Transient radiative bursts have been observed in systematic time-correlation with dust influx across the separatrix in JET ILW campaigns, [1]. Dust influx from PFC materials can produce transient temperature and density perturbations in the SOL. Are such local perturbations detectable by existing diagnostics? Could they be used as an additional monitor for safe operation? Results from the model illustrated here indicate that local densities of dust can produce bursts of acoustic waves in the low-temperature, low-density SOL, that could be detected by existing diagnostics.

Evidence that density shoulder broadening is dependent on high main-chamber neutral density is presented. Shoulder broadening does not occur when the sources for main-chamber neutrals are minimized using divertor baffles and wide gaps to the first wall (~3× the density decay length). Removing the baffles or reducing the gap to the inner wall both act to increase the density shoulder amplitude in otherwise identical TCV discharges. Radial turbulent transport is correlated with shoulder amplitude.

JT60SA is a tokamak constructed in Naka, Japan to support the operation of the ITER reactor [1]. In the initial research phase the device will be equipped with a full carbon first wall with no active cooling. For this reason, the preparation and assessment of the full inductive scenario with 26.5 MW of input power (the maximum available during this phase) is particularly challenging. The main aim of this work is to assess the realistic heat loads on the first wall of the vessel and to check the overall compatibility of the scrape-off layer plasma (SOL) regime with the desired core performance. To achieve this scope, the entire SOL volume within the vessel and the subdivertor region is simulated by the fluid transport code SOLEDGE3X [2], which uses a mesh extended up to the first wall and can besides handle the secondary X-point situated at the top of the tokamak vessel. Depending on the regime of operation, the interplay between the transport and atomic processes (ionization/recombination) makes this area interesting from the point of view of energy dissipation. The neutrals playing a key role in the process of detachment are treated kinetically by EIRENE Monte Carlo code [3] including additional reactions involving D2 molecules (charge exchange and elastic collisions) [4,5]. A scan on the input power is performed to estimate the maximum total power that can be handled by the divertor targets (maximum power density 10 MW/m2) and other plasma facing components: the dome (10 MW/m2), baffles (up to 1 MW/m2) and the rest of the wall (0.3 MW/m2), as specified in the Plant Integration Document. It is found that in certain areas a substantial fraction of the deposited power is delivered by the neutrals (up to 40%). This leads to the assesment of the desired range of radiated power fraction by intrinsic carbon or possible seeded impurities with respect to the mentioned material limits and H-mode operation requirements (power through separatrix >= ~10 MW) [6]. The results are compared with the previous simulations of the nominal fully inductive scenario (auxiliary heating power 41 MW) with Ar and Ne seeding [7].

Understanding the plasma dynamics in the edge and Scrape-Off Layer (SOL) regions of a tokamak is mandatory for future exploitation of fusion power as a source of energy, since the SOL transport properties determine the peak particle and heat fluxes flowing towards the plasma-facing components (PFCs). The inherently turbulent nature of SOL transport makes this assessment particularly challenging. In the first part of this contribution, an investigation is reported on a set of H-mode discharges on the TCV tokamak, focused on the role of gas fueling in determining the properties of transport and kinetic profiles in the near and far SOL. In all discharges additional NBI heating was used for H-mode access, while different gas settings allowed to span a wide interval of divertor neutral pressure ... . The analysis has been conducted by first estimating the SOL power width ?... from a fit of the upstream-remapped parallel heat flux profile as measured by IR cameras. This allowed to estimate the separatrix temperature T...,?... under the assumption of conduction-dominated parallel transport (justified by the considerably lower temperatures at the target than upstream). After fitting the Thomson Scattering density and temperature profiles, the separatrix position and density n...,?... could be estimated. The level of turbulent transport has been quantified by the ?? ? Z... R...q...?...? n...,?...T...,?... ? parameter introduced in [1], this being a measure of the effect of the interchange instability drive on drift waves and thus effectively an estimate of the interchange turbulence level, being directly related to the separatrix collisionality. An increase in p...,?...? is seen to translate into an increase of n...,?... and ultimately of ??. Limited to the high-density part of the database, a p...,?...? scan between 20 and 120 mPa produces a variation of ?? between 0.4 and 0.95. This is accompanied by a broadening of ?...by up to a factor ~2.5 (from ~4 to ~10 mm), coherently with recent modelling results predicting a larger SOL power width in H-mode at high separatrix collisionality [2]. A similar increase of the near-SOL electron density, temperature and pressure e-folding lengths is observed as well. In the far-SOL the density profile shows a progressive flattening at increasing ??, leading to formation of the so-called density shoulder. This observation can be accounted for by a progressive increase of filamentary transport into the far SOL at higher ??, as observed by wall-mounted and reciprocating Langmuir Probes. Given that ?? shows an explicit dependence on q...?... , a scan in upper triangularity has been performed in H-mode on TCV as well, where ?... has been varied between 0 and ~0.45. Preliminary results show a transition from a Type-I ELMy regime towards a small ELM/QCE regime as ??... increases at similar fuelling levels, coherently with previous experimental observations [3, 4]. In the second part of this work the effect of such a ??... variation on ?? and consequently on SOL profile and turbulence properties will be investigated.

