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In its first phase of exploitation, JT-60SA will be equipped with an inertially cooled divertor, which can sustain heat loads of 10 MW/m[sup 2] on the targets for a few seconds, which is much shorter than the intended discharge duration. Therefore, in order to maximize the duration of discharges, it is crucial to develop operational scenarios with a high radiated fraction in the plasma edge region without unacceptably compromising the scenario performance. In this study, the core and edge conditions of unseeded and neon-seeded deuterium H-mode scenarios in JT-60SA were investigated using METIS and SOLEDGE3X-EIRENE codes. The aim was to determine whether, and under which operational conditions, it would be possible to achieve heat loads at the targets significantly lower than 10 MW/m[sup 2] and potentially establish a divertordetached regime while keeping favorable plasma core conditions. In first analysis, an investigation of the edge parameter space of unseeded scenarios was carried out. Simulations at an intermediate edge power of 15 MW indicate that, without seeded impurities, the heat loads at the targets are higher than 10 MW/m[sup 2] in attached cases, and achieving detachment is challenging, requiring upstream electron densities at least above 4 χ 10[sup 19] m[sup -3]. This points toward the need for impurity injection during the first period of exploitation of the machine. Therefore, neon seeding simulations were carried out, performing a seeding rate scan and an injected power scan while keeping the upstream electron density at the separatrix at 3 χ 10[sup 19] m[sup -3]. They show that at 15 MW of power injected into the edge plasma, the inner target is easily detached and presents low heat loads when neon is injected. However, at the outer target, the heat fluxes are not lowered below 10 MW/m[sup 2], even when the power losses in the edge plasma are equal to 50% of the power crossing the separatrix. Therefore, the tokamak will probably need to be operated in a deep detached regime in its first phase of exploitation for discharges longer than a few seconds. In the framework of core-edge integrated modeling, using METIS, the power radiated in the core was computed for the most interesting cases. [ABSTRACT FROM AUTHOR]

Predictions of the scrape‐off layer (SOL) plasma conditions from the 2D multi‐fluid code EDGE2D‐EIRENE are compared to the OEDGE 1D fluid code for JET low‐confinement mode (L‐mode) plasmas. In the low‐recycling divertor conditions (divertor target plasma collisionality νe≈30$$ {\nu}_e\approx 30 $$), OEDGE and EDGE2D‐EIRENE agree on the electron temperature, the ion temperature, and the electron density within 10% in the low‐field side (LFS) divertor X‐point region (divertor SOL) and within 25% in the LFS midplane region. In high‐recycling conditions (νe≈55,000$$ {\nu}_e\approx 55,000 $$), the predictions of both codes agree on the electron temperature within 20%, while differences in the ion temperature and electron density increase to as high as 50% in the divertor SOL. [ABSTRACT FROM AUTHOR]

X-Point Radiator (XPR) regimes have been obtained in WEST tokamak experiments with nitrogen seeding during the experimental campaigns of 2023 and 2024. These experiments showed the formation of a stable toroidal radiating ring near the X-point, similar to observations in other devices such as JET, ASDEX-Upgrade, TCV, and COMPASS. In WEST, the onset of this regime is associated with a sharp transition of the divertor plasma from hot to cold and dense conditions, with increased particle fluxes, indicating that the plasma is not detached. At the same time, core conditions are significantly improved. These scenarios were successfully controlled in WEST using an interferometry line-of-sight passing through the X-point. Interpretative modeling of these discharges with the SOLEDGE3X-EIRENE code reveals that a physics mechanism needed to stabilize WEST nitrogen XPRs is not present when driving the simulations at constant power. On the contrary, a stable XPR can be obtained by increasing the power injected at the time of the XPR onset to represent the reduction of W contamination, highlighting the need to describe the plasma dynamics and go toward integrated core–edge simulations.

In its first phase of exploitation, JT-60SA will be equipped with an inertially cooled divertor, which can sustain heat loads of 10 MW/m2 on the targets for a few seconds, which is much shorter than the intended discharge duration. Therefore, in order to maximize the duration of discharges, it is crucial to develop operational scenarios with a high radiated fraction in the plasma edge region without unacceptably compromising the scenario performance. In this study, the core and edge conditions of unseeded and neon-seeded deuterium H-mode scenarios in JT-60SA were investigated using METIS and SOLEDGE3X–EIRENE codes. The aim was to determine whether, and under which operational conditions, it would be possible to achieve heat loads at the targets significantly lower than 10 MW/m2 and potentially establish a divertor-detached regime while keeping favorable plasma core conditions. In first analysis, an investigation of the edge parameter space of unseeded scenarios was carried out. Simulations at an intermediate edge power of 15 MW indicate that, without seeded impurities, the heat loads at the targets are higher than 10 MW/m2 in attached cases, and achieving detachment is challenging, requiring upstream electron densities at least above 4 × 1019 m−3. This points toward the need for impurity injection during the first period of exploitation of the machine. Therefore, neon seeding simulations were carried out, performing a seeding rate scan and an injected power scan while keeping the upstream electron density at the separatrix at 3 × 1019 m−3. They show that at 15 MW of power injected into the edge plasma, the inner target is easily detached and presents low heat loads when neon is injected. However, at the outer target, the heat fluxes are not lowered below 10 MW/m2, even when the power losses in the edge plasma are equal to 50% of the power crossing the separatrix. Therefore, the tokamak will probably need to be operated in a deep detached regime in its first phase of exploitation for discharges longer than a few seconds. In the framework of core–edge integrated modeling, using METIS, the power radiated in the core was computed for the most interesting cases.

