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Surfacing Eroding Thermocouples (SETC) have been used to provide high spatial and temporal resolution heat flux measurement in DIII-D. SETCs were first tested in the lower divertor of DIII-D using the Divertor Material Evaluation System (DiMES), then an array of SETCS was permanently installed in the upper SAS-VW divertor for studying the heat flux mitigation in closed divertor geometry. SETCs proved that inducing divertor detachment with various methods is an effective way to reduce the peak heat flux at the divertor targets. In the investigated discharges the heat flux is reduced to less than 30% under detachment compared to attached conditions. A new method of using the combination of recessed SETC and flush SETC was demonstrated to be able to measure the heat flux from volumetric radiation and the charge exchange neutrals. It was observed that the magnitude of the heat flux from radiation and charge exchange neutrals increased during the process of detachment. While 30% of the total incident heat flux is attributable to the volumetric radiation and the charge exchange neutrals in the fully detached divertor condition with B×∇B drift into the divertor, it could reach more than 60% with B×∇B drift away from the divertor.

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.

The transition from small edge-localized modes (ELMs) to large ELMs has been repetitively observed in minute-scale long-pulse high-confinement mode (H-mode) discharges during the 2017 EAST campaign. The appearance of large ELMs is found to be strongly correlated with the decrease in separatrix density due to the gradual decrease in fuel recycling during long-pulse H-mode operations (LPHOs). By the numerical scanning of separatrix density with a fixed temperature profile, it has been found that the dependence of ELM instability on separatrix density is related to the competition between the ion diamagnetic stabilizing effect and destabilizing effect of pressure gradient and current density in the pedestal region. This sheds light on a comprehensive understanding of the different roles of separatrix density in ELM instability observed during EAST experiments. With a high separatrix density, the ideal ballooning mode can be destabilized near the separatrix, which is thought to help achieve small ELMs in EAST LPHOs. During the 2021 EAST campaign, an experiment of large ELM control was performed through actively changing fuel recycling by moving the strike point location on the lower tungsten divertor target plate. It has been demonstrated that the mitigation of large ELMs is strongly correlated with the significant increase in separatrix density, which is thought to be attributed to a higher ionization source in the scrape-off layer (SOL) region by SOLPS-ITER simulation. The high ionization source in the SOL region is believed to provide a strong fueling effect near the separatrix and thus raise the local density, which is considered an important reason for triggering ballooning instabilities near the separatrix and achieving small ELMs.

To address the needs for a fusion pilot plan design, DIII-D/EAST joint experiments on DIII-D have demonstrated high normalized beta β _N ∼ 4.2, toroidal beta β _T ∼ 3.3% with q _min > 2, q _95 ⩽ 8 sustained for more than six energy confinement times in high poloidal beta regime. The excellent energy confinement quality ( H _98y2 ∼ 1.8) is achieved with an internal transport barrier at high line-averaged Greenwald density fraction f _Gr > 0.9. The trapped gyro-Landau fluid (TGLF) modeling of the transport characteristics shows that the beam-driven rotation does not play an important role in the high confinement quality. The modeling also captures very well several transport features, giving us confidence in using integrated modeling to project these experimental results to future machines. The high-performance phase is terminated by fast-growing modes triggered near the n = 1 ideal-wall kink stability limit. New radio frequency (RF) capabilities for off-axis current drive could remove the residual ohmic current to achieve a fully non-inductive state, and improve the mode–wall coupling to increase the ideal-wall β _N limit, enabling sustainment of the fully non-inductive high performance plasma in stationary conditions.