JT60SA is a tokamak constructed in Naka, Japan to support the operation of the ITER reactor [1]. In the initial research phase the device will be equipped with a full carbon first wall with no active cooling. For this reason, the preparation and assessment of the full inductive scenario with 26.5 MW of input power (the maximum available during this phase) is particularly challenging. The main aim of this work is to assess the realistic heat loads on the first wall of the vessel and to check the overall compatibility of the scrape-off layer plasma (SOL) regime with the desired core performance. To achieve this scope, the entire SOL volume within the vessel and the subdivertor region is simulated by the fluid transport code SOLEDGE3X [2], which uses a mesh extended up to the first wall and can besides handle the secondary X-point situated at the top of the tokamak vessel. Depending on the regime of operation, the interplay between the transport and atomic processes (ionization/recombination) makes this area interesting from the point of view of energy dissipation. The neutrals playing a key role in the process of detachment are treated kinetically by EIRENE Monte Carlo code [3] including additional reactions involving D2 molecules (charge exchange and elastic collisions) [4,5]. A scan on the input power is performed to estimate the maximum total power that can be handled by the divertor targets (maximum power density 10 MW/m2) and other plasma facing components: the dome (10 MW/m2), baffles (up to 1 MW/m2) and the rest of the wall (0.3 MW/m2), as specified in the Plant Integration Document. It is found that in certain areas a substantial fraction of the deposited power is delivered by the neutrals (up to 40%). This leads to the assesment of the desired range of radiated power fraction by intrinsic carbon or possible seeded impurities with respect to the mentioned material limits and H-mode operation requirements (power through separatrix >= ~10 MW) [6]. The results are compared with the previous simulations of the nominal fully inductive scenario (auxiliary heating power 41 MW) with Ar and Ne seeding [7].

A multi-function divertor feedback control system has been built on Experimental Advanced Superconducting Tokamak (EAST) to treat the divertor heat load issue. With the real-time data of the Langmuir probes and the impurity seeding, the divertor electron temperature (Te,div) is well controlled to achieve the partial detachment phase. The first trial of the CD4 seeding for the Te,div reduction has been achieved on EAST long-pulse discharge. In the seeding phase, the Te,div was maintained close to 5 eV, and the surface temperature of the target plate (Tsurface,div) had a reduction of about 150 oC. The plasma stored energy had a reduction in the control phase, so it is necessary to find a way to keep the good plasma confinement in the next step. The CD4 injection also mitigated the low hybrid wave coupling rate in some degree. The big volume of the CD4 injection lifted the Greenwald density fraction from ~0.4 to ~0.7, which made the SOL into high recycling state. Most of the injected carbon particles were in the high ionized state, and with the lower of the Te,div, the tungsten line emission was suppressed obviously.