Modelling of the plasma transport for inherently three-dimensional (3D) problems as in stellarators requires dedicated complex codes. FINDIF is a 3D multifluid plasma edge transport code that has been previously successfully used for the analysis of energy transport in the TEXTOR-DED tokamak [1], where 3D perturbations led to an ergodic structure of field lines in the plasma edge. The ongoing efforts to apply it meaningfully to Wendelstein 7-X (W7-X) plasma problems resulted in advancements in the main model and accompanying tools for mesh generation and post-processing. In order to verify the applicability of the code and to compare with the reported simulation (EMC3-EIRENE) and experimental (OP1.1) results, a series of simulations for varying plasma density, temperature and anomalous transport coefficients as well as for fixed input power were performed. The connection length pattern of FINDIF traced magnetic field lines on the limiter was reproduced and its impact on heat loads was confirmed. An increase in the peak heat load on the limiter with a rise in plasma density, temperature and anomalous plasma transport coefficients was observed. The decay lengths of density, electron temperature and heat flux did not change with density, and were decreasing with temperature and increasing with anomalous plasma transport coefficient, which was compared to the simple scrape-off layer (SOL) model. [ABSTRACT FROM AUTHOR]

Analysis of helicon wave heating and current drive capability in EXL-50 spherical torus plasmas has been conducted. It is found that the driven current increases with the launched parallel refractive index n | | and peaks around n | | = 4.0 when the frequency of the helicon wave is between 300 and 380 MHz. The helicon wave current drive efficiency shows a relatively stable upward trend with increasing plasma temperature. Moreover, the driven current decreases as the plasma density increases. We also analyzed the current drive with helicon waves of 150 MHz and 170 MHz and found that the driven current at a lower frequency was lower than that at a higher frequency. A positive proportional relationship exists between the driven current and n | | . Besides, as n | | increases, the profile of the driven current becomes wider. Finally, the effect of the scrape-off layer (SOL) region on the helicon wave current drive was also investigated. [ABSTRACT FROM AUTHOR]

We report the observation of a quasi-coherent fluctuation (QCF) by the Doppler backscattering system in the scrape-off layer (SOL) region of the DIII-D tokamak. This QCF is observed in high power, high performance hybrid plasmas with near double-null divertor shape during the electron cyclotron heating period. This mode is correlated with a steepened SOL density profile, and leads to significantly elevated particle and heat fluxes between edge localized modes. The SOL QCF is a long-wavelength ion-scale fluctuation (kθ ρs ≈ 0.2-0.4 and kr ρs ≈ 0.03), and propagates in the ion diamagnetic direction in the plasma frame. Its radial expanse is about 1.5-2 cm, well beyond the typical width of heat flux λ q on DIII-D. Also, the SOL QCF does not show any clear dependence on the effective SOL collisionality, and thus may raise issues on the control of plasma-material interactions in low collisionality plasmas in which the blob-induced transport is reduced. A linear simulation using BOUT++ with a five-field reduced model is performed and compared with experimental observations. In simulation results, an interchange-like density perturbation can be driven by the SOL density gradient, and its peak location and the radial width of the density perturbation are in agreement with the experimental observations.

The plasma electron density is measured with multiple diagnostics in the scrape-off layer of the ASDEX Upgrade tokamak. The recorded values are typically in good agreement in the case of exponentially decaying profiles at low densities, but they become inconsistent as soon as a density shoulder is formed, and up to a factor of 3 deviation is experienced at high densities. An idea is put forward; namely, that the local plasma-wall proximity has an influence on the local density and thus the measurement results depend on the local limiter-plasma and limiter-diagnostic distances at the same time. The effect of local wall clearance dependence is demonstrated by creating a plasma having an extremely high scrape-off layer density $\left(n_\mathrm{e} \gt 3\cdot10^{19}\,\textrm{m}^{-3}\right)$ after the H-L back transition and varying the position of the separatrix in different directions. As a possible physical process, neutral recycling at the plasma facing components is identified. Similar behaviour is also observed in the QCE regime, where the appearance of small ELMs barely impacts the scrape-off layer profiles. There are also indications that the local wall clearance does have an impact on the density profile not only in the scrape-off layer, but also in the confined region.

The heat flux control of the divertor plate via strike-point splitting generated by biased targets was proposed in the HL-2A tokamak (Cui et al 2021 Fusion Eng. Des. 173 112963). To understand the helical scrape-off layer (SOL) currents driven by hybrid biasing, two SOL current models (model A and B) are employed. Model A is a simplified 2D model that focuses on investigating the effect of biasing on the sheath and elucidating the fundamental physical mechanism of bias-driven SOL current paths. The potential, charge density, electric field and current densities are calculated. Model B takes into account the actual tokamak geometry and computes the resonant magnetic perturbations (RMPs) generated by bias-driven linear decay currents. Additionally, strike-point splitting is observed in the HL-2A tokamak, indicating that the SOL currents generated by hybrid biasing are capable of generating strong RMPs and consequently influence the magnetic topology. These results confirm the potential of heat/particle flux control by hybrid divertor biased targets.