Integration of transient and steady-state divertor heat fluxes control with a high-performance core is necessary for future fusion reactors. In recent DIII-D high- ${\beta _{\text{p}}}$ experiments, divertor detachment and simultaneous edge localized mode (ELM) suppression are demonstrated while the plasma confinement quality is maintained high in ITER-similar shape. By optimizing the neon injection in high- ${\beta _{\text{p}}}$ scenario with ITER-similar shape, deep detachment and ELM suppression are achieved with a high-performance core ( ${\beta _{\text{N}}}$ ∼ 2.8, ${\beta _{\text{p}}}$ ∼ 2.3) at ${q_{95}}$ ∼ 7.5. Partial divertor detachment and suppression of large ELMs are achieved at ${q_{95}}$ ∼ 6. The stability analyses suggest that with low neon injection, the density pedestal becomes higher and steeper and the ${T_{\text{i}}}$ profile also increases, therefore the increased edge pressure and higher current density destabilize the Peeling-Ballooning mode (PBM), which would lead to a large ELM collapse. With strong neon gas puffing, the significantly reduced pedestal pressure and current density, due to the degraded ${T_{\text{e}}}$ pedestal, lead to the stabilization of PBM and ELMs are suppressed. For both cases, the coupling between the large radius internal transport barrier (ITB) and edge pedestal is the key reason for maintaining high global performance. The formation of large radius ITB compensates for pedestal degradation. Such results could provide an attractive scenario to well control the transient and steady-state heat flux onto the divertor plates while maintaining good plasma performance, which is an important step toward the steady-state operation of future fusion reactors.

Experiments performed during strongly-shaped high-power diverted negative triangularity (NT) experiments in DIII-D achieved detached divertor conditions and a transient-free edge, showcasing the potential for application of NT to a core-edge integrated reactor-like scenario and providing the first characterization of the parametric dependencies for detachment onset. Detached divertor conditions will be required in future devices to mitigate divertor heat fluxes. Access to dissipative divertor conditions was investigated via an increase in upstream density. Detachment onset at the outer strike point was achieved with H-mode level confinement $H_{98-y2}\sim 1$ and reactor-relevant normalized pressures $\beta_\mathrm N \sim 2$ . Confinement degradation was observed with deeper detachment, associated with the loss of an electron temperature pedestal. Differences in geometry, radial transport, impact of cross field drifts are discussed to explain differences in access to detachment in NT discharges. Higher normalized densities, with respect to equivalent discharges in positive triangularity, were necessary to achieve detachment, partially explained by the shorter parallel connection length to the targets. The effect of cross-field particle drifts ( E × B , B $\times \nabla$ B ) on access to detachment was demonstrated by the lower upstream density needed to access detachment with ion B $\times \nabla$ B drift directed outside of the active divertor (Greenwald fraction $f_{\mathrm{Gw}}\sim$ 0.9–1.0 vs $f_{\mathrm{Gw}}\sim$ 1.3). The upstream density at detachment onset was observed to increase linearly with plasma current with ion B $\times \nabla$ B drift into the divertor, consistent with the observed narrowing of the scrape-off layer heat flux width λ _q . Edge fluid simulations capture separatrix densities needed to achieve detachment in NT plasma and their dependence on drift direction. The ability to reproduce detachment dynamics in NT plasma increases the confidence in future design studies for NT divertors.

Experiments in DIII-D demonstrate that the upstream plasma density to detach an un-pumped slot divertor is similar for a V-shaped and a flat-end slot, despite significantly higher neutral pressure in the V-shaped slot and in contrast to SOLPS-ITER predictions. The detachment threshold can be reduced by using in-slot instead of main-chamber gas fuelling or by placing the strike point on the inner slanted slot baffle instead of the slot end, as described by simulations with full drift physics. When increasing the plasma line-averaged density (without extrinsic impurities), the transition to detachment in DIII-D slot divertor is sharp and requires a high value of plasma density with the ion $\overrightarrow {\mathbf{B}} {{ \times }}\nabla B$ drift into the slot, whereas it is smooth and requires a lower value of plasma density with the opposite drift direction, in accord with detachment experiments in the DIII-D open lower divertor. Unique experiments on DIII-D and comparison to advanced simulations expand the scientific understanding of slot-shaped divertors, considered highly desirable for next step fusion devices.