Plasma Exhaust and Plasma Wall Interaction are subjects of intense studies in fusion energy research for the understanding of the amount of heat loads and the lifetime of Plasma Facing Components. In order to ensure reliable predictive edge modeling in this context, it is mandatory to determine the transport properties of the Scrape Off Layer (SOL), a region largely influenced by the presence of turbulent filaments which contribute to particle and energy losses in both L and H modes. From the ITER divertor perspective, to keep the power fluxes acceptable for target material, high neutral pressure and partial detachment are needed to ensure maximum tolerable loads 1. Thus experimental investigation of SOL transport needs to be extended to these regimes.

Plasma Exhaust and Plasma Wall Interaction are subjects of intense studies in fusion energy research for the understanding of the amount of heat loads and the lifetime of Plasma Facing Components. In order to ensure reliable predictive edge modeling in this context, it is mandatory to determine the transport properties of the Scrape Off Layer (SOL), a region largely influenced by the presence of turbulent filaments which contribute to particle and energy losses in both L and H modes. From the ITER divertor perspective, to keep the power fluxes acceptable for target material, high neutral pressure and partial detachment are needed to ensure maximum tolerable loads 1. Thus experimental investigation of SOL transport needs to be extended to these regimes.

An accurate study of the motion of particles of variable mass and charge, in interaction with the SOL plasma, requires considering in the dynamic model, all the effects of the same order of magnitude. Here we present an extension of the conventional point-like particle model introducing a discussion of the rocket acceleration consequent to the asymmetric mass evaporation of a particle traveling in the SOL. A simplified but realistic description of the asymmetric loss of mass is constructed in analogy with the physics of comets.

A detailed cross-device investigation on the role of filamentary dynamics in high-density regimes has been performed within the EUROfusion framework, comparing the ASDEX Upgrade (AUG) and TCV tokamaks. Both devices run density ramp experiments at different levels of plasma current, keeping the toroidal field or q95 constant in order to disentangle the role of the parallel connection length and the current. During the scan at a constant toroidal field, in both devices the scrape-off layer (SOL) profiles tend to develop a clear SOL density shoulder at a lower edge density whenever the current is reduced. Different current behaviour is substantially reconciled in terms of the edge density normalized to the Greenwald fraction. During the scan at constant q95 AUG exhibits similar behaviour, whereas in TCV no upstream profile modification signature has been observed at lower current levels. The latter behaviour has been ascribed to the lack of target density rollover. The relation between the upstream density profile modification and detachment condition has been investigated. For both devices the relation between blob size and the SOL density e-folding length is found independent of the plasma current, with the observation of a clear increase in blob size and the edge density normalized to a Greenwald fraction. ASDEX Upgrade has also explored filamentary behaviour in the H-mode. The experiments in AUG have focused on the role of neutrals, performing discharges with and without cryogenic pumps, highlighting how high neutral pressure, not only in the divertor but also at the midplane, is needed in order to develop an H-mode SOL profile shoulder in AUG.

The formation of flat density profile in the low field side Scrape Off Layer (SOL), the so called shoulder formation, is characterized in the ASDEX Upgrade tokamak. The two main candidate mechanisms for its formation, i.e. modification of SOL turbulence and divertor plasma-neutrals interaction, are studied. Turbulence and blobs are characterised with midplane Langmuir probes; divertor neutrals are estimated measuring the 2D map of Da and Dg emission through the tomographic inversion of their brightness measured by two cameras. The strong interplay between upstream profiles, up-stream turbulence and divertor conditions will be highlighted.In the considered ASDEX Upgrade data set, upstream midplane density shoulder forms when increasing the electron density and it is linked with detachment. The shoulder formation is correlated with the change of edge turbulent transport, resulting in an increase of the blobs perpendicular size, and also in a change in the neutral emission, reflecting a modification in the neutral distribution in the divertor region. In particular, while the neutral D emission is localized in the inner divertor at low electron density, the emission moves towards the outer divertor target and then covers the whole lower divertor region when the upstream electron density shoulder forms.