The impact of plasma shaping on the properties of high density H-mode scrape-off layer (SOL) profiles and transport at the outer midplane has been investigated on Tokamakà configuration variable. The experimental dataset has been acquired by evolving the upper triangularity while keeping the other parameters constant. The scan comprises $\delta_\mathrm{up}$ values between 0.0 and 0.6, excluding negative triangularity scenarios. Within this study, a transition from type-I edge localised modes to the quasi-continuous exhaust regime takes place from low to high $\delta_\mathrm{up}$ . The modification of the upstream SOL profiles has been assessed, in terms of separatrix quantities, within the $\alpha_\mathrm{t}$ turbulence control parameter theoretical framework (Eich et al 2020 Nucl. Fusion 60 056016). The target parallel heat load and the upstream near-SOL density profiles have been shown to broaden significantly for increasing $\alpha_\mathrm{t}$ . Correspondingly, in the far SOL a density shoulder formation is observed when moving from low to high $\delta_\mathrm{up}$ . These behaviours have been correlated with an enhancement of the SOL fluctuation level, as registered by wall-mounted Langmuir probes as well as the thermal helium beam diagnostic. Specifically, both the background and the filamentary-induced fluctuating parts of the first wall ion saturation current signal are larger at higher $\delta_\mathrm{up}$ , with filaments being ejected more frequently into the SOL. Comparison of two pulses at the extremes of the $\delta_\mathrm{up}$ scan range, but with otherwise same input parameters, shows that the midplane neutral pressure does not change much during the H-mode phase of the discharge. This indicates that indirect effects of the change in geometry, linked to first wall recycling sources, should not play a significant role. The total core radiation increases at high $\delta_\mathrm{up}$ , on account of a stronger plasma–wall interaction and resulting larger carbon impurity intake from the first wall. This is likely associated to the enhanced first wall fluctuations, as well as a smaller outer gap and the close-to-double-null magnetic topology at high shaping.

The EMC3-Eirene code was used to study the main impurity leakage mechanism for the island divertor in the standard magnetic field configuration. It was found that under experimentally accessible plasma scenarios in the last experimental campaign, the majority of the island scrape-off layer was friction-force dominated. The impurity force balance was only thermal force dominated for upstream locations closed to the last closed flux surface, beyond the island X-point. No impurity neutral ionization was found in this location and hence the parallel impurity transport provides excellent impurity retention. It was found that impurities approach the confinement region nonetheless via perpendicular transport across the island O-point near the parallel flow stagnation region. This finding points out the specific role of the parallel flow stagnation region in providing lower parallel convective transport and long impurity residence times, which makes non-parallel transport channels more important or even the dominant driver of impurity leakage. In line with the relevance of the particle build-up in the flow stagnation region, different retention behavior as a function of density is seen for various species, which is shown to be due to ionization length changes as the plasma background density is increased.

This paper presents a quantitative validation of Scrape-Off Layer Plasma Simulation-ITER (SOLPS-ITER) simulations against the TCV-X21 reference case and provides insights into the neutral dynamics and ionization source distribution in this scenario. TCV-X21 is a well-diagnosed diverted L-mode sheath-limited plasma scenario in both toroidal field directions, designed specifically for the validation of turbulence codes (Oliveira, Body et al 2022 Nucl. Fusion 62 096001). Five new, neutrals-related observables are added here to the extensive, publicly available TCV-X21 dataset. These are three deuterium Balmer lines in the divertor and neutral pressure measurements in the common and private flux regions. The quantitative agreement metric used in the validation is combined with the conjugate gradient method to approach the SOLPS-ITER input parameters that return the best overall agreement with the experiment. A proof-of-principle test of this method results in a modest improvement in the level-of-agreement; the shortcomings impacting the result and how to improve the methodology are discussed. Alternatively, a scan of the particle and heat diffusion coefficients shows an improvement of 10.4% in the level-of-agreement, approximately twice as high as that achieved by the gradient method. This result is found for an increased transport coefficient compared to what is usually used for TCV L-mode plasmas. The simulations further indicate that ∼65% of the total ionization occurs in the SOL, from which ∼70% in the divertor regions, despite being a sheath-limited regime, motivating the inclusion of self-consistent neutral models in future turbulence simulations on the path towards improved agreement with the experiment.

Coupled plasma-neutral transport simulations are performed on ADITYA-Upgrade tokamak scrape-off layer (SOL) plasma, where flows in the core and SOL were measured to reverse signs with density variation. The simulations performed using the EMC3-Eirene plasma-neutral code combination incorporate the toroidally continuous high-field-side belt limiter placed in a moderate circular tokamak equilibrium. The development of mutually counter-propagating toroidal plasma flows in the top and bottom regions of both the SOL and core is recovered for relatively high upstream density cases with high input power (300 kW and 3 m ^2 s ^−1 ). The origin of the flows is traced to the poloidal density variation introduced by high recycling on the inboard localized belt limiter. The results are compared with similar observations, for example, in Doppler-shifted passive charge exchange line emission on the ADITYA-Upgrade (ADITYA-U) tokamak, highlighting the role played by residual stress in the total Reynolds stress. The external stimuli, such as a localized gas puff, are discussed as potential drivers of flow, via residual stress, based on the existing resonant model of the tokamak plasma rotation.

We study a reduced electrostatic fluid model for the tokamak scrape-off layer, which incorporates temperature gradient and vorticity gradient as two free energy fields. Two scenarios of field coupling are addressed: (1) sheath condition; (2) vortex wave coupling. For the sheath condition induced field coupling, the poloidal $\mathbf{E} \times \mathbf{B}$ flow shear is coupled with the temperature gradient. Combining an eigenmode analysis and the nonlinear phase dynamics approach, our findings indicate that in the absence of a vorticity gradient, the overall effect of the sheath condition induced flow shear can either stabilize or destabilize the interchange mode, depending on the competition between the flow shear suppression and the temperature gradient driving. This is different from the case where the gradient drive and shear damping are decoupled. When the field coupling is mediated by wave interactions, by setting an idealized step-like temperature and vorticity profiles, a joint mode forms through resonant interaction between the interfacial waves driven by the temperature and vorticity gradients, respectively. Near the phase locking condition, the joint mode can be more unstable than pure temperature gradient driven mode.

