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Experimental data from NSTX and DIII-D discharges with the snowflake (SF) divertor configurations are analyzed toward the development of the X-point radiator (XPR) concept. The XPR divertor regime was recently realized in standard divertor configurations in several tokamaks The SF divertor configuration, with an additional poloidal field null nearby the main X-point, could provide additional benefits for the XPR: a higher flux expansion inside the separatrix and an extended private flux region. This may lead to lower temperatures and higher neutral and electron densities, which are thought to be essential for XPR stability, initiation, and impurity containment. In this work, 4 MW NBI-heated H-mode NSTX discharges and 3–5 MW NBI-heated H-mode DIII-D discharges with SF-minus and SF-plus divertors, with the ion B×∇B drift toward the lower divertor, with and without D2 and CD4 seeding, were analyzed. Many experimental XPR features were found, including good or slightly degraded H-mode confinement, significant ELM size reduction, nearly complete divertor power detachment and a significant divertor radiated power loss. However, evidence of the XPR extending into the confined region was inconclusive in the NSTX tokamak, while in DIII-D, a number of discharges demonstrated a stable MARFE-like structure inside the separatrix over a wide operating space. The present analysis supports the SF divertor as a good candidate for further XPR scenario development in DIII-D and NSTX-U.

The MAST-U tokamak has recently undergone upgrades to investigate any detachment advantages of the “Super-X Double-Null” magnetic configuration, a concept featuring up-down symmetry with two magnetic nulls and elongated divertor legs with strong wall baffling. This study presents SOLPS-ITER simulations of MAST-U Super-X Double Null discharges, in H-mode conditions, compared to initial 2D imaging measurements of divertor detachment characteristics. In the simulations, the Super-X divertor targets achieve detachment at a substantially lower upstream separatrix density compared to a shorter leg, conventional configuration (CD). As the density increases further in the Super-X configuration, the cold detached front moves upstream towards the X-point. Its speed in the poloidal plane, calculated as the change in front location divided by the corresponding increase in upstream plasma density, decreases by a factor of ∼ 3–4 as the front transitions from the baffled divertor chamber to the main plasma chamber, reaching a speed comparable to that observed in the CD configuration. In a chosen experimental Super-X density ramp discharge, the ionization front, identified using D2 Fulcher emission as a proxy, is already displaced from the target at the start of the divertor fueling ramp, whereas modeling of the same initial phase shows the ionization front still close to the target. As divertor fueling increases, both experiment and modeling show the ionization front moving further upstream towards the divertor throat. However, the associated variation in upstream density is ∼ 2 times larger in the modeling compared to the experiment, resulting in a broader detachment window. Comprehensive modeling efforts are underway to address these discrepancies, with several inaccurate modeling assumptions identified as potential causes of the deviation.

DIII-D experiments have achieved promising core–edge integrated plasma scenarios which combine a high-temperature low-collisionality pedestal (pedestal top temperature T _e _,ped > 0.8 keV and collisionality ν * _ped < 1) with a partially detached divertor by leveraging the benefits of a low-density-gradient pedestal in a closed divertor. It is found that with a closed divertor and high heating power, strong gas puffing to achieve detachment moves the peak density gradient outward with respect to the maximum gradient of electron temperature and reduces the density gradient at the pedestal region, which correlates with shallow pedestal fuelling due to the closed divertor geometry. In high-current plasmas in particular, the pedestal top density is found to change little with gas puffing while the separatrix, density increases to allow access for divertor detachment. The separation between density and temperature pedestals results in a high- η _e well above the electron-temperature-gradient stability threshold. Electron turbulence is found to be enhanced in the pedestal and correlated with high η _e resulting from the pedestal shift. The pedestal is wider than the EPED scaling. A revised empirical width scaling is derived based on the combination of EPED scaling with η _e and highlights the important role of additional turbulence on the pedestal structure. The wide temperature pedestal facilitates the achievement of a high-temperature, low-collisionality pedestal and high global performance. Simultaneously, the outward shift of the density pedestal facilitates access to detached divertor conditions with low temperature and heat flux towards the target plate. This approach may be promising for closing the core–edge integration gap for future fusion reactors, which may have a weak-gradient density pedestal due to the highly opaque boundary plasmas.

