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Combining high plasma core performance with a suitable power exhaust solution is one of the major challenges in magnetic confinement fusion research. One of the most important power exhaust parameters is the power fall-off length in the scrape-off layer. Two infrared thermography based power fall-off length data sets from JET operated with carbon and ITER-like wall are revisited and compared to recently published scaling laws as well as to confinement and pedestal top parameters. It is shown that the power fall-off length is correlated to confinement, with the highest correlation among the tested parameters being the pedestal top density. The power fall-off length decreases with increasing pedestal top density in variance to the multi-machine scaling law. A similar trend is observed for the pedestal top pressure. This is in agreement with findings at C-Mod showing a scaling of the power fall-off length in various confinement regimes with the volume averaged pressure. Further, it is shown that a variation of the safety factor at constant pedestal top density is not changing the power fall-off length significantly in the two JET data sets. [ABSTRACT FROM AUTHOR]

A cross-regime (L-mode, I-mode and H-mode) database combining scrape-off layer (SOL) power decay length λq divertor measurements and upstream SOL electron pressure, temperature and density decay lengths has been assembled at ASDEX Upgrade. It is found that a cross-regime λq scaling is best described by a local edge quantity, such as the edge electron pressure evaluated at ρpol = 0.95. Furthermore, λq exhibits a clear correlation with edge electron pressure gradient lengths, no matter if taken inside or outside the separatrix. In addition, the database reveals that SOL and pedestal electron pressure gradients are remarkably well correlated across all confinement regimes. The physical interpretation of this observation is discussed with regard to an edge pressure critical gradient paradigm governing the edge physics and to a turbulence spreading in the SOL. Moreover, it is shown that the Spitzer–Härm electron conduction regime is a reasonable approximation to estimate λq across different confinement regimes. The main implication of these findings is that a widening of λq is linked to a reduction of edge electron pressure gradients. [ABSTRACT FROM AUTHOR]

Probe measurements of plasma fluctuations in the divertor region of ASDEX Upgrade reveal the existence of two quiescent regions close to the separatrix: one on the low field side scrape-off layer and one on the high field side of the private flux region. The X-point manipulator at ASDEX Upgrade provides the unique possibility of obtaining continuous measurements of the ion-saturation current from the low field side scrape-off layer through the private flux region and into the high field side scrape-off layer. [ABSTRACT FROM AUTHOR]

Tearing modes are a major concern for large tokamak devices. Therefore their detection and characterization is of importance for timely countermeasures to avoid significant impact on the discharge or at least prevent possible machine damage. In case of phase-locked tearing modes, the poloidal variation of induced magnetic field fluctuation depends on the amplitudes of and the phase difference between the individual modes. This affects mode detection and identification when not considered appropriately. The phase between phase-locked (2, 1) and (3, 1) tearing modes in ASDEX Upgrade has been determined from local electron temperature and magnetic measurements independently. It is shown that the modes can be in phase at any poloidal position starting from the low field side plasma midplane over the plasma top to the high field side midplane. This observation invalidates the widespread assumption that phase-locked tearing modes are in phase near the low field side midplane. Dependence of the in-phase position on both, the plasma pressure and the plasma rotation velocity, is observed. [ABSTRACT FROM AUTHOR]

For the design and operation of large fusion devices, a detailed understanding of the power exhaust processes is necessary. This paper will give an overview of the current research on divertor power load mechanisms. The results shown are obtained in JET with the ITER-like wall (ILW)and ASDEX-Upgrade with tungsten coated plasma-facing components (PFCs). The challenges of infrared thermography on an ITER-like bulk tungsten divertor are presented. For the steady-state heat load, the power fall-off length λq in JET-ILW is compared to an empirical scaling found in JET and the ASDEX-Upgrade with carbon PFCs. A first attempt to scale the divertor broadening S in the ASDEX-Upgrade with tungsten PFCs is shown. The edge localized mode (ELM) duration tELM in JET-C and JET-ILW is compared. For similar pedestal conditions (Te,ped and ne,ped), similar ELM durations are found in JET-C and JET-ILW. For higher ne,ped at the same pedestal pressure pe,ped, longer ELM durations are found in JET-ILW. The pedestal pressure pe,ped is found to be a good qualifier for the ELM energy fluency in both JET-C and JET-ILW. Improved diagnostic capabilities reveal ELM substructures on the divertor target occurring a few milliseconds before the ELM crash. [ABSTRACT FROM AUTHOR]

The detachment cliff is a bifurcative transition to partial detachment recently discovered at the DIII-D tokamak (McLean et al 2015 J. Nucl. Mater. 463 533–6). This work presents a database analysis of target parameters in L-mode and H-mode discharges to search for a detachment cliff at ASDEX Upgrade (AUG). Most of the transitions from attached to partially detached divertor conditions observed in H- and L-mode discharges in AUG show bifurcative-like characteristics that are consistent with the properties of the detachment cliff if the B × ∇ B drift is directed towards the active X-point. In the operational space of power and density, the bifurcative transitions identified during an L-mode discharge occur at injected power and density higher than a threshold value ( P tot > 0.7 MW and n e > 1.6 × 10 19 m−3, respectively). Furthermore, the temperatures at which the transitions start are found to be insensitive to the injected impurity, the injected power and the value of the upstream density. Finally, the study of the evolution of the target parameters, of the intensity of the D α line and of specific manometers and bolometer lines of sights shows that the physical process underlying the detachment cliff and the self-sustained divertor oscillations (Heinrich 2020 Nucl. Fusion 60 076013) might be the same. [ABSTRACT FROM AUTHOR]

L-mode negative triangularity (NT) operation is a promising alternative to the positive triangularity (PT) H-mode as a high-confinement edge localised mode-free operational regime. In this work, two TCV Ohmic L-mode core density ramps with opposite triangularity δ ≃ ± 0.3 are investigated using SOLPS-ITER modelling. This numerical study aims to investigate the power exhaust differences between NT and PT focusing, in particular, on the geometrical effect of triangularity. To disentangle the latter from differences related to cross-field transport, anomalous diffusivities for particle ( D n A N ) and energy ( χ e / i A N ) transport are fixed to the same values in PT and NT. The simulation results clearly show dissimilar transport and accumulation of neutral particles in the scrape-off layer for the two configurations. This gives rise to different ionization sources in the edge and divertor regions and produces differences in the poloidal and cross-field fluxes, ultimately leading to different power and particle divertor fluxes in the two configurations. Simulations recover the experimental feature of a hotter and attached outer target ( T e , OSP NT ≳ 5 eV ) in the NT scenario compared to the PT counterpart. [ABSTRACT FROM AUTHOR]