The main goal of the Divertor Tokamak Test facility (DTT) is to operate with a high value of power-exhaust-relevant parameter PSOL/R in plasma scenarios similar to those foreseen for the Demonstration Fusion Power Plant (DEMO) in terms of low collisionality and neutral opacity. For these unique characteristics, accurate modelling of the principal scenario is necessary for machine designing.In edge numerical codes, cross-field transport profiles have a high impact on modelling results. This work aims at providing a coherent set of transport parameters for DTT full-power (FP) single-null (SN) scenario edge modelling. To evaluate such parameters for DTT, a transport analysis on the current machine has been performed using SOLEDGE2D-EIRENE and SOLPS-ITER.The transport parameters to be used in the simulations of the DTT single-null scenario were selected using two complementary methods. The first is the modelling of JET and Alcator C-Mod (C-Mod) with SOLEDGE2D-EIRENE and SOLPS-ITER, validating transport parameters by comparing modelling results to experimental data from pulses which are considered DTT-relevant. JET pulses were selected with the highest auxiliary input power (from 29 to 33 MW), plasma current and toroidal field to better match DTT parameters; nitrogen and neon seeded pulses were selected to check possible seeding material dependencies. The considered C-Mod pulse better matches DTT plasma density and neutral opacity. Transport parameters are then scaled to DTT according to scaling laws. The second method derives the transport parameters by tuning their values inside the DTT separatrix to reproduce the pedestal profiles predicted by the EPED model via the Europed code and applied in DTT.The derived set of DTT transport parameters is consistent with the results obtained by modelling present machines, reproduces the expected heat flux decay length in detached conditions and, inside the separatrix, reproduces the predicted pedestal using transport parameters which are coherent with what is predicted by the quasi-linear turbulent model QuaLiKiz.

The effect of sheared E  ×  B flow on the blob dynamics in the scrape-off layer (SOL) of HL-2A tokamak has been studied during the plasma current ramp-up in ohmically heated deuterium plasmas by the combination of poloidal and radial Langmuir probe arrays. The experimental results indicate that the SOL sheared E  ×  B flow is substantially enhanced as the plasma current exceeds a certain value and the strong sheared E  ×  B flow has the ability to slow the blob radial motion via stretching its poloidal correlation length. The locally accumulated blobs are suggested to be responsible for the increase of plasma density just outside the Last Closed Flux Surface (LCFS) observed in this experiment. The results presented here reveal the significant role played by the strong sheared E  ×  B flow on the blob dynamics, which provides a potential method to control the SOL width by modifying the sheared E  ×  B flow in future tokamak plasmas. [ABSTRACT FROM AUTHOR]

The plasma background and impurity transport in the scrape-off layer (SOL) of the WEST tokamak are modeled for three discharges over a Psep power scan ranging between 1.5 and 2.35 MW. An extended modeling grid is used to investigate far-SOL sourcing from the WEST baffle and lower divertor. Divertor Langmuir probe data at the targets and upstream data from reciprocating Langmuir probes and interferometry/reflectometry is used to create a simulated background plasma with the OSM and EIRENE codes. W-I spectroscopic data from WEST at the lower divertor and baffle is used to estimate the sputtered W flux as an input to the impurity transport code DIVIMP. A potential mechanism for far-SOL impurities from the baffle to transport into the near-SOL is proposed and discussed in terms of the model’s force balances. Analysis of simulated impurity transport showed similar impurity density at the separatrix for the mid and high-power cases despite the high-power case having 10% more sputtered W. This is explained by the combination of differing trends in the sourcing and screening with scaling power.

Estimates of plasma conditions in the far scrape-off layer (SOL) and of first wall (FW) fluxes in ITER are key input parameters to first wall erosion and impurity migration models, which are in turn involved in the assessment of FW panel lifetime and fuel retention studies. SOLEDGE3X up-to-the wall boundary plasma simulations are performed for ITER, based on an expected Pre-Fusion-Power-Operation (PFPO-1) scenario at PSOL= 20MW, including an impact study of enhanced far-SOL transport. This latter study concerns the possible formation of density shoulders, which are modelled here by applying an increase to the prescribed perpendicular particle and heat diffusivity coefficients maps in the far-SOL in the code, in order to flatten the density and temperature profiles there. Several kinds of such obtained “shoulders” are considered. A brief comparison with SOLPS-ITER is performed on the reference case with uniform coefficients, and shows good agreement. When far-SOL transport is increased, temperatures computed on the first wall rise to 20–30 eV for ions, and to 10–20 eV for electrons. It is also found that for first wall quantities of interest in the ITER machine, the assumed level of perpendicular transport in the far-SOL is the most relevant parameter, with the location at which transport is increased being much less important.

The L-mode density limit in ASDEX Upgrade is investigated with focus on the scrape-off layer. We observe a chain of events that involves the collapse of the E×B outer shear layer (just outside the separatrix), widening of the electron pressure fall-off length, outer divertor detachment, density shoulder formation, reduction and flattening of the parallel Mach number, and enhanced turbulent cross-field transport. The latter effect was associated with higher amplitude density fluctuations at low frequencies, where the turbulent cross-phase between density and velocity fluctuations pointing radially outward was favourable.