Recent results from MAST Upgrade are presented, emphasising understanding the capabilities of this new device and deepening understanding of key physics issues for the operation of ITER and the design of future fusion power plants. The impact of MHD instabilities on fast ion confinement have been studied, including the first observation of fast ion losses correlated with Compressional and Global Alfvén Eigenmodes. High-performance plasma scenarios have been developed by tailoring the early plasma current ramp phase to avoid internal reconnection events, resulting in a more monotonic q profile with low central shear. The impact of m / n = 3/2, 2/1 and 1/1 modes on thermal plasma confinement and rotation profiles has been quantified, and scenarios optimised to avoid them have transiently reached values of normalised beta approaching 4.2. In pedestal and ELM physics, a maximum pedestal top temperature of ∼350 eV has been achieved, exceeding the value achieved on MAST at similar heating power. Mitigation of type-I ELMs with n = 1 RMPs has been observed. Studies of plasma exhaust have concentrated on comparing conventional and Super-X divertor configurations, while X-point target, X-divertor and snowflake configurations have been developed and studied in parallel. In L-mode discharges, the separatrix density required to detach the outer divertors is approximately a factor 2 lower in the Super-X than the conventional configuration, in agreement with simulations. Detailed analysis of spectroscopy data from studies of the Super-X configuration reveal the importance of including plasma-molecule interactions and D _2 Fulcher band emission to properly quantify the rates of ionisation, plasma-molecule interactions and volumetric recombination processes governing divertor detachment. In H-mode with conventional and Super-X configurations, the outer divertors are attached in the former and detached in the latter with no impact on core or pedestal confinement.

Divertor heat flux and its correlation with pedestal parameters within various small edge localized mode (ELM) regimes, including high beta poloidal, type-II and ELMs with negative triangularity H-modes were investigated in DIII-D. The parallel energy fluences of type-II and high beta poloidal small ELM regimes fall below the linear scaling with pedestal electron pressure for type-I ELMs put forward in Eich et al 2017 ( Nucl. Mater. Energy 12 84–90). The negative triangularity of H-mode ELMs follow the Eich scaling for type-I ELMs. The parallel heat flux and total heat loads to the divertor were determined using high-time resolution infrared thermography, while pedestal parameters were obtained through self-consistent kinetic equilibrium reconstructions. Linear regressions for the type-II and high beta poloidal regimes demonstrate that an equivalent 7.5 MA small ELM scenario in ITER would fall below the ~5 ${\mathrm{MJ}}\,{\mathrm m^{-2}}$ leading edge melting limit for tungsten (Gunn et al 2017 Nucl. Fusion 57 046025). Utilizing fast thermography, the scrape-off layer power fall-off length for both inter-ELM and intra-ELM was determined and compared to the Eich scaling with poloidal magnetic field in Eich et al (ASDEX Upgrade Team and JET EFDA Contributors 2013 Nucl. Fusion 53 093031). Except for the high beta poloidal scenario, all the small ELM regimes during both inter- and intra-ELM periods had power fall-off lengths $\left(\lambda_q\right)$ larger then would be expected from the $B^{-1}_{\mathrm{pol}, \mathrm{MP}}$ scaling associated with type-I ELMs, signifying their potential in managing heat loads and offering a solution for core–edge integration.

Comparisons of profiles of the electron temperature (Te), density (ne) and pressure (pe) measured with Divertor Thomson Scattering in DIII-D low-confinement mode discharges to predictions from the edge fluid code EDGE2D-EIRENE [1] show that the models implemented in EDGE2D-EIRENE predict the measurements within their collective uncertainties if the Te at the separatrix (Te,sep) is 10 eV, or higher. The simulations do not predict, however, the peaked Te and ne profiles measured adjacent to the target plate when Te,sep, is below 10 eV, i.e., for the plasma downstream from the region of ionization of deuterium atoms. Inclusion of cross-field drifts and a fivefold reduction of radial transport cannot reconcile the discrepancy between the measurements and predictions.

