regression scaling expression for n e , sep , both requiring only engineering parameters as input. The theory-based expression is obtained by coupling two-point model equations with simple geometrical relations, and successfully reproduces experimental measurements across all three devices, with the exception of a common multiplicative constant. The regression confirms similar parameter dependencies, revealing a positive dependence on divertor neutral pressure and the ratio of the power entering the scrape-off layer to the major radius, a negative dependence on the toroidal magnetic field and minor radius, and no significant dependence on the plasma current. Both the resulting scaling and theory-based expressions predict n e , sep within a factor of 1.5 across the three machines, and provide projections to next-step devices (ITER, SPARC, DTT, JT-60SA and COMPASS-U) that are in agreement with available SOLPS simulations.

The compact shunts that are mounted in the upper divertor of ASDEX Upgrade are now operational and measure the thermocurrents generated by the temperature difference between the two target plates of the upper divertor cassette. A thermocurrent is measured when the strike point is on the tile. When a divertor coil current is introduced, the location of the strike point changes. This results in thermocurrent being measured on the corresponding tile. However, the shunt measurements are not toroidally symmetric when gas puffing alters the local conditions in front of a tile, or when a locked mode is present.

The ASDEX Upgrade tokamak (AUG) has been in operation since 1991. After a step-wise introduction of tungsten as plasma facing material, since 2007, AUG has complete tungsten/metal plasma facing components (PFCs). The conversion to tungsten PFCs required some adaptions to the plasma operation, namely enhancements of central wave heating for tungsten accumulation avoidance and a minimum gas puffing rate for reduction of the tungsten source at the outer plasma. AUG operates a versatile heating and current drive system, consisting of neutral beam injection, electron cyclotron resonance heating and current drive and ion cyclotron range of frequencies (ICRF) heating systems as foreseen in ITER. Since high central radiative losses by tungsten were observed during ICRF operation, an improved antenna pair with three antenna stripes was installed which considerably reduced the tungsten source related to ICRF operation. This paper describes aspects of the AUG operation preparing the scientific exploitation, like boronization schedule, plasma breakdown, and startup. Program setup and performance and failure statistics are also presented. Main failure causes are related to deterioration of electronics and the implementation of new actuators into existing, complicated systems.

In next-step fusion tokamaks such as SPARC and ITER, achieving high levels of scrape-off-layer power dissipation will be essential to protect the divertor while maintaining good core plasma performance. The Lengyel model for power dissipation is easy to interpret and fast enough to incorporate into plasma control and scoping tools, but it systematically overestimates the impurity concentration required to reach detachment by a factor of ∼ 5 relative to experiments and higher-fidelity simulations. In this work, we extended the Lengyel model to match the semi-empirical Kallenbach scaling, which successfully describes detachment access on several operating tokamaks. We found that we can reproduce the experimental scaling by accounting for cross-field transport in the divertor, power and momentum loss due to neutral ionization close to the divertor target and turbulent broadening of the upstream heat flux channel. These corrections cause the impurity concentration required for detachment to decrease faster than n e , u 2 , reproducing the c z ∝ 1 / n e , u 2.7 − 3.2 scalings found in experiment. The model also quantitatively reproduces the impurity concentration needed to reach detachment in experiment, demonstrating that the extended Lengyel model can be used as a simple, accurate model for detachment access.

Extrapolating the observed behavior of helium exhaust in current tokamaks towards future reactors requires the understanding of the underlying physical mechanisms determining helium transport, recycling and pumping. Helium compression is the main physics-based figure of merit characterizing how efficiently helium is transported towards the divertor and recycled at the target plates. Moreover, helium gas transport in the subdivertor region towards the pumps is strongly influenced by vessel geometry and installed pumps. The SOLPS-ITER code package is used to model H-mode He-seeded deuterium plasmas at the ASDEX Upgrade tokamak, and compared to recent experiments. The simulations generally indicate a poor recycling of helium in the divertor, compared to that of deuterium, in qualitative agreement with the experiment. This is mainly determined by a deeper edge transport barrier and a weaker parallel SOL transport of He ions, with respect to D ions, and by the higher first ionization energy of He atoms, which results in a deeper penetration of recycled atoms into the plasma. The simulated He compression is, however, much smaller than the experimentally measured one, despite the introduction of additional, non-default physics components into the code. Helium gas transport in the subdivertor region towards the pumps is conductance-limited, but moderately enhanced by the entrainment of He atoms into the stronger, viscous deuterium gas flow via friction. The observed poor helium recycling poses challenges in view of the requirements of helium exhaust in future reactors. Our results emphasize the need to investigate further strategies to optimize helium pumping, to guarantee an efficient removal of helium ash in future burning plasmas. Additionally, the observed difficulty of SOLPS-ITER in reproducing the experimental observations suggests a careful evaluation of the currently available extrapolations of impurity transport towards future devices obtained via edge transport modelling.

Since the publication of the review Progress in the ITER Physics Basis (PIPB) in 2007, significant progress has been made in understanding the processes at the plasma-material interface. This review, part of the ITPA Nuclear Fusion Special Issue On the Path to Burning Plasma Operation, presents these developments, focusing on key areas such as the physics of plasma exhaust, plasma-material interactions, and the properties of plasma-facing materials and their evolution under plasma exposure. The coordinated efforts of the ITPA Topical Group on Scrape-Off Layer and Divertor Physics (DivSOL) have been instrumental in identifying and addressing critical research and development issues in numerous collaborative experimental and modelling projects.

Electron cyclotron resonance heating (ECRH) staircase discharges in strongly shaped plasmas were performed at the full-tungsten ASDEX Upgrade tokamak to investigate the enhanced D _α (EDA) H-mode, a high-confinement regime without edge localized modes (ELMs) that exhibits numerous desirable qualities for future reactors. Heating power, fueling, and plasma current scans reveal rich dynamics as the plasma traverses different confinement regimes. The L-H transition typically occurs with a brief I-phase, sometimes followed by a short nonstationary ELM-free H-mode, before the quasi-coherent mode (QCM) sets in, marking the start of the EDA H-mode. After the pedestal fully develops, the plasma remains stationary until the heating power is raised above a certain threshold, causing ELMs. A novel criterion based on the normality of the divertor shunt current distribution is introduced to identify phases with ELMs, showing general applicability under a wide range of discharges and conditions. The no-ELM power boundary is found to increase with fueling, and too little deuterium gas puff results in a pathological nonstationary ELM-free H-mode without the QCM. Empirical scalings are derived for core, pedestal, and global parameters in EDA H-mode. These show, for example, that pedestal electron pressure increases sublinearly with power and almost quadratically with current. Line-averaged density is approximately proportional to plasma current but very weakly affected by power and fueling, whereas energy confinement time decreases sublinearly with power and increases supralinearly with current. The EDA H-mode achieves several reactor-relevant dimensionless parameters, most notably high Greenwald fraction and confinement enhancement factor over the entire heating power range. This dataset constitutes a versatile resource to plan EDA experiments in present and upcoming devices, also serving as a testbed for validating physics-based theories and models of the regime. Overall, the EDA H-mode remains promising and could become an important no-ELM scenario in future reactors such as SPARC and the full-tungsten ITER.