Two gas‐puffs are used near limiters in the FT‐2 tokamak for the purpose of hydrogen refuelling during plasma discharges. This creates toroidal and poloidal asymmetry in particle sources near limiters which has to be considered in modelling. Here, the effect of toroidal asymmetry is simulated using the gyrokinetic code ELMFIRE. Two slightly different toroidal particle sources are used in simulations, and their results are compared with each other, and with experimental measurements to understand the impact of toroidal particle sources on Scrape‐off Layer (SOL) physics. [ABSTRACT FROM AUTHOR]

An ion temperature anisotropy model including cross‐field drifts has been implemented in the 2D transport code UEDGE. Two ion temperature equations are solved in the model to give separate ion temperatures perpendicular and parallel to the magnetic field, which results in an additional mirror force for ion parallel transport and also leads to an additional curvature drift in the perpendicular direction. In our previous work, an ion temperature anisotropy model was implemented and applied to the full 2D version of UEDGE including recycled neutrals. The simulation results showed that the mirror force due to anisotropy reduces ion parallel flow velocity, however, the simulations were presented without cross‐field electric and magnetic drifts. Nevertheless, in present medium‐size tokamaks or future compact fusion devices, cross‐fields are important. In this work, the anisotropy model is extended to be compatible with the inclusion of cross‐field drifts. Preliminary results on comparison between the model solutions with and without cross‐field drifts are presented. [ABSTRACT FROM AUTHOR]

Tokamak edge (scrape-off layer (SOL)) plasmas can exhibit non-local transport in the direction parallel to the magnetic field due to steep temperature gradients. This effect along with its consequences has been explored at equilibrium for a range of conditions, from sheath-limited to detached, using the 1D kinetic electron code SOL-KiT, where the electrons are treated kinetically and compared to a self-consistent fluid model. Line-averaged suppression of the kinetic heat flux (compared to Spitzer-Härm) of up to 50% is observed, contrasting with up to 98% enhancement of the sheath heat transmission coefficient, γ _e . Simple scaling laws in terms of basic SOL parameters for both effects are presented. By implementing these scalings as corrections to the fluid model, we find good agreement with the kinetic model for target electron temperatures. It is found that the strongest kinetic effects in γ _e are observed at low-intermediate collisionalities, and tend to increase (keeping upstream collisionality fixed) at increasing upstream densities and temperatures. On the other hand, the heat flux suppression is found to increase monotonically as upstream collisionality decreases. The conditions simulated encompass collisionalities relevant to current and future tokamaks.

Plasma filaments have been measured with alkali beam emission spectroscopy in the plasma edge, divertor island, and scrape-off layer of Wendelstein 7-X. Due to the high intensity of a 1–2 kHz plasma mode, a new, correlation based conditional averaging algorithm was used to search for filaments in the signals. With that method, effects of different magnetic configurations and density levels on filament properties are observed. In configurations where the islands are small and do not play an important role for the connection length topology, filaments behave similar to tokamaks. In contrast, in configurations with larger magnetic islands and more complex connection length profiles, filaments behave quite differently, for instance they may or may not appear in the inner side of the divertor island depending on the plasma parameters. Coupling between the filaments and lower frequency events are also showed. The role of filaments in the global and local particle transport is briefly discussed.

Drifts affect particle, momentum, and energy transport in the scrape-off layer (SOL) of tokamaks and stellarators, altering plasma flows and creating asymmetries between divertors. To understand how drifts affect SOL transport in the W7-X island divertor, an experiment was performed to compare plasmas with matched core parameters but opposite magnetic field directions, and therefore opposite drift transport directions. Parallel flow measurements made with coherence imaging spectroscopy are interpreted with the aid of a diagnostic forward model and a 1D simple SOL model that includes the $\boldsymbol{\mathbf{E}} \times \boldsymbol{\mathbf{B}}$ drift. In low-density plasmas ( $\overline{n}_\mathrm{e} \lt 2 \times 10^{19}$ m ^−3 ), the poloidal $\boldsymbol{\mathbf{E}} \times \boldsymbol{\mathbf{B}}$ drift induces a large poloidal density asymmetry within the island SOL, as measured by divertor Langmuir probes. This in turn causes the parallel flow stagnation point to shift from the position halfway between targets to the X-point in the drift direction, leading to near-unidirectional flow throughout the SOL. As density increases, the effects of the poloidal $\boldsymbol{\mathbf{E}} \times \boldsymbol{\mathbf{B}}$ drift decrease substantially, resulting in a smaller density asymmetry and the development of a counter-streaming flow pattern. For the entire density range probed in this experiment ( $\overline{n}_\mathrm{e} = 1.5-6\times 10^{19}$ m ^−3 ), the experimental observations are more consistent with the effects of the poloidal $\boldsymbol{\mathbf{E}} \times \boldsymbol{\mathbf{B}}$ drift than the radial $\boldsymbol{\mathbf{E}} \times \boldsymbol{\mathbf{B}}$ drift.