EAST implements the double feedback control experiments of the detachment by using two control systems at the same time to explore the new detachment control way. We hire two PID controllers and two gas puffing valves dividedly to inject impurity gases from different locations, increasing the radiated power of the bulk plasma and reducing the divertor electron temperature (Te,t) simultaneously. This control method is applied in the EAST long-pulse H-mode plasma, the controlled variables of the radiative feedback control is the local radiative intensity and the total radiated power respectively. In the experiments, the longest double control duration has been up to 13.5 s, the divertor heat flux is reduced significantly in the control duration, while the particle flux has an overall drop at the end of the control phase. The chosen impurity species in the control system are Neon (Ne) and Argon (Ar): Ne is injected from the upper divertor to raise the core radiated power, and the Ar is injected from the lower divertor to reduce the Te,t beneath 8 eV. However, the Ar injection is not only decreasing the Te,t, but also generates an unfavorable increment of the core radiated power. This increment makes the bulk radiated power is higher than the target value, the radiative feedback control thus is paused for a long time. This mechanism results in the double feedback control back to the sequential implementations of the two feedback controls, which does not accord with the original expection(the both controls run at the same time), and degrades the control precision. In the future, many mitigating ways will be tested to improve the control effect. We will also develop a dynamic response model of the impurity seeding to simulate the time evolution of the divertor heat load with two kinds of impurities. The modeling results will be used to optimize the double feedback control system to gain a better experimental results.

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.

Divertor detachment without serious core confinement quality loss in DIII-D’s high poloidal β scenario has been combined with impurity-induced ELM mitigation without disruption. Use of Ne previously granted access to a detached, non-ELMing regime that retained high confinement quality due to stimulation of Internal Transport Barrier (ITB) growth, but suffered from on-going core fuel dilution and high disruptivity. Excess Ne accumulation in the core plasma has now been avoided by feeding back core radiated power (Prad) measurements to control Ne seeding, rather than using attachment fraction (Afrac) control with Ne; this also reduces disruptivity. At the same time, N2 seeding is used in a feedback loop with Afrac measurements, which previously posed low disruption risk. In this way, the effect of Ne in the core is managed while avoiding excess seeding, and N2 acts to correct for any excess heat exhaust that might interfere with detachment. The average Ne flow rate was 38% of what was used in pure Ne Afrac control, plus average N2 flow that was 43% of pure N2 Afrac control, even while meeting an even deeper detachment target. Meanwhile, a steep pressure gradient in the core plasma reduces sensitivity to impurity-induced degradation of the pedestal and reasonable confinement quality was maintained despite operational challenges that blocked formation of an ITB in these experiments.

Dedicated experiments in DIII-D find that magnetic shaping and divertor target geometry significantly affect the divertor plasma conditions and divertor detachment process in the small-angle-slot (SAS) divertor. The compact SAS divertor in DIII-D provides a good testbed for understanding the effects of a tightly closed divertor on particle and power dissipation, and for application to core–edge integration solutions. A longer outer leg facilitates the achievement of divertor dissipation in the SAS divertor, while a shorter leg leads to higher electron temperatures near the divertor target plate and requires higher upstream densities to achieve the same level of divertor detachment. In addition, with the ion B × ∇B drift away from the SAS divertor and the outer strike point (OSP) near the outer corner, the target temperature is lower for a particular upstream density than with the OSP on a slanted or flat surface, leading to lower heat flux even when the particle flux remains similar. In contrast, with the ion B × ∇B drift into the SAS divertor, a strike point at the inner slanted surface exhibits a lower upstream density to achieve divertor detachment than a strike point either at the outer corner or the outer slanted target. Experimental results and SOLPS-ITER simulations with full drifts suggest the strong interplay between drift flows and the neutral distribution resulted from target shaping. Furthermore, in-slot gas puffing has been shown to achieve global divertor detachment with an onset density about 10 % lower than that using main-chamber gas puffing when the outer strike point is placed at the inner slanted surface. Corresponding modelling reveals that the local gas puffing enhances the neutral ionization which potentially facilitates the achievement of divertor dissipation. However, such improvement diminishes when the strike point is at the outer corner, which also indicates the geometric dependence on divertor performance in the SAS divertor. Even with different strike point locations, complete divertor detachment with very low particle and heat fluxes at the divertor targets and a high confinement core with normalized energy confinement factor H98 > 1.0 can be simultanesouly achieved with the SAS divertor with ion B × ∇B drift into SAS divertor, demonstrating the benefit of a closed divertor for exploration of core–edge integration.