An extensive analysis of the global energy balance on the ASDEX Upgrade (AUG) tokamak has been performed. For such a purpose a complete revamping of the AUG cooling water calorimetry originally described in (Richter and Vernickel 1994 Rev. Sci. Instrum. 65 5) has been done. The system, which has been largely improved, consists of 88 cooling units, which provides an almost perfect toroidal and poloidal coverage of the in-vessel components of the AUG tokamak. To measure the energy loss due to radiation, which is only partially captured by the cooling water calorimetry, a novel framework is introduced to calculate the losses utilizing bolometric data. The analysis of the global energy balance based on almost 600 discharges reveal the remarkable capability to account more than 95% of the total injected energy on average. [ABSTRACT FROM AUTHOR]

The magnetic field dependence of intermittently expelled density filaments (blobs) is investigated in the scrape-off layer of ASDEX Upgrade low confinement (L-mode) plasmas. It is demonstrated that lithium beam emission spectroscopy can be used to determine the frequency, radial size and velocity of the blobs. The measured radial blob sizes depend only weakly on magnetic field B. Normalizing the blob sizes to the drift parameter ρs ∝ B−1 results in a large variation beneficial for a quantitative comparison with theoretical blob scaling laws. The blob velocity scales inversely proportional to the square of the blob size in agreement with analytic models for blobs in the sheath-connected regime. The measurements point towards an influence of finite ion temperature on radial blob transport. [ABSTRACT FROM AUTHOR]

To facilitate the estimation of heat loads on plasma-facing components in fusion devices in various different magnetic geometries, a heat load proxy model was developed based on anisotropic diffusion. In this work, this model is compared to the so-called field-line diffusion approach. To facilitate the evaluation of these models, a novel synthetic camera-based approach for obtaining heat load distributions from Monte Carlo samples was also developed and implemented. With the assistance of this synthetic camera, heat load predictions for the Wendelstein 7-X divertor were obtained and compared with infrared camera observations. It was found that the anisotropic diffusion-based model achieved a closer match to infrared camera observations, while still being suitable in computational effort for large magnetic configuration database scans. [ABSTRACT FROM AUTHOR]

Infinite toroidal mode number (n = ∞) ballooning mode analysis of negative triangularity discharges on DIII–D shows that the access to 2nd stability becomes strongly restricted when the top triangularity decreases even modestly from −0.18 to −0.36. This is observed in experiment to coincide with the suppression of the L–H-transition. Further theoretical analysis with ballooning mode limited pedestals shows that the threshold for opening the 2nd stability access rises from a pedestal temperature of 0.3–1 keV when the top triangularity is decreased from −0.18 to −0.36 indicating that to access the 2nd stability with the more negative triangular shape would require unrealistically high pedestal temperature. The pedestal predicted by the EPED code agrees with the experimental H-mode profile for the negative triangularity case but in contrast to positive triangular shapes the prediction shows no increase in pedestal height with increasing core pressure when triangularity is negative. This work provides a first model to predict when negative triangularity plasmas can be expected to access the H-mode. [ABSTRACT FROM AUTHOR]

It is critical to keep a low level of tungsten impurity concentration in the core plasma of the fusion reactor. Using the OEDGE code, the influence on the tungsten concentration due to the cross-field transport in the scrape-off layer is studied. Based on the design parameters of the China Fusion Engineering Test Reactor, the background plasma with the peak target electron temperature of 15 eV is generated using the onion-skin model. The neon impurity with the assumed concentration of 0.2% inside the separatrix is included for estimating the tungsten sputtering source. In a series of simulations of tungsten transport, the diffusion coefficient D⊥ is varied from 0.1 to 10 m2 s−1, and the hypothetical inward pinch velocity VIMP is in the range of 0–5 m s−1. The tungsten concentration inside the separatrix increases with decreasing diffusion coefficient and exceeds 10−5 for the diffusion coefficient lower than about 0.2 m2 s−1. When the hypothetical inward pinch is included, the tungsten concentration increases. The relation between the relative increase in tungsten concentration and VIMP/D⊥ is fitted. [ABSTRACT FROM AUTHOR]

To investigate the optimal scenario of impurity seeding to obtain divertor plasma detachment for target protection, experiments with Ar&D2 seeding from two different poloidal locations, the upper outer (UO) divertor target and lower outer (LO) target, were carried out on EAST. Partial energy detachment (the electron temperature near the strike point Te,spt ⩽ 10 eV) were obtained with Ar&D2 mixture puffing from the UO target and, for the first time, from the LO target into H-mode plasmas in the upper single null (USN) configuration. The peak heat flux qt on the UO target was significantly reduced (by ∼80%). The rollover of ion flux density js did not appear probably due to insufficient momentum loss, which is independent of the puff locations. The poloidal asymmetries of particle and heat fluxes on the targets have also been investigated. The UO-dominant asymmetry of particle flux was reversed, while the UO-dominant asymmetry of heat flux was mitigated but not reversed. The plasma confinement dropped by only 14% and 8.2% in the LO-puff case and UO-puff case, respectively, during detachment. The high level of C may contributed greatly to the higher radiation in the bulk plasma region and the greater decline in WMHD in the LO-puff case. [ABSTRACT FROM AUTHOR]

A database of DIII-D plasmas without edge-localized modes (ELMs) compares the operating space and plasma performance of stationary no-ELM regimes found in conventional tokamaks: ELM suppression with resonant magnetic perturbations (RMPs), quiescent H-mode (QH, including wide-pedestal variant), improved confinement mode (I-mode), enhanced D-alpha H-mode (EDA-H), conventional low-confinement mode (L-mode), and negative triangularity L-mode (Neg-D). Operational space is documented in terms of engineering and physics parameters, revealing divergent constraints for each regime. Some operational space discriminants (such as pedestal collisionality) are well known, while others, such as low torque & safety factor, or high power & density, are less commonly emphasized. Normalized performance (confinement quality and normalized pressure) also discriminate the no-ELM regimes and favor the regimes tolerant to power in DIII-D: RMP, QH, and Neg-D. Absolute performance (volume-averaged pressure , confinement time τ, and triple product) also discriminates no-ELM regimes and is found to rise linearly with IaB (a metric for the magnetic configuration strength, the product of current I, minor radius a, and field B that has units of force), and also benefits from tolerance to power. The highest normalized performance using the metric is found in QH and RMP regimes. Focusing on ITER-shaped no-ELM plasmas, Q = 10 at 15 MA scaled global performance is met with some metrics (,), but not others (,), and only thus far at high torque (βN is normalized pressure, H89,98 is the confinement quality by scaling law, and q95 is the safety factor). Though comparable QH and RMP performance is found, the pedestal pressure (pped ≈ 2 pe,ped) is very different. pped in RMP plasmas is relatively low, and the best performance is found with a high core fraction alongside high core rotation, consistent with an ExB shear confinement enhancement. pped of QH plasmas is significantly higher than RMP, and QH performance does not correlate with core rotation. However, the highest QH pped are found with high carbon fraction. While normalized performance of Neg-D plasmas is comparable to QH and RMP plasmas, the absolute confinement of Neg-D is lower, owing to low elongation and low pped achieved thus far. Considering integration with electron cyclotron heating, the operational space for RMP and QH plasmas narrow, while that for EDA-H plasmas open, and the high pped/ regimes (EDA-H and QH) preserve the highest performance. Only the EDA-H, Neg-D, and L-mode scenarios have approached divertor-compatible high separatrix density conditions, with Neg-D preserving the highest performance owing to its compatibility with both high power and density. Comparison to ELMing plasmas highlighted in the literature finds a clearer pedestal correlation to plasma performance, but also reveals that the peak performance of plasmas without ELMs is significantly lower in DIII-D, owing to limits in operational space accessed so far without ELMs. [ABSTRACT FROM AUTHOR]