Multi-spectral imaging of helium atomic emission (HeMSI) has been used to create 2D poloidal maps of ${{T}}_{\mathrm{e}}$ and ${{n}}_{\mathrm{e}}$ in TCV’s divertor. To achieve these measurements, TCV’s MANTIS multispectral cameras (Perek et al 2019 Rev. Sci. Instrum. 90 123514) simultaneously imaged four He I lines (two singlet and two triplet) and a He II line (468 nm) from passively present He and He ^+ . The images, which were absolutely calibrated and covered the whole divertor region, were inverted through the assumption of toroidal symmetry to create emissivity profiles and, consequently, line-ratio profiles. A collisional-radiative model (CRM) was applied to the line-ratio profiles to produce 2D poloidal maps of ${{T}}_{\mathrm{e}}$ and ${{n}}_{\mathrm{e}}$ . The collisional-radiative modeling was accomplished with the Goto helium CRM code (Zholobenko et al 2018 Nucl. Fusion 58 126006, Zholobenko et al 2018 Technical Report , Goto 2003 J. Quant. Spectrosc. Radiat. Transfer 76 331–44) which accounts for electron-impact excitation (EIE) and deexcitation, and electron–ion recombination (EIR) with ${\textrm{He}}^{+}$ . The HeMSI ${{T}}_{\mathrm {e}}$ and ${{n}}_{\mathrm{e}}$ measurements were compared with co-local Thomson scattering measurements. The two sets of measurements exhibited good agreement for ionizing plasmas: $({5 \,{\textrm{eV}} \leqslant {{T}}_{\mathrm{e}} \leqslant 60 \,{\textrm{eV}}}$ , and ${2 \times 10^{18}\, {\textrm{m}}^{-3} \leqslant{{n}}_{\mathrm{e}} \leqslant 3 \times 10^{19}\,{\textrm{m}}^{-3}})$ in the case of majority helium plasmas, and $({10 \,{\textrm{eV}} \leqslant {{T}}_{\mathrm{e}} \leqslant 40 \,{\textrm{eV}}},{2 \times 10^{18}\,{\textrm{m}}^{-3} \leqslant{{n}}_{\mathrm{e}} \leqslant 3 \times 10^{19} \,{\textrm{m}}^{-3}})$ in the case of majority deuterium plasmas. However, there were instances where HeMSI measurements diverged from Thomson scattering. When ${{T}}_{\mathrm{e}} \leqslant 10 \,{\textrm{eV}}$ in majority deuterium plasmas, HeMSI deduced inaccurately high values of ${{T}}_{\mathrm{e}}{}$ . This disagreement cannot be rectified within the CRM’s EIE and EIR framework. Second, on sporadic occasions within the private flux region, HeMSI produced erroneously high measurements of ${{n}}_{\mathrm{e}}$ . Multi-spectral imaging of Helium emission has been demonstrated to produce accurate 2D poloidal maps of ${{T}}_{\mathrm{e}}$ and ${{n}}_{\mathrm{e}}$ within the divertor of a tokamak for plasma conditions relevant to contemporary divertor studies.

Midplane gas puffing close to the ion cyclotron range of frequencies (ICRF) antennae has been demonstrated to be robust in improving ICRF power coupling in current tokamaks. It is also shown in a previous study (Zhang 2019 Nucl. Mater. Energy 19 364–71) that in ITER, midplane gas puffing with a puff rate of ∼4.5 × 10 ^22 electrons s ^−1 can increase the antenna loading/coupling resistance by about a factor of two. In this paper, a comprehensive parametric study has been carried out to characterize the influence of midplane gas puffing on ICRF power coupling in additional and broader range of parameter scans. The new parameter scans include the gas puff rate, the poloidal location of the gas pipe orifices (GPOs), the separatrix density, the particle perpendicular diffusion coefficient, the radial distance between the plasma and antenna as well as the antenna phasing. The 3D edge plasma fluid and neutral transport code EMC3-EIRENE code has been used to simulate the 3D distributions of plasma density in the presence of gas puffing, which are then used in the antenna code ANTITER II to calculate the antenna coupling. The simulation results indicate that the ITER ICRF local midplane gas injection layout (with the GPOs located on one side of each antenna port) increases the ICRF power coupling significantly in all studied plasma scenarios and antenna parameters. We are hence confident that the chosen layout for the ICRF local gas injection on ITER is appropriate. We are also confident that the ITER local gas injection will allow boosting the ICRF coupling with all studied plasma conditions and antenna phasings.

The dynamics of the screening effect of Ar impurity by the injection of additional Ne has been studied through time-dependent analysis with the integrated divertor code SONIC. In the preceding study (Yamoto et al 2020 Plasma Phys. Control. Fusion 62 045006), the predictive simulation of JT-60SA plasma by SONIC has shown that the injection of additional Ne into Ar-seeded plasma results in lower Ar density and radiation power in the SOL and core edge than in the Ar-only seeded case. The results have demonstrated that the mixed impurity seeding of Ar and Ne may be advantageous for maintaining a high core plasma performance with a low divertor heat load. It was found that the friction force induced by the high D ^+ flow in the SOL towards the inner divertor (ID) region in the Ar + Ne seeded case pushes Ar impurities to the ID. However, the dynamics of D ^+ flow acceleration cannot be interpreted in the previous study because SONIC was a steady state code. In this study, we have developed the time-dependent version of SONIC and applied it to the transient analysis of the injection of additional Ne into Ar-seeded plasma in JT-60SA. When additional Ne is injected, Ne ions stay in the ID plasma near the X-point. As a result, the Ne radiation power increases near the X-point. The electron pressure then decreases due to the radiation cooling and the D ^+ flow is accelerated by the electron pressure gradient. The ion pressure also decreases due to the convection by the accelerated D ^+ flow by electron pressure gradient. The resulting ion pressure gradient further accelerates the D ^+ flow velocity towards the ID. The results suggest that both the high-performance core plasma and the low divertor heat load can be achieved by the Ar + Ne mixed impurity seeding.

This paper focuses on scrape-off layer and divertor modelling of the medium-density single-null scenario of the Divertor Test Tokamak facility (DTT), under construction in Italy. The modelling was performed using the 2D coupled fluid-Monte Carlo code SOLEDGE2D-EIRENE. For DTT pump designing, neutral pressure at the pump aperture below the dome is calculated in deuterium-only cases as well as with impurity seeding with various puffing levels. This scenario analysis also allowed the characterization of detachment in DTT and the influence of pumping on detachment itself. Two different radiating impurities, neon and nitrogen, were tested in the high power scenario to evaluate the minimum impurity concentration required to achieve sustainable conditions at DTT divertor. The sensitivity of the model was studied by varying the impurity concentration; the model shows a hysteresis-like behaviour between the impurity influx and the total impurity content by which detachment is strongly influenced.