The effect of magnetic geometry on scrape-off layer (SOL) transport and detachment behaviour is investigated on the TCV tokamak with the goal of assessing the potential of alternative divertor geometries and for the validation of theoretical models. L-mode experiments reveal that increasing connection length and hence divertor volume by either increasing poloidal flux expansion or divertor leg length have different effects on the boundary plasma. In attached conditions, the SOL heat flux width q inferred from target infrared thermography measurements is weakly dependent on poloidal flux expansion but increases approximately with the square root of the divertor leg length. The divertor spreading factor S shows no clear trend with leg length but decreases with flux expansion. TOKAM3X turbulence simulations of the leg length scan are in qualitative agreement with the experiment and can explain observations by a strongly asymmetric (ballooning) transport at and below the X-point. Evidence for increased transport in the region of low poloidal field is obtained in the Snowflake minus geometry. The presence of an additional X-point in the low-field side SOL increases the effective SOL width by approximately a factor two. Increasing flux expansion and leg length both result in enhanced divertor radiation levels, with the effect being much larger in the latter case. This behaviour, together with the observed trend in q, is consistent with a substantial drop in the density threshold for divertor detachment with increasing leg length and a weak variation with flux expansion. Novel spectroscopic techniques reveal that the drop in target ion current and access to detachment is caused by a reduction of the divertor ionization source due to power starvation, while volume recombination is only a small contributor. This interpretation is confirmed by SOLPS modelling. TCV alternative divertor studies are being extended to neutral beam heated H-mode plasmas. The H-mode power threshold is found to vary weakly between standard, X-, and Super- X geometries. In all cases, ELMy H-mode is obtained at intermediate current, while the discharges are ELM-free at high current. Signs of detachment have so far only been observed in the latter case. Ongoing experiments further investigate H-mode detachment in these plasmas and will be extended to Snowflake configurations.

Addressing the role of scrape off layer filamentary transport is a subject of intense studies in fusion science. Intermittent structures dominate transport in L-Mode and strongly contribute to particle and energy losses in H-mode. The role of convective radial losses has become even more important due to its contribution to the shoulder formation in L-Mode, describing the progressive flattening of the density scrape off layer profile at high density r1- 3s. Investigation of this process revealed the strong relationship between divertor conditions and the upstream profiles, mediated by filaments dynamics which varies according to the downstream conditions. Preliminary investigations suggested that similar mechanisms occur in H-Mode r1s and that filaments contribute the SOL transport in H-mode density limit (HDL) as well r4s. The present contribution will report on results obtained on ASDEXUpgrade and TCV tokamaks, to address the role of filamentary transport in high density regimes both in L- and H-Mode. The combined results enlarge the operational space, from a device with a closed divertor, metallic first wall and cryogenic pumping system to a carbon machine with a completely open divertor. The mechanism of shoulder formation and the role of filaments have been tested against variation of plasma current, magnetic configuration (single and double null plasmas), and divertor neutral densities, through modification of cryopump efficiency. At constant magnetic field the density decay length increases with filament-size independently of the plasma current for both machines in L-mode, consistently with the fact that upstream profiles and divertor neutral pressure exhibit the same trend with normalized Greenwald fraction.

In a wide variety of natural and laboratory magnetized plasmas, filaments appear as a result of interchange instability. These convective structures substantially enhance transport in the direction perpendicular to the magnetic field. According to filament models, their propagation may follow different regimes depending on the parallel closure of charge conservation. This is of paramount importance in magnetic fusion plasmas, as high collisionality in the scrape-off layer may trigger a regime transition leading to strongly enhanced perpendicular particle fluxes. This work reports for the first time on an experimental verification of this process, linking enhanced transport with a regime transition as predicted by models. Based on these results, a novel scaling for global perpendicular particle transport in reactor relevant tokamaks such as ASDEX-Upgrade and JET is found, leading to important implications for next generation fusion devices.