The effects of a coherent mode (CM) at a frequency of f = 13–16 kHz on the redistribution of divertor particle flux have been studied in HL-2A H-mode plasmas using multiple diagnostics. The experimental results demonstrate that a CM localized at the middle of the pedestal results in outward particle transport, limiting the growth of the pedestal density gradient in the recovery phase of the pedestal, leading to a substantial increase in the localized particle flux striking the outer divertor target, which clearly deviates from the primary peak at the strike point. These observations indicate that the CM might play a crucial role in redistributing the divertor particle flux, which is important for understanding and controlling the divertor heat flux in future fusion-relevant devices. [ABSTRACT FROM AUTHOR]

The ion (Ti) and electron (Te) temperatures at the last closed flux surface (LCFS) during density scanning discharges and in the scrape-off layer (SOL) with a constant plasma density have been measured in the J-TEXT tokamak using a retarding field analyzer and four-tip Langmuir probe, respectively. Both the LCFS Ti and Te are found to reduce with increasing central line-averaged density (). The temperature ratio (τ = Ti/Te) is about 2 for a low (1.5 → 2.5 × 1019 m−3), decays in a moderate (2.5 → 3.5 × 1019 m−3) and tends to unity at a high (3.5 → 4.2 × 1019 m−3). This suggests that the ions and electrons at the LCFS (even in the SOL) are thermally decoupled at low or moderate and tend to full thermal coupling in the high-density plasmas. In addition, both Ti and Te decrease with the radius and Te falls more steeply than Ti due to faster parallel energy loss of electrons in the SOL. In contrast, τ rises with the radius, which indicates that the ions and electrons are much more thermally decoupled in the deeper SOL. [ABSTRACT FROM AUTHOR]

Internal kink (IK) instability is investigated for European demonstration fusion reactor (EU DEMO) plasmas in both negative triangularity (NT) and positive triangularity (PT) configurations. For NT plasmas, the IK becomes more unstable as an ideal conformal wall moves away from the plasma boundary, with the mode growth rate saturating at the wall radial location of about , where a is the plasma minor radius and b the wall radial location. The plasma resistivity destabilizes the IK mode. The effect of sub-sonic toroidal plasma flow is sufficiently weak and can thus be ignored for these EU DEMO equilibria. These results are consistent with those for PT plasmas, albeit with larger mode growth rate in the NT configuration. Both perturbative and self-consistent magneto-hydrodynamic (MHD)-kinetic hybrid calculations predict (partial) stabilization of the IK modes in both NT and PT configurations, with inclusion of various kinetic contributions. Precessional drift motion of trapped fusion-born alphas in EU DEMO produces weak stabilization to the IK mode. Stronger stabilization occurs with the toroidal precession of trapped thermal particles (ions and electrons) and the bounce-transit motion of thermal ions. The stabilization is similar between the NT and PT configurations, due to the similarity of the mode eigenfunction (occupying a nearly circular region in the plasma core) despite the sign difference in the triangularity. The non-perturbative MHD-kinetic hybrid model predicts much less stabilization of the mode than the perturbative model, primarily due to the self-consistent determination of the mode eigenvalue in the former. Generally, no significant difference in the IK mode stability is found between the NT and PT plasmas in EU DEMO. [ABSTRACT FROM AUTHOR]

The study of edge localized mode (ELM) behavior, stationary heat flux and transient heat flux during the ELM crash phase is performed for a China Fusion Engineering Test Reactor (CFETR) 1 GW hybrid mode operation scenario (= 7.2 m, = 6.5 T, = 13.78 MA). Modeling and simulation start with a scenario obtained by multi-code integrated modeling on the one modeling framework for integrated tasks framework. Linear stability and nonlinear simulations of ELM dynamics are carried out using the BOUT++ six-field reduced magnetohydrodynamic module, which show a much smaller ELM energy loss (ΔELM ~ 0.13%) compared to that of a Type-I ELM. Parametric analysis of the weak linear growth rate and small ELM energy loss characteristics shed light on physics corresponding to a grassy ELM regime for CFETR 1 GW hybrid scenario. The transient heat flux on the divertor target during this small ELM phase is investigated using BOUT++. We found that upstream radial transport in the scrape-off-layer (SOL) induced by small ELMs is weak, which keeps it in the drift-dominated region. However, the heat flux width is still broadened to = 4.64 mm by the increase of separatrix temperature during ELM nonlinear evolution. The impact of transient peak heat load and ELM energy fluence on tungsten melting and net erosion rate of divertor target is evaluated for the first-time using physics-based transport that connects the pedestal with the SOL. Energy fluence caused by a single ELM pulse is below the tungsten melting limit, while tungsten erosion would exceed the material requirements. We conclude that external mitigation methods, such as divertor detachment and advanced divertor geometry are likely needed for the steady state operation of CFETR. [ABSTRACT FROM AUTHOR]

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

Current 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 with different physical nature. Besides the known Pfirsch–Schlüter currents and thermoelectric currents, so-called 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 the SOL and private flux region (PFR). In particular, the current flowing to the outer plate in the PFR, opposite to the thermoelectric current, is predicted for the standard single-null configuration and favorable direction of drift. In addition, a pair of currents should flow to and away from the outer plate. In the single-null configuration, they are often masked by a larger thermoelectric current. However, for the connected double null (CDN) case, where the thermoelectric current is strongly reduced due to smaller temperature asymmetry, these currents dominate. The suggested physical model is supported by the results of simulations performed with the SOLPS-ITER transport code. Simulations were done for ASDEX Upgrade (AUG), L- and H-modes, single-null configurations, and for Globus-M H-mode, both disconnected and CDN configurations. Results of the simulations are compared with probe measurements for AUG and Globus-M tokamaks, and reasonable agreement has been found. The role of parallel currents in the formation of the potential maximum/minimum in the vicinity of the X-point for strongly detached regimes is also analyzed. [ABSTRACT FROM AUTHOR]