A staged divertor program is currently under discussion to advance DIII-D research on core-edge integration. One phase could address optimization of power and particle exhaust, and supporting modeling of several slot divertor options is underway, including variations in wall baffling, slot depth and divertor leg length. This paper focuses on the role of slot depth to achieve highly dissipative (detached) divertor conditions, in both BT directions. For ion B×∇B into the divertor and PSOL= 4 MW, SOLPS-ITER finds that increasing the slot depth from 18 to 50 cm reduces the upstream separatrix electron density needed to detach by 15%, due to increased divertor radiation. A dedicated run of the EIRENE neutral transport code, in which neutrals are launched from the outer target and followed until ionization, finds that neutral leakage is strongly reduced in the deep slot compared to the shallow slot, explaining the increased divertor radiation and, thus, lower detachment density threshold. Reversing the BT direction cools and densifies the plasma in the slot, such that both slot options are detached at all simulated densities. As for the opposite BT direction, the deep slot has lower target temperature compared to the shallow slot, as a result of lower neutral leakage. Increasing the depth of a slot divertor is, therefore, beneficial to achieve highly dissipative divertor conditions for both field directions. Additional modeling will build on these results to evaluate whether an increased slot depth can also improve trapping of low-Z radiating impurities.

The integration of a high-performance core and a dissipative divertor, or the so-called ‘core–edge integration,’ has been widely identified as a critical gap in the design of future fusion reactors. In this letter, we report, for the first time, direct experimental evidence of electron turbulence at the DIII-D H-mode pedestal that correlates with the broadening of the pedestal and thus facilitates core–edge integration. In agreement with gyrokinetic simulations, this electron turbulence is enhanced by high η _e ( η _e = L _n / L _T _e , where L _n is the density scale length and L _T _e is the electron temperature scale length), which is due to a strong shift between the density and temperature pedestal profiles associated with a closed divertor. The modeled turbulence drives significant heat transport with a lower pressure gradient that may broaden the pedestal to a greater degree than the empirical and theoretically predicted pedestal width scalings. Such a wide pedestal, coupled with a closed divertor, enables us to achieve a good core–edge scenario that integrates a high-temperature low-collisionality pedestal (pedestal top temperature T _e,ped > 0.8 keV and a pedestal top collisionality ν * _ped < 1) under detached divertor conditions. This paves a new path toward solving the core–edge integration issue in future fusion reactors.

Experiments performed in DIII-D demonstrate that higher plasma current and heating power combined with impurity seeding facilitate the achievement of divertor detachment with a higher pedestal pressure and higher plasma performance in H-mode plasmas with a baffled closed divertor compared with an open divertor. Dedicated experiments were carried out to study the impact of power, plasma current and impurity seeding on divertor detachment with ion $B \times \nabla B$ directed into the divertor favorable for the L–H transition. With a factor of three variation in heating power and with only D _2 puffing, no significant difference in the separatrix density at detachment onset was found. The higher heating power leads to higher impurity concentration and wider scrape-off layer (SOL) width, and reduces the detachment onset density to one similar to that in lower-power plasmas. Higher current requires higher pedestal and line-averaged densities to achieve divertor detachment; however, the increase in separatrix density at increasing plasma current is found to be less pronounced. Initial calculations found that both power scan and plasma current scan datasets are qualitatively consistent with theory after considering the change in impurity concentration and heat flux width. This also motivates the future extensive study of transport and divertor impurity behavior in order to have a quantitative comparison between experiment and theory. Compared with an open divertor, a closed divertor facilitates detachment onset at ∼40% lower line-averaged plasma density. Additional N _2 seeding facilitates the achievement of detachment at a lower separatrix density and thus a higher pedestal temperature, which is beneficial for advanced tokamak scenarios. Higher heating power requires a higher N _2 puffing rate to achieve the same degree of detachment, while a higher N _2 puffing rate leads to lower detachment onset line-averaged density, both of which agree with theory. In contrast to the narrower pedestal in an open divertor approaching detachment, the pedestal density width in a closed divertor increases with density. The density gradient increases with line-averaged density at higher plasma current, but remains nearly unchanged at lower plasma current. In particular, compared with discharges with low power, at high heating power the pedestal density gradient is much weaker, while the SOL density is significantly higher and wider. At the same plasma current, both pedestal pressure gradient and temperature gradient decrease linearly with the line-averaged density but remain similar across different heating powers. Even with different plasma current and heating power, the normalized pressure gradient remains identical. As a result, achievement of divertor detachment with a higher pedestal pressure and higher plasma performance is shown in a closed divertor, which is important for improving core–edge integration as one of the critical issues for future tokamak fusion reactors.

