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Plasma-facing materials in future large-scale fusion reactors must be designed to withstand high heat fluxes from extreme off-normal events such as edge localized modes and unmitigated plasma disruptions. The erosion rates of possible tungsten-alternative materials are tested under high heat flux conditions at the DIII-D National Fusion Facility. High-purity β-3C CVD silicon carbide was exposed alongside MAX phase ceramic Ti3SiC2 to both L- and H-mode plasma discharges in the DIII-D divertor. Samples survived average heat fluxes ranging from 2–10 MW/m2 over 16 s. A new micro-trench erosion measurement technique was successfully implemented and measured Ti3SiC2 and SiC erosion rates of 0–9 nm/s and 27–73 nm/s, respectively. Additionally, average ion impact angle estimates for an incident B-field angle of ∼1.5° from surface parallel were made using micro-trench impact patterns. Measurements ranged from θ = 24o–34o with respect to Bt and ϕ = 51.5o–55o below the surface normal.

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 2016 DIII-D DiMES tungsten (W) leading edge experiment in support of ITER studied heat loads in helium (He) plasmas with ELMs and compound ELMs (C-ELMs). The regime was close to the threshold for L-mode to H-mode transitions. In shots with ECH, C-ELMs produced in H-L back-transitions dominated the transient particle exhaust. Their duration (10–20 ms) was much longer than regular ELMs, and their heat loads higher by 2–3 times. The C-ELMS reduced the plasma density significantly, and some triggered (automated) gas puffing. This regime, with high particle and heat exhaust, may have relevance for He plasmas in ITER's start-up phase. Regular ELMs occurred during neutral beam (NB) powered shots. This paper describes new analyses on heat loads during and between C-ELMs in ECH-powered shot 166843. The results are generally consistent with a geometric solution for the parallel heat load at the plasma edge striking the leading edge. The power in the C-ELMs is sufficient to cause a temperature rise of ∼

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 D 2 -injection and by factor of 3 – 4 in detachment by N 2 -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 N 2 - and D 2 -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].

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

Divertor detachment offers a promising solution to the challenge of plasma-wall interactions for steady-state operation of fusion reactors. Here, we demonstrate the excellent compatibility of actively controlled full divertor detachment with a high-performance (βN ~ 3, H98 ~ 1.5) core plasma, using high-βp (poloidal beta, βp > 2) scenario characterized by a sustained core internal transport barrier (ITB) and a modest edge transport barrier (ETB) in DIII-D tokamak. The high-βp high-confinement scenario facilitates divertor detachment which, in turn, promotes the development of an even stronger ITB at large radius with a weaker ETB. This self-organized synergy between ITB and ETB, leads to a net gain in energy confinement, in contrast to the net confinement loss caused by divertor detachment in standard H-modes. These results show the potential of integrating excellent core plasma performance with an efficient divertor solution, an essential step towards steady-state operation of reactor-grade plasmas., Plasma fusion devices like tokamaks are important for energy generation but there are many challenges for their steady state operation. Here, the authors show that full divertor detachment is compatible with high-confinement high-poloidal-beta core plasmas and this prevents the damage to the divertor target plates and the first wall.

Relativistic electron (RE) beams at high current density (low safety factor, q ( a )) yet very low free-electron density accessed with D-2 secondary injection in the DIII-D and JET tokamak are found to exhibit large-scale MHD instabilities that benignly terminate the RE beam. In JET, this technique has enabled termination of MA-level RE currents without measurable first-wall heating. This scenario thus offers an unexpected alternate pathway to achieve RE mitigation without collisional dissipation. Benign termination is explained by two synergistic effects. First, during the MHD-driven RE loss events both experiment and MHD orbit-loss modeling supports a significant increase in the wetted area of the RE loss. Second, as previously identified at JET and DIII-D, the fast kink loss timescale precludes RE beam regeneration and the resulting dangerous conversion of magnetic to RE kinetic energy. During the termination, the RE kinetic energy is lost to the wall, but the current fully transfers to the cold bulk thus enabling benign Ohmic dissipation of the magnetic energy on longer timescales via a conventional current quench. Hydrogenic (D-2) secondary injection is found to be the only injected species that enables access to the benign termination. D-2 injection: (1) facilitates access to low q ( a ) in existing devices (via reduced collisionality & resistivity), (2) minimizes the RE avalanche by 'purging' the high-Z atoms from the RE beam, (3) drives recombination of the background plasma, reducing the density and Alfven time, thus accelerating the MHD growth. This phenomenon is found to be accessible when crossing the low q ( a ) stability boundary with rising current, falling toroidal field, or contracting minor radius-the latter being the expected scenario for vertically unstable RE beams in ITER. While unexpected, this path scales favorably to fusion-grade tokamaks and offers a novel RE mitigation scenario in principle accessible with the day-one disruption mitigation system of ITER.