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 have been conducted in the DIII-D tokamak to explore the in-situ growth of silicon-rich layers as a potential technique for real-time replenishment of surface coatings on plasma-facing components (PFCs) during steady-state long-pulse reactor operation. Silicon (Si) pellets of 1 mm diameter were injected into low- and high-confinement (L-mode and H-mode) plasma discharges with densities ranging from 3.9– $7.5\times10^{19}$ m ^−3 and input powers ranging from 5.5 to 9 MW. The small Si pellets were delivered with the impurity granule injector at frequencies ranging from 4 to 16 Hz corresponding to mass flow rates of 5–19 mg s ^−1 (1– $4.2\times10^{20}$ Si s ^−1 ) at cumulative amounts of up to 34 mg of Si per five-second discharge. Graphite samples were exposed to the scrape-off layer and private flux region plasmas through the divertor material evaluation system to evaluate the Si deposition on the divertor targets. The Si II emission at the sample correlates with silicon injection and suggests net surface Si-deposition in measurable amounts. Post-mortem analysis showed Si-rich coatings containing silicon oxides, of which SiO _2 is the dominant component. No evidence of SiC was found, which is attributed to low divertor surface temperatures. The in-situ and ex-situ analysis found that Si-rich coatings of at least 0.4–1.2 nm thickness have been deposited at 0.4–0.7 nm s ^−1 . The technique is estimated to coat a surface area of at least 0.94 m ^2 on the outer divertor. These results demonstrate the potential of using real-time material injection to form Si-enriched layers on divertor PFCs during reactor operation.

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.

A new, comprehensive set of evidence reveals that Micro-Tearing Modes (MTMs) dominate pedestal electron heat transport in an H-mode experiment in the DIII-D tokamak. The experiment investigates the role of MTMs by scanning pedestal collisionality, a main drive of MTM instability, from 0.43 to 0.84 on the pedestal top. Broadband (150–800 kHz) magnetic and density fluctuations originating from the pedestal gradient region and highly consistent with MTMs are observed, with amplitude increasing during the scan. The higher magnetic fluctuation amplitude correlates with a lower pedestal electron temperature gradient, implying MTMs may regulate the pedestal electron heat transport. The collisionality scan results in profile and transport changes consistent with predicted transport capability of MTMs: (1) experimentally-determined electron heat diffusivity increases ∼40% at the location where the broadband density fluctuations peak; (2) ion heat diffusivity has less increase (

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,

The tungsten erosion process for an H-mode discharge from the DIII-D Metal Rings Campaign is modeled using OEDGE and TRIM.SP. The OEDGE code is employed to calculate tungsten erosion between edge-localized modes (ELMs). Then a newly developed semi-analytical carbon–tungsten mixed material model based on TRIM.SP is used to simulate the intra-ELM tungsten gross erosion profiles. The tungsten erosion is found to be dominated by carbon, with different origin for carbon between ELMs and during ELMs. For inter-ELM, the tungsten is mainly eroded by locally redeposited low charge state carbon, while for intra-ELM, the C6+ originated from the pedestal region is found to dominate the tungsten erosion in the near separatrix region, whereas the locally redeposited low charge state C fluxes lead to a nonnegligible tungsten erosion in the outer SOL region. These results suggest that modeling of W erosion during ELMs needs to include impurity transport from the pedestal to the divertor during an ELM. In addition, for both inter- and intra-ELM simulation, a carbon coverage of 30% on the tungsten surface is needed to reproduce the measured erosion at the divertor target. Keywords: Plasma material interaction, Tungsten, Carbon, Erosion

We report measurements of a+/− 5 mm toroidal variation of the outer strike point radial position using an array of three identical Langmuir probes distributed at 90° intervals around the torus (90°, 180°, 270°). The strike point radial location is determined from the profiles of floating potential (Vf) measured by the three 6 mm diameter domed Langmuir probes as the strike point is swept radially on a horizontal tile surface just outside of the upper small angle slot (SAS1) divertor. Based on the three probe measurements, the strike point variation is consistent with previous error field measurements by Schaffer [1,2] and estimates by Luxon [3] which indicated the strike point error could appear as an n = 1 radial variation of 4.5 mm at the outer mid plane and thus could be effectively described with a three point measurement. The results are also consistent with field line tracing calculations using the MAFOT code [4]. The small angle slot (SAS1) divertor performance is particularly sensitive to a misalignment with the divertor plasma since enhanced neutral confinement and recycling in the slot and distribution of neutrals along the slot surfaces are important for achieving divertor detachment at the lowest possible core plasma separatrix density. These strike point measurements are discussed with regard to the slot divertor alignment.