Actively enhancing plasma radiated power has been used to reduce divertor heat flux in high power, high-performance double null divertor (DND) and near-DND plasmas in DIII-D, while at the same time maintaining acceptable energy confinement. One form of this approach that we focus on here is the radiating mantle. The effectiveness of the radiating mantle was found to depend strongly on the location of electron cyclotron (EC) heating deposition and on the effect that seed impurities have on triggering inimical MHD activity inside the plasma. In addition, we present experimental support for the predictions made by the ELITE boundary stability code (Snyder et al., 2002, [1]) for ways to improve confinement and fueling in these high-performance plasmas. Representative AT operating conditions were H98 ≈ 1.4–1.7, βN ≈ 3–4, and q95 ≈ 4.7–6.5. JEL classification: Divertor Materials (D0500), Impurities (10100), Plasma Properties (P0600), Keywords: DIII-D, Edge pedestal, Energy confinement, Gas injection and fueling impurity, Power deposition

The contributions of deuterium molecular emission to the total deuterium radiation was assessed in DIII-D ohmically-confined plasmas in high-recycling divertor conditions. Radial profiles of the deuterium Ly-α line intensity across the low-field side divertor leg were obtained with the recently installed divertor Survey Poor Resolution, Extended Spectrometer [1]. A high-resolution spectrometer was used to measure the poloidal profiles of the deuterium Balmer-α and the deuterium Fulcher-α band intensity in the visible wavelength range. The scrape-off layer plasma and neutral distributions were simulated using the edge fluid EDGE2D-EIRENE [2], and the numerical solutions constrained utilizing Thomson scattering and Langmuir probe measurements at the low-field side midplane and the divertor target plate. The studies show that for these conditions molecular emission plays a negligible role in the total radiative power balance of the low-field side divertor, but molecular processes are important when evaluating deuterium Balmer-α line intensity for code-experiment validation. Keywords: Deuterium atomic and molecular emission, Lyman-Werner and Fulcher band, DIII-D, Divertor edge fluid simulations, EDGE2D-EIRENE, EIRENE

Radiative divertor experiments and 2D fluid simulations show strong impact of cross-field drifts on the low field side (LFS) divertor target heat flux and volumetric radiation profiles when evolving to detached conditions in DIII-D high confinement mode (H-mode) plasmas with forward and reversed toroidal field configurations. In both field configurations, the peak heat flux is reduced by about a factor of 2 in detachment by D2-injection and by factor of 3 – 4 in detachment by N2-injection. Operating with the B × ∇B-drift towards the X-point (fwd. BT), the LFS divertor radiation front is observed to shift step-like from the target to near to the X-point at the onset of detachment. In contrast, operating with the B × ∇B-drift away from the X-point (rev. BT), the radiation front is observed to remain closer to the target plate and to be radially shifted towards the far SOL. These phenomena occur with detachment induced both with N2- and D2-injection. The step-like detachment onset is consistent with recently published theory of the role of poloidal E × B-drift in the private flux region in driving highly non-linear detachment onset in fwd. BT [22]. Keywords: Divertor, Detachment, Power exhaust, Drifts, UEDGE

Experiments at DIII-D use different divertor geometries to measure the effect of divertor closure on detachment onset. We compare matched H-mode discharges in the closed (upper) divertor and the open (lower) divertor and use as a detachment onset metric the rollover in peak ion flux to the target, Jsat, as a function of upstream pedestal top density, ne,ped. Measurements of electron temperature in the lower divertor confirm that this metric captures the transition from attached to detached divertor for this data set. Experiments at two different neutral beam injection powers demonstrate an ≈ 20–40% decrease in pedestal density at detachment onset in the closed case compared to the open. High-resolution edge density and temperature measurements and power balance analysis allow us to determine the difference in upstream separatrix density ne,sep at detachment onset, which we find to be ≈ 15–25% lower in the closed divertor case. The effect of closure on detachment seen in separatrix density is hence only about 60–75% the effect seen in pedestal density, with the remainder due to differences in ne,sep/ne,ped. Future devices are expected to have a scrape off layer that is opaque to neutrals, and hence will not have the same pedestal fueling to affect ne,sep/ne,ped; this suggests that the effect of closure on detachment onset seen in future devices will be less than that seen in current experiments, and closer to the magnitude of the effect we see here in the separatrix for a similar degree of divertor closure. Keywords: Tokamak, Divertor, Detachment, Fusion