The impact of triangularity on edge plasma transport and turbulence is addressed from full 3D turbulence simulations performed with TOKAM3X. Flux driven fluid simulations are run on analytical magnetic equilibria generated with positive and negative triangularity δ in a bottom limiter configuration. The conservation of the energy is assured by the increase of the bottom limiter radial position from δ>0to δ

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 present paper shows the results of the investigation about the effects of the flux-expansion variation at the inner strike point location on the tungsten content in the plasma core, and on the bulk radiation in JET-ILW. The divertor geometries have been designed using the plasma equilibrium code CREATE-NL. The designed divertor geometries were experimentally achieved via a suitable optimization of the divertor coils current on a few discharges with similar additional heating. Diagnostics were used to examine: the tungsten source from the divertor tiles in both inter-ELM and intra-ELM phases, the tungsten content and profile shape inside the plasma core, and the total radiation. Preliminary results suggest that there is not a strong impact of the flux expansion on the tungsten erosion at the strike point. A discharge with high flux compression shows nonetheless a slight increase in core tungsten concentration but this might also be the result of changes in ELMs frequency due to flux expansion variation. The effect on ELMs dynamic has been clearly observed making more difficult the present analysis and should be further studied in future experiments. Edge analysis performed with the SOLEDGE2D code seems to exclude a significant variation of edge transport with flux expansion, pointing to an indirect effect on pumping efficiency and consequently on edge density.

The search for scenarios in which tolerable power exhaust is combined with good confinement must involve high separatrix densities. We present here the first infrared thermography measurements at ASDEX Upgrade in a high density, high confinement H-mode regime in which no type-I ELMs are present. This regime was formerly called type-II ELM or small ELM regime to distinguish it from type-I ELMs. We report on a broadening of the power fall-off length in this regime of up to a factor of four compared to low density inter type-I ELM conditions. This broadening is correlated to an increased filament detection rate as well as an increase in the pressure close to the separatrix. The broadening of the fall-off length decreases the peak heat flux, while the filaments lead to a quasi-continuous transport into the scrape-off layer and onto the divertor.

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

Currents structure in the scrape-off layer (SOL) of a tokamak is analyzed. It is demonstrated that poloidal currents measured in the experiments are a combination of several current types of different physical nature. Besides known Pfirsch-Schlüter (PS) currents and thermoelectric currents, so-called PCC (plate closing currents) flowing to/from the divertor plates are also analyzed. The latter close radial currents in the SOL and below/above the X-point. In particular, current flowing to the outer plate in the private flux region (PFR), opposite to thermolelectric current is predicted for the standard single-null configuration and favorable direction of ∇B drift (∇B drift of ions is directed towards active X-point). In addition, a pair of currents to and away from the outer plate should flow. In the single-null configuration they are often masked by a larger thermoelectric current, however for the connected double null (CDN) case, where thermoelectric current is strongly reduced due to less temperature asymmetry, these currents dominate. The suggested physical model is supported by results of simulations performed with SOLPS-ITER transport code. Simulations were done for ASDEX-Upgrade (AUG), L-mode , single-null configurations, and for Globus-M H-mode, both disconnected and connected double null configurations. Results of the simulations are compared with probe measurements for AUG and Globus-M tokamaks, and reasonable agreement has been found.

High speed visible imaging allows to visualize the 2-dimensional dynamics of fast phenomena in the boundary layer of fusion plasmas. Here we describe the two high speed visible cameras currently operating in the JET tokamak, enabling the simultaneous observation of a large fraction of the JET tokamak vacuum vessel with the appropriate configuration for many different plasma phenomena. As an example we discuss recent observations on the spatial and temporal dynamics of visible emission at the outer wall during Edge Localized Modes (ELMs) and the penetration of Shattered Injected Pellet (SPI) into the plasma.

The thermal helium beam diagnostic at ASDEX Upgrade can access the whole plasma edge region simultaneously with a high spatiotemporal resolution. Manifold dynamics of plasma edge transport is observed for different plasma confinement regimes, comprising L- and H-mode scenarios with type-I and small ELMs as well as ELM free EDA regimes. The measurements show that the plasma dynamics in the confined region close to the separatrix is often dominated by regular high frequency modes (10–100 kHz), whereas filaments with an intermittent occurrence rate of 0.1–2 kHz dominate in particular the far scrape-off layer (SOL). The comparison among different regimes shows a similar filamentary activity between L-mode and inter-ELM H-mode phases. An increased filamentary activity in H-mode, however, is found for rising densities n∕nGW. This is the case for the regime of quasi continuous exhaust (QCE) with three times higher filament velocities compared to a clean inter-ELM H-mode.

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.