A first set of divertor gas baffles has recently been installed in the TCV tokamak. In order to explore the physics determining the benefits and limitations of divertor baffling and to guide the design of a possible second generation of baffles, the effect of baffle closure is investigated using the 2D transport code SolEdge2D-EIRENE with realistic wall geometries. The baffle extension is scanned, first imposing the same upstream conditions as in previous SOLPS-ITER studies, then extending the parameter space to access detached plasma conditions. In attached plasma cases, divertor neutral compression is maximised by a Low-Field Side baffle length with an opening between the separatrix and the baffle tip of approximately 5 λq, resulting in an increase in neutral compression by a factor 4 with respect to the unbaffled case. In detached cases this ratio can be improved by up to a factor 25 using higher baffle closures. This difference in behaviour between attached and detached conditions is explained by a model based on the ionisation mean free path of neutral particles recycled from the target. In some conditions, the optimal baffle extension in terms of neutral compression is found to be subject to high levels of intercepted upstream heat flux, which results in a peak heat flux on the baffles comparable to the one impinging on the outer target. The individual roles of the High-Field Side and Low-Field Side baffles are disentangled by means of dedicated simulations, which show a lower global impact of the inner baffle. This study suggests that an outer baffle with a gap of approximately 3 λq, slightly more closed than the one presently installed, could further enhance the neutral compression ratio in cases where the ionisation front is detached. The biggest unknown in these simulations is related to far SOL particle transport, which could result in higher levels of baffle recycling and thus limit baffle performance. [ABSTRACT FROM AUTHOR]

In divertor tokamaks electric currents flowing in the Scrape-Off Layer (SOL) are a frequently observed and well known feature. However, the convective heat flux carried by these currents is often assumed to be small compared to the total parallel heat flux. In this paper it is shown that in ASDEX Upgrade the electric current, which was measured by Langmuir probes at the outer target, can be several times larger than the ion saturation current in low density L- and H-mode discharges. Therefore they have to be taken into account when calculating the target heat flux from the Langmuir probe data. Moreover, a detailed analysis reveals that for low density L- and H-mode discharges the heat flux profile is mainly determined by the current profile. By applying a 1D heat flux transport model it is demonstrated that the electric current can carry a large fraction (> 50 %) of the electron heat flux along the SOL convectively. Thus a Spitzer-Härm approach, where it is assumed that heat is transported only conductively, might not be valid in the investigated regime. [ABSTRACT FROM AUTHOR]

It has been found that the X-divertor (XD) configuration can flare magnetic field lines near the divertor targets and reduce heat flux on the divertor plate (Kotschenreuther 2007 Phys. Plasmas 14 072502). To solve severe heat exhaust problems in the China Fusion Engineering Test Reactor, the XD configuration is proposed. Equilibria calculations indicate that the XD can be realized and is compatible with the long-leg divertor. The position of the extra X-point was optimized to reduce the currents in the central solenoid (CS) coils, which dominates the flattop length. Two internal coils were considered to assist in the optimization of the XD. Different plasma currents were also used to further reduce the CS coil currents. The results indicate that the aforementioned approaches can effectively extend the flattop. [ABSTRACT FROM AUTHOR]

Simultaneous control of the transient heat load induced by the large-amplitude edge-localized modes (ELMs) and the time-averaged heat and particle fluxes to the divertor targets is a critical issue for the steady-state operation of a future tokamak fusion reactor. The combination of divertor detachment and grassy-ELM regime provides a candidate solution to this issue. The strong particle exhaust capability of the grassy-ELM regime greatly facilitates the achievement of steady-state operation of a detached plasma with divertor impurity seeding. Stable complete detachment at inner target and partial detachment at outer target in the grassy-ELM regime have been achieved with seeding of 5% neon (Ne) and 95% D2 mixture since 2018 in the EAST superconducting tokamak with ITER-like tungsten monoblock upper divertor. The peak ion fluxes (Γion) on the upper outer and inner divertor targets were reduced by 55% and 92%, respectively. However, it was accompanied by confinement degradation with 18% reduction in the plasma stored energy WMHD. In contrast, partial detachment on the upper outer divertor target with confinement improvement has been achieved with the power across the separatrix (Psep) around the H-mode threshold power. In addition, at relatively lower q95 (∼5.7) with unfavorable Bt direction, Ne seeding leads to a transition from mixed (large/small) ELM regime into grassy-ELM regime with significantly increased ELM frequency, which extends the accessible parameter range of the grassy-ELM regime towards lower q95. [ABSTRACT FROM AUTHOR]

A new analysis technique for Balmer line spectroscopy that enables recombination rate (particle sink) and ionization rate (particle source) inference in a closed divertor configuration is reported. Bayesian inference is employed to systematically utilize all available information from multiple Balmer lines and constrain parameter ranges by using prior knowledge about plasmas. While a closed-divertor facilitates detachment, neutral plugging typically leads to large spatial variations in plasma parameters. A forward model is developed to take into account non-uniformity in the plasma parameters and applied to test data generated by divertor plasma simulations. It is shown that the forward model robustly provides particle source and sink inference over a wide parameter range. In addition, the precision improves as more Balmer lines are resolved simultaneously. The new analysis technique is also applied to an L-mode ASDEX Upgrade plasma in the high-recycling regime. The inferred quantities and their profiles are consistent with the expectations of a high-recycling divertor plasma. The further insight into the detachment physics will be provided by using this new analysis technique. [ABSTRACT FROM AUTHOR]

The phenomenon of multifaceted asymmetric radiation from the edge (MARFE) is investigated with impurity gas puff from the upper divertor on the Experimental Advanced Superconducting Tokamak (EAST). A typical process in which dense radiation region in the vicinity of the X point (X-point MARFE) further evolves into MARFE on the high field side (wall MARFE) after the plasma makes a transition from H-mode to L-mode confinement is observed in discharges with the impurity gas seeding. The electron density distribution and evolution in the divertor volume are measured by means of spectroscopy. It is observed in the experiments that the final position of MARFE is related to the ion ∇ B drift direction. After the phase difference of n = 1 resonant magnetic perturbation (RMP) change from 90° to 270°, MARFE begins to appear on the high field side, which may be caused by the density pump-out induced by RMP. At the same heating power, the density threshold for MARFE formation is somewhat higher in the Ne seeding discharge than in the Ar seeding discharge. This may be attributed to the fact that the divertor radiation fraction, Prad,div/Prad,main, in the Ne seeding discharge is lower than that in the Ar seeding discharge. In addition, MARFE is successfully suppressed by total radiated energy feedback control in radiative divertor experiments. Therefore, radiation feedback control may play a vital role in avoiding major disruption induced by impurity radiation in radiative divertor experiments. [ABSTRACT FROM AUTHOR]