Local neutral pressure measurements in the closed small angle slot (SAS) divertor [1,2] on DIII-D show a large increase when the divertor plasma shifts from high recycling into detachment. In-tile pressure gauges were installed to measure the pressure in the near- and far-SOL regions to examine the predicted high neutral pressures and compression. Cross-field drift effects lead to ∼ 10 × higher peak neutral pressure in detachment with the ion ∇B drift toward the divertor compared to out of the divertor, with similar pressures in attached conditions. Drifts also play a role in the neutral distribution in the slot while attached but become less pronounced in detachment once gradient drives reduce. Variation in the outer strike point location found higher neutral pressure and detachment at modestly lower main-plasma density with the strike point positioned away from the designed operation point. Reducing the fraction of the SOL width allowed into the slot increases neutral leakage into the main chamber and increases the main-plasma density required for detachment. Preliminary modeling with the SOLPS code without drifts over-predicts the neutral pressure in detachment by a factor of 2 with the strike point in the designed operation point while more significantly over-predicts the neutral slot compression; experiments show a broader distribution of neutrals through the slot. These measurements are used to help understand detachment and validate divertor design metrics. Keywords: DIII-D, Divertor, Neutral pressure, Detachment,

Feedback control of radiated power from the lower divertor Prad, div, L has been implemented in the DIII-D Plasma Control System (PCS). A realtime sensor for Prad, div, L has been constructed from 12 foil bolometer channels which agrees with standard post-shot analysis to within 20%. Results with the 12-channel sensor are compared to initial proof-of-concept tests with a single channel as a proxy for Prad, div, L, showing that the upgraded sensor is necessary to overcome limitations of the proxy channel strategy in DIII-D. Using N2 seeding under feedback control, Prad, div, L has been increased by up to 150% above unseeded levels, and a radiated power fraction frad of 80% has been demonstrated, although feedback controlled gas flow is steadier at frad = 55%. Spatial coverage is broad enough to enable Prad control during the strike point sweeps which are commonly used to generate pseudo-2D divertor Thomson measurements in DIII-D divertor experiments. Use of this control reveals challenges that may affect next step devices, which will require actively controlled extrinsic impurity seeding in order to manage heat loads. When operating at high frad, changes in pedestal Te (caused by ELMs in these experiments but could come from other disturbances) resulted in large perturbations to Prad which were destabilizing to the feedback controller. Keywords: Plasma control, Radiation, Tokamak, Fusion, Divertor

A refined version of the Fundamenksi-Moulton 'free-streaming' model (FSM) for the dynamics of divertor density, particle flux, and heat flux during edge localized modes (ELMs) is presented. This model depends only on inter-ELM pedestal and divertor conditions and, crucially, incorporates particle recycling: a FSM with recycling model, FSRM. The effective particle recycling coefficient, Reff, is the only empirical fitting parameter in the FSRM. The predictions of the FSRM are systematically tested against a DIII-D database of ELM ion and energy fluence measurements and are shown to be consistent with the model across a wide range of pedestal and divertor conditions using a constant value of 0.96 for Reff . Predictions for W sputtering during ELMs are developed based on the FSRM. It is concluded that energetic free-streaming D+ ions and C6+ impurities are the dominant contributors to the intra-ELM gross erosion of W in the DIII-D divertor, i.e., recycling ions and impurities have relatively little impact on the total W sputtering rate. These calculations are also shown to be consistent with spectroscopic measurements of W gross erosion for three different pedestal conditions after incorporating the strong electron density dependence of the WI 400.8 nm ionizations/photon (S/XB) coefficient. Keywords: Tungsten sputtering, Tungsten Erosion, WI spectroscopy, Edge Localized Modes, Recycling