With the ongoing effort to establish the Integrated Modelling and Analysis Suite (IMAS) and the Interface Data Structures (IDSs) as the standard relevant software infrastructure for ITER Organization and its parties for all ITER related activities, we report on steps towards the integration of the B2.5-Eirene plasma fluid solver, one of the codes in the SOLPS-ITER code package, within IMAS workflows, by means of a set of modules which allow post-processing of solution data into the IMAS model. This way, after a case is successfully completed, the solver simulation output results can be stored to the IDSs. To provide the means for visualization of the edge plasma state from the IDSs, a plugin for ParaView, an open-source scientific data analysis and visualization application, named ReadUALEdge was developed. With the data displayed within the ParaView application, all tools and utilities provided by ParaView can be used on the data for convenient and efficient analysis. In this article, we give an overview of the IMAS framework and SOLPS-ITER status and present the main aspects and features of the latest ReadUALEdge plugin for the ParaView application. [ABSTRACT FROM AUTHOR]

SOLPS‐ITER L‐mode‐like simulations with the full set of currents and drift velocities activated, and fluid neutrals have been carried out to interpret experimental results obtained in AUG. Drifts are critical to quantitatively reproduce the experimental results; however, simulations without drifts can also reproduce some trends qualitatively. The magnitude and dependence of the peak heat flux onto both targets on the upstream collisionality are, in general, in quantitative agreement within uncertainties with infrared thermography measurements in favourable field direction. The onset of power detachment is observed. In unfavourable toroidal field direction, a more symmetrical inner/outer target solution with regards to the power distribution is predicted, in agreement with experimental observations. However, also in unfavourable toroidal field direction, insufficient power is dissipated in the simulations and therefore qpeak, inn is overpredicted by up to a factor of 4 and qpeak, out by up to a factor of 1.5. The largest contribution to the sources due to radial transport in the energy balance equation is the radial divergence of the energy flux due to VE × B. [ABSTRACT FROM AUTHOR]

Advanced divertor configurations modify the magnetic geometry of the divertor to achieve a combination of strong magnetic flux expansion, increased connection length and higher divertor volume—to improve detachment stability, neutral/impurity confinement and heat‐channel broadening. In this paper, we discuss the modification of the flux‐coordinate independent (FCI) turbulence code GRILLIX to treat generalized magnetic geometry, to allow for the investigation of the effect of magnetic geometry on turbulent structures in the edge and scrape‐off layer (SOL). The development of grids and parallel operators from numerically defined magnetic equilibria is discussed, as is the application of boundary conditions via penalization, with the finite‐width method generalized to treat complex non‐conformal boundaries. Initial testing of hyperbolic (advection) and parabolic (diffusion) test cases are presented for the Snowflake scenario. [ABSTRACT FROM AUTHOR]

Applying gyrokinetic simulations in theoretical turbulence and transport studies for the plasma edge and scrape‐off layer (SOL) presents significant challenges. To particularly account for steep density and temperature gradients in the SOL, the "full‐f" code PICLS was developed. PICLS is a gyrokinetic particle‐in‐cell (PIC) code, is based on an electrostatic model with a linearized field equation, and uses kinetic electrons. In previously published results, we applied PICLS to the well‐studied 1D parallel transport problem during an edge‐localized mode (ELM) in the SOL without collisions. As an extension to this collision‐less case and in preparation for 3D simulations, in this work, a collisional model will be introduced. The implemented Lenard–Bernstein collision operator and its Langevin discretization will be shown. Conservation properties of the collision operator, as well as a comparison of the collisional and non‐collisional case, will be discussed. [ABSTRACT FROM AUTHOR]

The goal of this article is to show that the "intermittent" convective transport in the scrape-off layer (SOL) of linear magnetic confinement plasmas actually takes place periodically. This is reflected in several features that are unveiled in this article. It is found that the distribution of the time between high-amplitude bursts, selected above a given threshold, has a central frequency in the SOL. Moreover, this distribution is found similar to the one taken inside the main plasma column as the threshold is reduced, suggesting links between the SOL and the main plasma column. The power spectrum of the radial flux is determined, indicating the presence of a coherent peak that accounts for about 50% of the transport. Using two probes that are poloidally and radially separated, it is shown that a burst in the SOL is correlated to a wave packet inside the main plasma column. The relation between the two distant regions is thus assured by avaloids that are nonlocal structures in nature as they have velocities much higher than the background. (C) 2003 American Institute of Physics.

• A new fast divertor IR system was put into operation on the COMPASS tokamak. The system provides in-situ calibration possibility using a special heated divertor tile. • Radial profiles of the divertor heat flux are routinely measured for both the inner and the outer divertor target with spatial resolution ∼1 mm and frequency up to ∼60 kHz. • First experimental divertor heat flux measurements using the new system were successfully performed in both L-mode and H-mode. A new fast divertor infra-red thermography system was put into operation on COMPASS. It provides full radial coverage of the bottom open divertor with pixel resolution ∼0.6–1.1 mm/px on the target surface and temporal resolution better than 20 μs. The system consists of fast IR camera TELOPS Fast-IR 2K placed in a magnetic shielding box, a positionable holder, a 1 m long IR endoscope consisting of 14 Ge and Si lenses securing off-axis view from an upper inner vertical port and a special graphite divertor tile optimized for IR thermography. The tile is equipped with a heating system allowing tile preheating up to 250 °C. Embedded thermoresistors and a calibration target (a deep narrow hole acting as a black body radiator) allows in-situ calibration of the system including estimation of the target surface emissivity. Furthermore, a roof-top shaped structure on top of the tile increases magnetic field incidence angles above 3 degrees. Laboratory tests of the system performed during its commissioning are presented. The global transmission of the optical system was found to be τ ≈ 40–50%. Poor spatial resolution compared to the design value was observed. Too large surface error of individual lenses was identified as the main cause and re-manufacturing of the most critical lens was suggested. First experimental results obtained using the IR system are presented: divertor heat flux profiles in L-mode with the heat flux decay length λ q omp = 2.1 − 3.3 mm and average H-mode heat flux profiles in an inter-ELM period and during an ELM heat flux maximum with λ q omp ≈ 0.6 mm and λ q omp ≈ 6.7 mm, respectively. [ABSTRACT FROM AUTHOR]