Neutral beam injection is the main heating system on a variety of fusion devices, and will be the main heating system of ITER. Especially during high-power operation in long pulse devices, it is important that transmission losses and power deposition in the beamline are well quantified. This paper presents the analytical beamlet code (ABC)3D. ABC3D is based on analytical Gaussian beamlets, and includes shadowing effects. The code is able to calculate emission spectra in the beamline, duct transmission losses, and neutral beam power loads on 3D components. Examples of its application are shown that highlight key results in beam emission spectroscopy, beamline design, and geometry verification in experiments. Limitations to the computational approach are discussed on the basis of a comparison with more advanced calculations. [ABSTRACT FROM AUTHOR]

The electron cyclotron resonance heating (ECRH) system of the ASDEX Upgrade tokomak has been upgraded over the last 15 years from a 2MW, 2 s, 140 GHz system to an 8MW, 10 s, dual frequency system (105/140 GHz). The power exceeds the L/H power threshold by at least a factor of two, even for high densities, and roughly equals the installed ion cyclotron range of frequencies power. The power of both wave heating systems together (>10MW in the plasma) is about half of the available neutral beam injection (NBI) power, allowing significant variations of torque input, of the shape of the heating profile and of Qe/Qi, even at high heating power. For applications at a low magnetic field an X3-heating scheme is routinely in use. Such a scenario is now also forseen for ITER to study the first H-modes at one third of the full field. This versatile system allows one to address important issues fundamental to a fusion reactor: H-mode operation with dominant electron heating, accessing low collisionalities in full metal devices (also related to suppression of edge localized modes with resonant magnetic perturbations), influence of Te/Ti and rotational shear on transport, and dependence of impurity accumulation on heating profiles. Experiments on all these subjects have been carried out over the last few years and will be presented in this contribution. The adjustable localized current drive capability of ECRH allows dedicated variations of the shape of the q-profile and the study of their influence on non-inductive tokamak operation (so far at q95 > 5.3). The ultimate goal of these experiments is to use the experimental findings to refine theoretical models such that they allow a reliable design of operational schemes for reactor size devices. In this respect, recent studies comparing a quasi-linear approach (TGLF) with fully non-linear modeling (GENE) of non-inductive high-beta plasmas will be reported. [ABSTRACT FROM AUTHOR]

The power load to the divertor surfaces is a key concern for future devices such as ITER, due to the thermal limits on the material surface. One factor that characterizes the heat flux to the divertor is the fall off length in the scrape-off layer (SOL), which recent empirical scalings have shown could be as small as 1�mm. These predictions are based on a multi-machine scaling of the heat flux width fitted using an expression for the divertor heat flux profile which includes a term for the exponential decay in the SOL and diffusion about the last closed flux surface (LCFS) in the private flux region. This expression has been used to fit a database of inter-ELM H mode profiles at the upper divertor and extract the fall off length, λq, for a range of different plasma parameters in double null plasmas. The midplane separation between the primary and secondary LCFS (δrsep) of the double null plasma used in the study are in the range 2�⩽�δrsep�⩽�7�mm and no correlation is seen between the fall off length and the δrsep. The MAST data shows good agreement with the formula, with the fitted fall off length spanning a range of 5–11�mm in the data base generated. Regression of this data has shown that the fall off length has the strongest dependence on the plasma current (or equivalently, the poloidal magnetic field at the outboard midplane) to the power −0.71. The scaling with the smallest χ2 error utilizes the poloidal magnetic field at the outboard midplane (Bpol,omp) and the power crossing the SOL in the relation with χ2�=�3.46 and R2�=�0.56 as a goodness of fit. The equivalent scaling with plasma current is with χ2�=�3.84 and R2�=�0.55. The moderate goodness of fit suggests that additional plasma parameters are required to accurately reproduce the observed variation in λq. [ABSTRACT FROM AUTHOR]

A new method was applied to indirectly obtain information about the features of the crash of the H-mode edge transport barrier in consequence of an edge localized mode (ELM). The method is based on a combination of fast measurements, without spatial resolution, and relatively slow measurements, with high spatial resolution. The comparison of two different ELM scenarios in the full metal tokamak ASDEX Upgrade—a standard scenario and one with additional nitrogen seeding—revealed a two-fold nature of the ELM crash. In the case with additional nitrogen only a part of the standard crash is observed. This suggests the standard ELM crash consists of two or more consecutive events instead of a single distinct one. Some of these events are observed to be suppressed with changes in plasma parameters. The effect of the impurity seeding on different plasma parameters is documented in detail and compared to measurements conducted in machines with a carbon wall.The radial extent of the phases observed during the ELM crash differs in the kinetic profiles, with one instability extending inside the pedestal top and the other being confined to the pedestal region. This picture can explain the differences in the loss of stored energy and the change in ELM frequency which are observed for the analysed pair of discharges. It also suggests that the ELM crash starts at the pedestal top and only then affects the steep gradient region. [ABSTRACT FROM AUTHOR]

The PDF file provided contains web links to all articles in this volume. [ABSTRACT FROM AUTHOR]

Numerical simulations of L-mode turbulence in the scrape-off layer (SOL) are used to construct power scaling laws for the characteristic decay lengths of the temperature, density and heat flux at the outer mid-plane. Most of the results obtained are in qualitative agreement with the experimental observations despite the known limitation of the model. Quantitative agreement is also obtained for some exponents. In particular, an almost linear inverse dependence of the heat flux decay length with the plasma current is recovered. The relative simplicity of the theoretical model used allows one to gain insight into the mechanisms determining the width of the SOL. [ABSTRACT FROM AUTHOR]

We present a method to obtain reliable edge profiles of the electron temperature by forward modelling of the electron cyclotron radiation transport. While for the core of ASDEX Upgrade plasmas, straightforward analysis of electron cyclotron intensity measurements based on the optically thick plasma approximation is usually justified, reasonable analysis of the steep and optically thin plasma edge needs to consider broadened emission and absorption profiles and radiation transport processes. This is carried out in the framework of integrated data analysis which applies Bayesian probability theory for joint analysis of the electron density and temperature with data of different interdependent and complementary diagnostics. By this means, electron cyclotron radiation intensity delivers highly spatially resolved electron temperature data for the plasma edge. In H-mode, the edge gradient of the temperature can be several times higher than the one of the radiation temperature. Furthermore, we are able to reproduce the 'shine-through' peak--the observation of increased radiation temperatures at frequencies resonant in the optically thin scrape-off layer. This phenomenon is caused by strongly down-shifted radiation of Maxwellian tail electrons located in the H-mode edge region and, therefore, contains valuable information about the electron temperature edge gradient. [ABSTRACT FROM AUTHOR]