The radial width of the exhaust heat flux flowing in the SOL of DIII-D is found to expand at high input power and plasma density, consistent with MHD ballooning stability limits. At low heating power, ~3 MW, the SOL width remains constant and consistent with established empirical scaling dependent only on the midplane poloidal field. At high heating power, ~ 13 MW a higher separatrix density, and resulting higher separatrix pressure is required for divertor detachment. The separatrix pressure gradient at the separatrix continues to increase with density until it begins to saturate at levels ~50% above the calculated ideal MHD ballooning limit. Examination of the separate contributions to the pressure gradient from electrons and ions reveals the ion pressure gradient to saturate more strongly than the electron pressure gradient. Potential analysis issues leading to the measured pressure gradient exceeding the ballooning limit are discussed. At high density, particularly for detached divertor plasmas, the SOL width for temperature and density expand modestly, ~30–50%. The divertor plasma density profile in detachment also reflects this trend, expanding in the radial direction a factor of 2–3. Despite the SOL width expansion at the highest power and density no degradation of the pedestal and resulting core confinement is observed with the additional density at high power. These results imply a more favorable scaling for divertor heat flux control in future reactor-scale tokamaks than predicted by existing empirical scaling.

The OEDGE code is used to model tungsten erosion and transport for experiments with toroidal rings of high-Z metal tiles in the DIII-D tokamak. Such modeling is needed for both experimental and diagnostic design to have estimates of the expected core and edge tungsten density and to understand the various factors contributing to the uncertainties in these calculations. OEDGE simulations are performed using the planned experimental magnetic geometries and plasma conditions typical of both L-mode and inter-ELM H-mode discharges in DIII-D. OEDGE plasma reconstruction based on specific representative discharges for similar geometries is used to determine the plasma conditions applied to tungsten plasma impurity simulations. A new model for tungsten erosion in OEDGE was developed which imports charge-state resolved carbon impurity fluxes and impact energies from a separate OEDGE run which models the carbon production, transport and deposition for the same plasma conditions as the tungsten simulations. These values are then used to calculate the gross tungsten physical sputtering due to carbon plasma impurities which is then added to any sputtering by deuterium ions; tungsten self-sputtering is also included. The code results are found to be dependent on the following factors: divertor geometry and closure, the choice of cross-field anomalous transport coefficients, divertor plasma conditions (affecting both tungsten source strength and transport), the choice of tungsten atomic physics data used in the model (in particular ionization rate for W-atoms), and the model of the carbon flux and energy used for calculating the tungsten source due to sputtering. Core tungsten density is found to be of order 1015m−3 (excluding effects of any core transport barrier and with significant variability depending on the other factors mentioned) with density decaying into the scrape off layer. For the typical core density in the plasma conditions examined of 2 to 4 ×1019m−3, this represents a concentration on the order of 5 × 10−5.

Joint DIII-D/EAST experiments in the radio-frequency (RF) heated H-mode scheme with comparison to that of neutral beam (NB) heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL) width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broader SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. The joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH) H-mode plasmas. Keywords: SOL width, Radio-frequency heating, Neutral beam heating, Low collisionality

Simulations using the fully kinetic code XGCa were undertaken to explore the impact of kinetic effects on scrape-off layer (SOL) physics in DIII-D H-mode plasmas. XGCa is a total-f, gyrokinetic code which self-consistently calculates the axisymmetric electrostatic potential and plasma dynamics, and includes modules for Monte Carlo neutral transport. Fluid simulations are normally used to simulate the SOL, due to its high collisionality. However, depending on plasma conditions, a number of discrepancies have been observed between experiment and leading SOL fluid codes (e.g. SOLPS), including underestimating outer target temperatures, radial electric field in the SOL, parallel ion SOL flows at the low field side, and impurity radiation. Many of these discrepancies may be linked to the fluid treatment, and might be resolved by including kinetic effects in SOL simulations.The XGCa simulation of the DIII-D tokamak in a nominally sheath-limited regime show many noteworthy features in the SOL. The density and ion temperature are higher at the low-field side, indicative of ion orbit loss. The SOL ion Mach flows are at experimentally relevant levels (Mi ∼ 0.5), with similar shapes and poloidal variation as observed in various tokamaks. Surprisingly, the ion Mach flows close to the sheath edge remain subsonic, in contrast to the typical fluid Bohm criterion requiring ion flows to be above sonic at the sheath edge. Related to this are the presence of elevated sheath potentials, eΔΦ/Te∼3−4, over most of the SOL, with regions in the near-SOL close to the separatrix having eΔΦ/Te > 4. These two results at the sheath edge are a consequence of non-Maxwellian features in the ions and electrons there.