UEDGE simulations highlight the role of cross-field drifts on the onset of detached conditions, and new calibrated divertor vacuum ultra violet (VUV) spectroscopy is used to challenge the predictions of radiative constituents in these simulations. UEDGE simulations for DIII-D H-mode plasmas with the open divertor with the ion ∇B-drift towards the X-point show a bifurcated onset of the low field side (LFS) divertor detachment, consistent with experimentally observed step-like detachment onset (Jaervinen A.E. et al 2018 Phys. Rev. Lett. 121 075001). The divertor plasma in the simulations exhibits hysteresis in upstream separatrix density between attached and detached solution branches. Reducing the drift magnitude by a factor of 3 eliminates the step-like detachment onset in the simulations, confirming the strong role of drifts in the bifurcated detachment onset. When measured local plasma densities and temperatures are within proximity of predicted values in the simulations, there is no shortfall of the local emission of the dominant resonant radiating lines. However, the simulations systematically predict a factor of two lower total integrated radiated power than measured by the bolometer with the difference lost through radial heat flow out of the computational domain. Even though there is no shortfall in the emission of the dominant lines, a shortfall of total radiated power can be caused by underpredicted spatial extent of the radiation front, indicating a potential upstream or divertor transport physics origin for the radiation shortfall, or shortfall of radiated power in the spectrum between the dominant lines. In addition to the underpredicted spatial extent, in detached conditions, the simulations overpredict the peak radiation and dominant carbon lines near the X-point, which can be alleviated by manually increasing divertor diffusivity in the simulations, highlighting the ad hoc cross-field transport as one of the key limitations of the predictive capability of these divertor fluid codes.

In ITER, first mirrors for the Upper Wide Angle Viewing System (UWAVS) are expected to experience contamination with beryllium and tungsten. Contamination levels of 10 nm can degrade mirror performance. In the UWAVS, a first mirror cleaning system will use radio-frequency (RF) gas discharge. A mockup in representative geometry was developed. The tests focused on power delivery, hardware survivability, cleaning uniformity, and removal of contaminants. Using RF discharges without introducing surface damage was confirmed. Sputtering was investigated for 10–20 nm thick Al-coatings used as proxies for Be due to health and safety issues. Helium was chosen for many tests due to its effectiveness and suitability for ITER operations. The most effective cleaning was obtained in Helium at 5 Pa, 36–43 MHz, and 50–150 W. The sputtering conditions were determined through ion energy measurements. Ion energies of 100–150 eV and currents of 0.5–0.8 A/m2 were found to be optimum. The measured cleaning rate in He for Al- and Al-oxide was 1.5–3 nm/hour. Aluminum was used as a proxy material for Be. The setup was operated for 500 h. No significant chamber contamination was identified.