We report the first detailed experimental verification of the polar deuterium ion impact angle distribution (IAD) on the DIII-D divertor surface in L-mode plasmas using micro-engineered trenches in samples mounted on the DiMES probe. These trenches were fabricated via focused ion beam (FIB) milling of a silicon surface partially coated with aluminum. The sample surfaces were exposed to eight repeat L-mode deuterium discharges (30 s total exposure time). The samples were examined by scanning electron microscopy (SEM), which revealed changes on the trench floor due to material deposition and evidence for shadowing of the incident deuterium ions by the trench walls. The areal distribution of carbon and aluminum deposition was measured by energy-dispersive X-ray spectroscopy (EDS). One-dimensional profiles of this deposition are in agreement with net erosion profiles calculated from a Monte Carlo micro-patterning and roughness (MPR) code for ion sputtering using as input the polar and azimuthal deuterium IADs reported previously (Chrobak et al., Nucl. Fusion 58, 106019 (2018)). The deposition profiles verified the characteristic shape of the polar IADs, which have a broad maximum from 79° to 86°, over the experimental range of 68°–83°, where 0° is the surface normal direction.

Experiments at DIII-D and EAST have demonstrated simultaneous high confinement, divertor detachment, and active control of detachment level, all of which are required for ITER. Comparing detachment control via Teand Jsat, it appears that Teis the most straightforward sensor to use for accessing detachment onset, while Jsatoffers more precise control of degree of detachment. Based on these results, control using nitrogen seeding has so far shown the best ability to follow a target value with the low disruptivity and little to no degradation of performance when an Internal Transport Barrier (ITB) is present, but not all facilities allow its use. Neon seeding also can be paired with feedback control with low impact on core performance as long as there is no disruption, however shots with neon seeding commonly disrupted during these experiments. Argon is effective in EAST, but tends to degrade performance (by ≈10%βp) when detachment is achieved. With ideal conditions and strike point position control, data from a single Langmuir probe are an acceptable input to the control algorithm, but this simple system is easily defeated by strike point displacement comparable to the Teor Jsatscale lengths. The presence of an ITB seems to be critical to retaining core performance in detachment in these parameter ranges, as the pedestal pressure tends to decrease as a result of impurity seeding.

Simultaneous control of transient heat load induced by large-amplitude edge-localized modes (ELMs) and steady-state heat load on divertor targets under metal wall environment is crucial for steady-state operation of future tokamak fusion reactors, such as ITER and the China Fusion Engineering Test Reactor (CFETR). In the recent experiments, sustained partial energy detachment without confinement degradation has been achieved in the Experimental Advanced Superconducting Tokamak (EAST) in high-performance grassy-ELM H-mode with q95 ~ 5.9 by a newly developed detachment feedback control scheme, in which we first used electron temperature (Tet) measured by divertor Langmuir probes to identify the onset of energy detachment, and then the system switched to the feedback control of total radiation power measured by absolute extreme ultraviolet (AXUV) system. Tet around the upper outer strike point was successfully maintained less than 8 eV with seeding of 80% Ne and 20% D2 mixture from upper outer divertor, and the total radiation power was maintained ~1.4 MW, around 52% of injected power. There was no significant decrease of the plasma stored energy and H98,y2 factor (~1) over the entire detachment feedback control process. These experiment results demonstrate good compatibility of the high-performance grassy-ELM regime with radiative divertor. In order to confirm the compatibility in a wider range, stable partial energy detachment in grassy-ELM H-mode with relatively lower q95 (~5.4) was also achieved in EAST through the newly developed integrated-feedback-control technique. The new detachment feedback control without confinement degradation in grassy-ELM H-mode provides a candidate mode for EAST long-pulse operation in the future with well control of ELM-induced transient and steady heat fluxes on the divertor target.