The impurity-seeded detached divertor is essential for heat exhaust in ITER and other reactor-relevant devices. Dedicated experiments with injection of N2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N2, Ne and Ar all promote divertor detachment, greatly reducing heat flux near the strike point. The upstream plasma density at the onset of detachment decreases with increasing impurity-puffing flow rates. For all injected impurity species, the confinement and pedestal pressure are correlated with the impurity content and the ratio of separatrix loss power to the l-H transition threshold power. As the divertor plasma approaches detachment, the high-Z impurity seeding tends to degrade the core confinement owing to the increased core radiation. In particular, Ar injection with up to 50% of the injected power radiating in the core cools the pedestal and core plasmas, thus significantly degrading the confinement. As for Ne seeding, medium confinement with H98∼0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N2 seeded plasmas, radiation is predominately confined in the boundary plasma, which leads to less effect on the confinement and pedestal. In the case of strong N2 gas puffing, the confinement recovers during the detachment, from ∼20% reduction at the onset of the detachment to greater than unity comparable to that before the seeding. The core and pedestal temperatures feature a reduction of 30% from the initial attached phase and remain nearly constant during the detachment phase. The improvement in confinement appears to arise from the increase in pedestal and core density despite the temperature reduction.

We present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ∼100nm thick were applied to ideal (smooth) and realistic (rough) surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non-spectroscopic measurements. The gross Al erosion yield was estimated from film thickness change measurements of small area samples, and was found to be ∼40–70% of the expected erosion yield based on theoretical physical sputtering yields after including sputtering by a 1–3% carbon impurity. The multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration patterns, were found to be influenced by the surface roughness and/or porosity. A time-dependent model of material migration accounting for deposit accumulation in hidden areas was developed to reproduce the measurements in these experiments and determine a redeposition probability distribution function for sputtered atoms. Keywords: Erosion, Deposition, Migration, Roughness, Aluminum, Mixed materials

Simulations using the multi-fluid code UEDGE indicates that, in low confinement (L-mode) plasmas in DIII-D, poloidal projection of the ionization driven flows dominate poloidal particle flows in the divertor near the divertor plates, whereas E ×B drift flows dominate the radial particle flows. In contrast, in high confinement (H-mode) conditions E ×B drift flows dominate both poloidal and radial particle flows in the divertor in the vicinity of the strong gradient region near the separatrix. UEDGE indicates that the toroidal C2+ flow velocities in the divertor plasmas are mainly entrained within 30% to the background deuterium flow in both L- and H-mode plasmas in the plasma region where the CIII 465nm emission is measured. Therefore, UEDGE indicates that the Doppler Coherence Imaging Spectroscopy (CIS), measuring the toroidal velocity of the C2+ ions, can provide insight to the deuterium flows in the divertor. Parallel-to-B velocity dominates the toroidal divertor flow; direct drift impact being less than 1%. Toroidal divertor flow is predicted to reverse when the magnetic field is reversed. This is explained by the parallel-B flow towards the nearest divertor plate corresponding to opposite toroidal directions in opposite toroidal field configurations.

Significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND) hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q⊥P) ∝ [PSOL x IP] 0.92 for PSOL= 8−19MW and IP= 1.0–1.4MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-D plasmas may be problematical at high power and H98 (≥ 1.5) due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q⊥P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot”) but also that heating near the slot opening is a significant source for impurity contamination of the core. PSI-22 keywords: DIII-D, Divertor geometry, Gas injection and fueling, Impurity sources, Power deposition

Detailed measurements of the ne, Te, and Ti profiles in the vicinity of the separatrix of ELMing H-mode discharges have been used to examine plasma stability at the extreme edge of the plasma and assess stability dependent models of the heat flux width. The results are strongly contrary to the critical gradient model, which posits that a ballooning instability determines a gradient scale length related to the heat flux width. The results of this analysis are not sensitive to the choice of location to evaluate stability. Significantly, it is also found that the results are completely consistent with the heuristic drift model for the heat flux width. Here the edge pressure gradient scales with plasma density and is proportional to the pressure gradient inferred from the equilibrium in accordance with the predictions of that theory.

As High-Z materials will likely be used as plasma-facing components (PFCs) in future fusion devices, the erosion of high-Z materials is a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion of Mo and W is found to be strongly suppressed by local injection of methane and deuterium gases. The 13C deposition resulting from local 13CH4 injection also provides information on radial transport due to E ×B drifts and cross field diffusion. Finally, D2 gas puffing is found to cause local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.