Most of the detachment experiments done to date on the tokamak à configuration variable (TCV), both in standard and alternative divertor geometries, focused on L-mode integrated core density ramps. In view of extending these studies to high-power, high-confinement regimes, where impurity seeding will be necessary for detachment, the properties of nitrogen seeded L-mode detachment in TCV are assessed here with the extensive set of edge and divertor diagnostics and similarities and differences with integrated core density () ramp detachment experiments are elucidated. It is found that in high current, reversed field plasmas, detachment at the outer target is achieved with N2-seeding and density ramps, with target heat flux reductions of up to 90%, while the inner target only detaches with seeding. The Scrape-Off Layer radiation fraction reaches values of 60%–80% and in all situations, a stable radiator can form around the X-point. The most striking difference between seeding and density ramp is the behavior of the upstream quantities. During the -ramp, a broadening of the upstream density profile (density ‘shoulder’) occurs, concurrent with the outer target ion flux roll-over, while no such behavior occurs during nitrogen seeded detachment. Separatrix density, electron temperature and pressure also evolve strongly with increasing density, and are largely unaffected by the injection of nitrogen. Comparison of upstream and target pressures reveals that, in all cases, the outer target ion flux reduction coincides with the development of a parallel gradient of the total pressure. Common to all cases is also a reduction of energy confinement time with detachment, although this effect is weak for seeding at relatively high density. Studying the impact of the ∇B-drift direction in both nitrogen seeding and core density ramps reveals that drifts mainly affect the behavior at the inner strike point, highlighting the need to include them in edge transport simulations. [ABSTRACT FROM AUTHOR]

The soft x-ray tomography diagnostic in the stellarator Wendelstein 7-X consists of twenty pinhole cameras, up–down symmetrically arranged in a poloidal, triangular cross-section of the plasma vessel. The x-ray emissivity is measured with 16 bit amplitude resolution at 2 MHz sampling rate along 360 lines-of-sight by silicon photodiode arrays. In the recent operation campaign data acquisition (DAQ) has been working reliable for the conducted plasma pulse lengths <1 min, however the DAQ system are ready for the foreseen 30 min plasma pulse lengths of upcoming campaigns. The bandwidth of the preamplifiers is ≈200 kHz and the sensitive energy range is approximately 1–12 keV. The measurements indicate the up–down symmetric emissivity distribution in the triangular poloidal cross-section. First tomographic reconstructions of different magnetic field configurations are consistent with the theoretically calculated flux surface topology. [ABSTRACT FROM AUTHOR]

An analytic 1D approximation for the divertor broadening S is introduced, depending only on the electron temperature between X-point and target. It is compared to simulations solving the 2D heat diffusion equation, in order to describe the divertor broadening along a field line solely by the ratio of the perpendicular to the parallel diffusivities. By assuming the temperature dependence of these two diffusivities an integral form of S is derived for the area along the separatrix between X-point and target. Integration along the separatrix results in an approximation for S, being in agreement with the 2D simulations. This approximation is furthermore compared to recent studies, which find a power law with negative exponent to describe S in terms of target temperature. This dependence is not reproduced in a pure conductive description, which instead shows a finite S for zero target temperature. This points to other mechanisms changing the shape of the heat flux profile—by additional widening or radiation losses—not included in the presented reduced approximation. [ABSTRACT FROM AUTHOR]

Wendelstein 7-X is a highly optimized stellarator that went into operation in 2015. With a 30 cubic meter volume, a superconducting coil system operating at 2.5 T, and steady-state heating capability of eventually up to 10 MW, it was built to demonstrate the benefits of optimized stellarators at parameters approaching those of a fusion power plant. We report here on the first results with the test divertor installed, during the second operation phase, which was performed in the second half of 2017. Operation with a divertor, and the addition of several new fueling systems, allowed higher density operation in hydrogen as well as helium. The effects that higher density operation had on both divertor operation and global confinement will be described. In particular, at high densities detachment was observed, and the highest fusion triple product for a stellarator was achieved. [ABSTRACT FROM AUTHOR]

An analysis of ASDEX Upgrade plasma is performed to understand the near SOL transport and develop the predictive basis for the electron temperature gradient, All of the unseeded L- and H-mode attached and N seeded H-mode discharges studied are shown that the analyzed ASDEX Upgrade dataset is in the conduction-limited regime, i.e. the parallel transport in the near SOL is dominated by Spitzer–Harm conduction. By studying a H–L back transition, it is shown that the ‘bifurcation’ in the core plasma between H- and L-mode regimes also exists in the perpendicular transport in the near SOL region. Through power balance and the Spitzer–Harm condution, the SOL perpendicular transport can be derived as with For detached plasmas, the SOL upstream electron profile is found to be broader than an equivalent attached plasma under certain conditions. By comparing with global energy confinement, it is found that the discharges with broadened profiles also have degraded confinement, while those with unchanged profiles have similar confinement to attached plasma. The widening of the SOL is also found to coincide with the dropping of upstream temperature. Finally, comparisons of a N seeded H-mode with high pedestal top pressure and an I-mode plasma with an L-mode reference are both found to break the generally observed correlation between the Te SOL decay length and the pedestal top pressure. Thus, the relationship is shown to be non-causal and must instead be due to similar dependences on other plasma parameters. This means that higher global energy confinement is not necessarily imply larger heat flux in the divertor and motivates the search for regimes that optimize both. The implications of these results for SOL transport are discussed. [ABSTRACT FROM AUTHOR]

Magnetic perturbation (MP) fields are currently studied in ASDEX Upgrade and many other tokamaks in terms of edge localized mode control and the implication onto steady state divertor power load and access to detachment. Previous studies at ASDEX Upgrade in low density, attached L-mode (Faitsch et al 2017 Plasma Phys. Control. Fusion59 095006) are combined with high density L- and H-mode studies (Brida et al 2017 Nucl. Fusion57 116006). A consistent interpretation is presented for the steady state power load variation due to the 3D MP. The deviation from an axisymmetric power load is reducing with increasing density, being an effect of the increasing broadening in the divertor region. No deleterious effect on the access to a detached divertor regime is found. Experimental results from ASDEX Upgrade reveal similar power load profiles for plasmas without and a toroidal averaged profile in presence of an external MP. EMC3-Eirene modeling is showing agreement only if screening currents in the plasma as a response to the external magnetic field are added. In plasma scenarios with field penetration and locking of internal modes a fundamentally different power load pattern is observed. The pattern is in agreement with edge ergodization and the extent can be used to determine the magnitude of current needed to describe the internal mode. [ABSTRACT FROM AUTHOR]