Edge localized modes (ELMs) are triggered using deuterium pellets injected into plasmas with ITER-relevant low collisionality pedestals, and the resulting peak ELM energy fluence is reduced by approximately 25%–50% relative to natural ELMs destabilized at similar pedestal pressures. Cryogenically frozen deuterium pellets are injected from the low-field side of the DIII-D tokamak at frequencies lower than the natural ELM frequency, and heat flux is measured by infrared cameras. Ideal MHD pedestal stability calculations show that without pellet injection, these low collisionality pedestals were limited by their current density (peeling-limited) rather than their pressure gradient (ballooning-limited). ELM triggering success correlates strongly with pellet mass, consistent with the theory that a large pressure perturbation is required to trigger an ELM in low collisionality discharges that are far from the ballooning stability boundary. For sufficiently large pellets, both instantaneous and time-integrated ELM energy deposition measured by infrared cameras is reduced with respect to naturally occurring ELMs at the inner strike point, which is the position where it is largest for natural ELMs. Energy fluence at the outer strike point is less effected. Cameras observing both heat flux and D-alpha emission often find significant toroidally asymmetric striations in the outboard far scrape-off layer resulting from ELMs that are triggered by pellets. Toroidal asymmetries at the inner strike point are similar between natural and pellet-triggered ELMs, suggesting that the reduction in peak heat flux and total fluence at that location is robust for the conditions reported here.

An optimized pedestal regime called the super-H (SH) mode is leveraged to couple a fusion relevant core plasma with a high density scrape-off layer appropriate for realistic reactor power exhaust solutions. Recent DIII-D experiments have expanded the operating space of the SH regime using advanced control algorithms and investigated optimization of impurity seeding, deuterium gas puffing, and 3D magnetic perturbations. Simultaneous real-time control of the pedestal density and radiated power with in-vessel coils and nitrogen seeding enable optimal coupled divertor and pedestal conditions. Four case studies are analysed with varied levels of radiated power in the divertor volume ranging from 0 (no seeding) to 8.5 MW radiated from carbon and nitrogen emission. Plasmas with a 4.5 MW radiated power target establish a radiative mantle, leading to divertor temperatures of ∼16 eV while maintaining SH-mode, and with only marginal impact on the pedestal and core performance. Increased levels of N2 seeding with a 7.5 MW radiated power target facilitate detachment onset and divertor temperatures 2.5{\tau }_{\text{E}}$?> > 2.5 τ E . Finally, a 8.5 MW radiated power target leads to partial detachment, which is so far associated with the loss of access to SH-mode pedestal conditions.

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.

One of the diagnostic systems being provided by the US is the Upper Wide Angle Viewing System (UWAVS) which provides real-time, simultaneous visible and infrared (IR) images of the ITER divertor regions via optical systems located in five upper ports. UWAVS is designed in three main sections: in-vessel, interspace and port cell assemblies. Each assembly utilizes multiple steering and relay mirrors to direct the in-vessel light out of the tokamak to the port cell camera sensors. The primary design challenge of the in-vessel assembly is maximizing performance of the overall system while surviving the severe electromagnetic and nuclear ITER environment. A first mirror material study was conducted and determined that single crystal molybdenum was the best choice for the first two mirrors in the optical train. A fail open, bellows actuated shutter with cross pivot flexure design was determined to be the most reliable mechanism to protect the first mirror. A geometrically representative glow discharge mirror cleaning system is being designed and will be tested to maximize cleaning effectiveness while minimizing optical degradation of the molybdenum mirrors adjacent to the plasma. The shutter and first mirror assemblies were packaged and designed for replacement via remote handling methods to minimize radiation waste and cost associated with eventual component replacement. The preliminary optical and structural design provides a robust and reliable system while maximizing the field of view. The R&D efforts, the technical challenges and issues, and the design and analysis results are presented.

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

The key remaining physics design issue for the ITER tungsten (W) divertor is the question of monoblock (MB) front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM) or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA) and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets. Keywords: ITER, Tungsten, Divertor, Shaping, Melting, MEMOS

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 13 C deposition resulting from local 13 CH 4 injection also provides information on radial transport due to E × B drifts and cross field diffusion. Finally, D 2 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.