Employing an array of surface eroding thermocouples (SETCs), the heat flux measured at the outer strike point of the DIII-D Small Angle Slot (SAS) divertor was found to be reduced by ~50% when the plasma reached the detachment regime compared to the attached state. Reduction in heat flux of similar magnitude has been found in both toroidal field directions (obtained by ramping up the plasma density) but with different characteristics: With the ion B × ∇B drift direction toward the SAS divertor, the heat flux measured by the SETCs first increased as plasma density increased, which may due to reduction in energy losses in the upstream and hence increase in the power flow into the divertor, and then began to roll over concurrent with a reduction in ion saturation current density on nearby Langmuir probes. In contrast, with the ion B × ∇B drift directed away from the SAS, the heat flux began at a higher value but decreases monotonically with increasing plasma density. This leads to lower heat flux in this BT direction at up to 20% lower line-averaged density. These significant differences in the onset of detachment in opposing BT directions are believed to be largely determined by the impacts of E×B drifts. Because ITER’s divertor must be operated with some degree of plasma detachment to radiate most of the power arriving in the scrape-off-layer, it is crucial to further understand the impact of magnetic drifts on plasma detachment behavior in a tight divertor geometry.

DIII-D L-mode experiments with local boron powder injection for real-time wall conditioning have been interpreted for the first time with the 3D plasma edge transport Monte Carlo code EMC3-EIRENE. Local B sourcing in plasma scenarios with upstream densities 1.5⋅1019m−3 and 2.2 MW heating results in a non-axisymmetric B distribution in the scrape-off layer (SOL) and on the divertor. The SOL frictional flows at high plasma density cause a strong inboard drag of injected impurities (≈90%), while lower background plasma densities tend to result in a more uniform distribution. The thermal forces prevent B deposition in the near SOL while the frictional force causes B fluxes to cover the divertor plasma-facing components in a region 7–10 cm beyond the strike line. Radiative dissipation occurs for B influxes above 1⋅1020s−1 and causes a moderate, non-axisymmetric reduction of the far SOL divertor heat fluxes. A comparison of top and midplane B injection shows no substantial difference in inboard vs. outboard asymmetries of the B distribution. On the other hand, erosion or recycling at the strike line may distribute the boron more uniformly in the SOL.

We assess the toroidal magnetic field B _t asymmetry in DIII-D due to a misalignment of the toroidal field coils with respect to the poloidal magnetic field coils and vacuum vessel. The peak-to-peak variation of the divertor strike point (SP) radius is measured to be 1 cm, with an n = 1 toroidal pattern. We use the centre of a narrow carbon deposition band on tungsten-coated divertor tiles just inside the outer strike point (OSP) as a proxy for the divertor SP location. The band occurred in a series of reverse B _t discharges with the OSP positioned on the divertor inserts due to strong E × B drift transport of C from the inner to the outer SP through the private flux region. The variation in band radius (and hence the magnetic SP) is a (4.89 ± 0.31) mm shift toward (310 ± 4)° toroidal direction. These measurements agree well with previous measurements of the 3D magnetic field distribution (Luxon 2003 Nucl. Fusion 43 1813), simulations performed by the mafot field line integration code, and recent Langmuir probe measurements in the small-angle-slot (SAS) divertor (Watkins et al 2019 Nucl. Mater. Energy 18 46). Comparison of these measurements in the SAS divertor also indicates that there is the possibility of a tilt (in conjunction with the shift) of the B _t coil field of (0.04 ± 0.07)° towards the toroidal angle of (215 ± 25)°. Previous measurements suggested a field misalignment of (4.6 ± 0.3) mm in the 270° toroidal direction, and a tilt of (0.06 ± 0.02)° toward the 114° toroidal direction, which is similar to the results reported here. These studies will be important for better understanding the radial variation of the toroidal strike line in DIII-D, for designing the new generation of SAS divertor, and for developing an understanding of the impact of error fields on tokamaks with tightly baffled slot divertors.

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 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.

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.

The planar Hall effect (PHE) of magnetite ([fe.sub.3][O.sub.4]) films is measured. The magnitude and stability of the PHE in the remanent state has made it a viable candidate for magnetic random access memory applications.