Predictions for the operation of tokamak divertors are reliant on edge plasma simulations typically consisting of a fluid plasma code in combination with a Monte-Carlo (MC) code for neutral species. Pilot-PSI is a linear device operating with a cascaded arc plasma source that produces plasmas comparable to those expected during the inter-ELM phase in the ITER divertor (Te ∼ 1 eV, ne ∼ 1020 m−3). In this study, plasma discharges in Pilot-PSI have been modelled using the Soledge2D fluid plasma code (Bufferand et al 2015 Nucl. Fusion55) coupled to the Eirene neutral MC code (Reiter et al 2005 Fusion Sci. Technol.47, 172–186) in order to (a) investigate which phenomena need to be included in the modelling to reproduce experimental trends and (b) provide new insights to the interpretation of experiments. The simulations highlight the key role of ion/molecule elastic collisions in determining the ion flux reaching the target. Recombination is likely to play a role at high molecular background pressure. However, even with the most advanced atomic and molecular model used in this work, Te at the target is overestimated with respect to the measurements using TS and spectroscopy. Te in the simulations appears to saturate at 0.7 eV for a wide range of parameters, while experimentally values of 0.1–0.3 eV are found. As a consequence, in the simulations the volume recombination is underestimated, which is a strong function of Te when it is below 1 eV. Further analysis of simulation results using a two-point formalism shows that inelastic collisions between electrons and neutral background particles remove most of the energy flux, mainly via dissociation of molecules and molecular ions. However this happens mostly in the upstream region of the beam where Te > 1 eV. For Te < 1 eV, there seems to be no significant energy removal mechanism in the simulated cases. The results also indicate that conclusions on the importance of volume processes, e.g. recombination, cannot be solely based on Te or the dominance of certain reaction rate coefficients over others, but rather the complete transport picture, including macroscopic flow, has to be taken into account. In the cases studied here, the plasma is typically advected to the wall too fast for recombination to remove a significant fraction of the particle flux. [ABSTRACT FROM AUTHOR]

At the ASDEX Upgrade tokamak the electron temperature (Te) profile is estimated from electron cyclotron emission (ECE) using radiation transport forward modeling within the integrated data analysis scheme. For the interpretation of ECE measurements in high plasmas, it was necessary to upgrade the forward model with a fully relativistic absorption coefficient including the relativistic Maxwell–Jüttner distribution. This model intrinsically enables the interpretation of ECE measurements affected by the so-called ‘pseudo radial displacement’ or by harmonic overlap. A numerically efficient implementation allows for the analysis of everyday ECE measurements at ASDEX Upgrade. Various ASDEX Upgrade plasma scenarios are discussed highlighting the benefits of the present radiation transport forward modeling for routine analysis. [ABSTRACT FROM AUTHOR]

The impact of the -drift in the cross-field transport and its effect on the density and power scrape-off layer (SOL) width in the limit of low anomalous transport is studied with the fluid code SolEdge2D. In the first part of the work, the simulations are run with an isothermal reduced fluid model. It is found that a -drift dominated regime is reached in all geometries studied (JET-like, ASDEX-like and circular analytic geometries), and that the transition toward this regime comes along with the apparition of supersonic shocks, and a complex parallel equilibrium. The parametric dependencies of the SOL width in this regime are investigated, and the temperature and the poloidal magnetic field are found to be the principal parameters governing the evolution of the SOL width. In the second part of this paper, the impact of additional physics is studied (inclusion of the centrifugal drift, self-consistent variation of temperature and the treatment of the neutral species). The addition of centrifugal drift and neutral species are shown to play a role in the establishment of the parallel equilibrium, impacting the SOL’s width, although the role of the centrifugal drift is limited to a low diffusion level. Finally, the numerical results are compared with the estimate of the Goldston’s heuristic drift based model (HD-model), the starting point of our study, and which has shown good agreement with experimental scaling laws. We find that the particles SOL widths in the -drift dominated regime are at least two times smaller than the estimate of the HD-model. Moreover, in the parametric dependencies proposed by the HD-model, the dependency with is retrieved, but not the one on T. [ABSTRACT FROM AUTHOR]

The scrape-off layer (SOL) power fall-off length, λq, for parameters relevant to ASDEX Upgrade (AUG) L-mode discharges is examined by means of numerical simulations. The simulation data is acquired using synthetic probe data from turbulence simulations, which gives a high temporal resolution on the full density and pressure fields, required for an accurate evaluation of λq due to the strongly intermittent signals in the SOL. Electron conduction is found to dominate the parallel heat flux close to the separatrix, while ion convection dominates in the far SOL. Good agreement is found with the experimental scaling for AUG L-mode Sieglin et al (2016 Plasma Phys. Control. Fusion58 055015), where λq is found to scale almost linearly with the safety factor, q, and to be weakly dependent on the power across the last-closed flux surface (LCFS), P. However, P depends on a wide range of parameters. In this paper we trace this dependence and the resulting fit of λq reveals a scaling proportional to the inverse square root of the electron temperature at the LCFS, , and a linear dependence on q. [ABSTRACT FROM AUTHOR]

It has been previously demonstrated in Li et al (2016 Nucl. Fusion56 126007) that the optimum upper/lower coil phase shift ΔΦopt for alignment of RMP coils for ELM mitigation depends sensitively on q95, and other equilibrium plasma parameters. Therefore, ΔΦopt is expected to vary widely during the current ramp of ITER plasmas, with negative implications for ELM mitigation during this period. A previously derived and numerically benchmarked parametrisation of the coil phase for optimal ELM mitigation on ASDEX Upgrade (Ryan et al 2017 Plasma Phys. Control. Fusion59 024005) is validated against experimental measurements of ΔΦopt, made by observing the changes to the ELM frequency as the coil phase is scanned. It is shown that the parametrisation may predict the optimal coil phase to within 32° of the experimental measurement for n = 2 applied perturbations. It is explained that this agreement is sufficient to ensure that the ELM mitigation is not compromised by poor coil alignment. It is also found that the phase which maximises ELM mitigation is shifted from the phase which maximizes density pump-out, in contrast to theoretical expectations that ELM mitigation and density pump out have the same ΔΦul dependence. A time lag between the ELM frequency response and density response to the RMP is suggested as the cause. The method for numerically deriving the parametrisation is repeated for the ITER coil set, using the baseline scenario as a reference equilibrium, and the parametrisation coefficients given for future use in a feedback coil alignment system. The relative merits of square or sinusoidal toroidal current waveforms for ELM mitigation are briefly discussed. [ABSTRACT FROM AUTHOR]

A comprehensive work of the effects of plasma current and heating schemes on divertor power footprint widths is carried out in the experimental advanced superconducting tokamak (EAST). The divertor power footprint widths, i.e., the scrape-off layer heat flux decay length λq and the heat spreading S, are crucial physical and engineering parameters for fusion reactors. Strong inverse scaling of λq and S with plasma current have been demonstrated for both neutral beam (NB) and lower hybrid wave (LHW) heated L-mode and H-mode plasmas at the inner divertor target. For plasmas heated by the combination of the two kinds of auxiliary heating schemes (NB and LHW), the divertor power widths tend to be larger in plasmas with higher ratio of LHW power. Comparison between experimental heat flux profiles at outer mid-plane (OMP) and divertor target for NB heated and LHW heated L-mode plasmas reveals that the magnetic topology changes induced by LHW may be the main reason to the wider divertor power widths in LHW heated discharges. The effect of heating schemes on divertor peak heat flux has also been investigated, and it is found that LHW heated discharges tend to have a lower divertor peak heat flux compared with NB heated discharges under similar input power. All these findings seem to suggest that plasmas with LHW auxiliary heating scheme are better heat exhaust scenarios for fusion reactors and should be the priorities for the design of next-step fusion reactors like China Fusion Engineering Test Reactor. [ABSTRACT FROM AUTHOR]

The tungsten (W) material in the high heat flux regions of the ITER divertor will be exposed to high fluxes of low-energy particles (e.g. H, D, T, He, Ne and/or N). Combined with long-pulse operations, this implies fluences well in excess of the highest values reached in today’s tokamak experiments. Shaping of the individual monoblock top surface and tilting of the vertical targets for leading-edge protection lead to an increased surface heat flux, and thus increased surface temperature and a reduced margin to remain below the temperature at which recrystallization and grain growth begin. Significant morphology changes are known to occur on W after exposure to high fluences of low-energy particles, be it H or He. An analysis of the formation conditions of these morphology changes is made in relation to the conditions expected at the vertical targets during different phases of operations. It is concluded that both H and He-related effects can occur in ITER. In particular, the case of He-induced nanostructure (also known as ‘fuzz’) is reviewed. Fuzz formation appears possible over a limited region of the outer vertical target, the inner target being generally a net Be deposition area. A simple analysis of the fuzz growth rate including the effect of edge-localized modes (ELMs) and the reduced thermal conductivity of fuzz shows that the fuzz thickness is likely to be limited by the occurrence of annealing during ELM-induced thermal excursions. Not only the morphology, but the material mechanical and thermal properties can be modified by plasma exposure. A review of the existing literature is made, but the existing data are insufficient to conclude quantitatively on the importance and extent of these effects for ITER. As a consequence of the high surface temperatures in ITER, W recrystallization is an important effect to consider, since it leads to a decrease in material strength. An approach is proposed here to develop an operational budget for the W material, i.e. the time the divertor material can be operated at a given temperature before a significant fraction of the material is recrystallized. In general, while it is clear that significant surface damage can occur during ITER operations, the tolerable level of damage in terms of plasma operations currently remains unknown. [ABSTRACT FROM AUTHOR]

This paper reports on experimental observations on TCV with a scan in upper triangularity , including negative triangularity, focusing on the power fall-off length in L-Mode. The upper triangularity is scanned from −0.28 to 0.47. Smaller is measured at the outer divertor target for decreasing together with higher edge temperature leading to increased confinement. This effect is observed for both magnetic drift directions for discharges in deuterium and helium. In helium larger values are observed compared to deuterium. The power fall-off length at the inner divertor target has a non-monotonic behaviour with changing triangularity. The largest values are around . The ratio increases for decreasing for positive triangularity and is approximately constant for negative triangularity. is compared to available scaling laws. Partial agreement is only observed for a scaling law containing a proxy for at ASDEX Upgrade (Sieglin 2016 Plasma Phys. Control. Fusion58 055015). Extending this scaling to TCV and using at suggests that is independent of machine size . Possible explanations for smaller for decreasing is a reduction in turbulence or a direct effect of increasing . [ABSTRACT FROM AUTHOR]

In this paper we present a computational study on the divertor heat load mitigation through impurity injection for the EU DEMO. The study is performed by means of the SOLPS5.1 code. The power crossing the separatrix is considered fixed and corresponding to H-mode operation, whereas the machine operating condition is defined by the outboard mid-plane upstream electron density and the impurity level. The selected impurity for this study is Ar, based on its high radiation efficiency at SOL characteristic temperatures. We consider a conventional vertical target geometry for the EU DEMO and monitor target conditions for different operational points, considering as acceptability criteria the target electron temperature (≤5 eV to provide sufficiently low W sputtering rate) and the peak heat flux (below 5–10 MW m−2 to guarantee safe steady-state cooling conditions). Our simulations suggest that, neglecting the radiated power deposition on the plate, it is possible to satisfy the desired constraints. However, this requires an upstream density of the order of at least 50% of the Greenwald limit and a sufficiently high argon fraction. Furthermore, if the radiated power deposition is taken into account, the peak heat flux on the outer plate could not be reduced below 15 MW m−2 in these simulations. As these simulations do not take into account neutron loading, they strongly indicate that the vertical target divertor solution with a radiative front distributed along the divertor leg has a very marginal operational space in an EU DEMO sized reactor. [ABSTRACT FROM AUTHOR]

A new thermal helium beam diagnostic has been implemented as plasma edge diagnostic at the ASDEX Upgrade (AUG) tokamak. The helium beam is built to measure the electron density ne and temperature Te simultaneously with high spatial and temporal resolution in order to investigate steady-state as well as fast transport processes in the plasma edge region. For the thermal helium beam emission line ratio spectroscopy, neutral helium is locally injected into the plasma by a piezo valve. This enabled the measurement of the line resolved emission intensities of seven He i lines for different plasma scenarios in AUG. The different line ratios can be used together with a collisional-radiative model (CRM) to reconstruct the underlying electron temperature and density. Ratios from the same spin species are used for the electron density reconstruction, whereas spin mixed ratios are sensitive to electron temperature changes. The different line ratios as well as different CRMs are tested for their suitability for diagnostic applications. Furthermore their consistency in calculating identical parameters is validated and the resulting profiles are compared to other available diagnostics at AUG. [ABSTRACT FROM AUTHOR]

A deep understanding of plasma transport at the edge of magnetically confined fusion plasmas is needed for the handling and control of heat loads on the machine first wall. Experimental observations collected on a number of tokamaks over the last three decades taught us that heat flux profiles at the divertor targets of X-point configurations can be parametrized by using two scale lengths for the scrape-off layer (SOL) transport, separately characterizing the main SOL () and the divertor SOL (Sq). In this work we challenge the current interpretation of these two scale lengths as well as their dependence on plasma parameters by studying the effect of divertor geometry modifications on heat exhaust in the Tokamak à Configuration Variable. In particular, a significant broadening of the heat flux profiles at the outer divertor target is diagnosed while increasing the length of the outer divertor leg in lower single null, Ohmic, L-mode discharges. Efforts to reproduce this experimental finding with both diffusive (SolEdge2D-EIRENE) and turbulent (TOKAM3X) modelling tools confirm the validity of a diffusive approach for simulating heat flux profiles in more traditional, short leg, configurations while highlighting the need of a turbulent description for modified, long leg, ones in which strongly asymmetric divertor perpendicular transport develops. [ABSTRACT FROM AUTHOR]