Experiments have been conducted on DIII-D investigating high repetition rate injection of non-fuel pellets as a tool for pacing Edge Localized Modes (ELMs) and mitigating their transient divertor heat loads. Effective ELM pacing was obtained with injection of Li granules in different H-mode scenarios, at frequencies 3–5 times larger than the natural ELM frequency, with subsequent reduction of strike-point heat flux (Bortolon et al., Nucl. Fus., 56, 056008, 2016). However, in scenarios with high pedestal density (∼6 × 1019 m−3), the magnitude of granule triggered ELMs shows a broad distribution, in terms of stored energy loss and peak heat flux, challenging the effectiveness of ELM mitigation. Furthermore, transient heat-flux deposition correlated with granule injections was observed far from the strike-points. Field line tracing suggest this phenomenon to be consistent with particle loss into the mid-plane far scrape-off layer, at toroidal location of the granule injection.

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3–5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower $n_{e}$ , and ELM energy and divertor peak heat flux reduction, especially prominent in radiative $D_{2}$ -seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative $D_{2}$ -seeded SF divertor at $P_{\text {SOL}}=3-4$ MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected $P_{\text {SOL}} \simeq 9$ MW case. The radiative SF divertor with carbon impurity provides a wider $n_{e}$ operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar $q_{\text {peak}}$ reduction factors (see standard divertor).

A new small angle slot (SAS) divertor concept has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Initial tests on DIII-D find a strong interplay between such anticipated 'SAS' effects and cross-field drifts, favouring operation with the ion B × ∇B drift away from the X-point, as currently employed for advanced tokamaks. This offers the following key improvements relative to DIII-D's open lower divertor or partially-closed upper divertor: (i) SAS allows for transition to low temperature moderately detached divertor conditions with T e ≲ 10 eV at very low main plasma densities, lower than are usually attainable at all in DIII-D high confinement (H-mode) plasmas as used in these tests; (ii) Pedestal performance and core confinement are significantly improved with SAS. The final confinement collapse associated with the onset of X-point MARFE (multifaceted asymmetric radiation from the edge) following deep detachment occurs at significantly higher pedestal densities, thus widening the window of H-mode operation compatible with a dissipative divertor. For operation with the ion B × ∇B drift toward the X-point, the divertor plasma transitions to a bifurcative detached state at much higher densities, similar to other divertor configurations in DIII-D. These results highlight the strong interplay between divertor closure and drifts, and point to an interesting divertor optimization path to explore that offers potential for future fusion reactors.

We observe the formation of a high-pressure staircase pedestal ($\ensuremath{\approx}16--20\text{ }\text{ }\mathrm{kPa}$) in the DIII-D tokamak when large amplitude edge localized modes are suppressed using resonant magnetic perturbations. The staircase pedestal is characterized by a flattening of the density and temperature profiles in midpedestal creating a two-step staircase pedestal structure correlated with the appearance of midpedestal broadband fluctuations. The pedestal oscillates between the staircase and single-step structure every 40--60 ms, correlated with oscillations in the heat and particle flux to the divertor. Gyrokinetic analysis using the cgyro code shows that when the heat and particle flux to the divertor decreases, the pedestal broadens and the $E\ifmmode\times\else\texttimes\fi{}B$ shear at the midpedestal decreases, triggering a transport bifurcation from the kinetic ballooning mode (KBM) to trapped electron mode (TEM) limited transport that flattens the density and temperature profiles at midpedestal and results in the formation of the staircase pedestal. As the heat flux to the divertor increases, the pedestal narrows and the $E\ifmmode\times\else\texttimes\fi{}B$ shear at the midpedestal increases, triggering a back transition from TEM to KBM limited transport. The pedestal pressure increases during the staircase phase, indicating that enhanced midpedestal turbulence can be beneficial for confinement.

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

The first rapid tokamak discharge shutdown using dispersive core payload deposition with shell pellets has been achieved in the DIII-D tokamak. Shell pellets are being investigated as a possible new path toward achieving tokamak disruption mitigation with both low conducted wall heat loads and slow current quench. Conventional disruption mitigation injects radiating impurities into the outer edge of the tokamak plasma, which tends to result in poor impurity assimilation and creates a strong edge cooling and outward heat flow, thus requiring undesirable high-$Z$ impurities to achieve low conducted heat loads. The shell pellet technique aims to produce a hollow temperature profile by using a thin, low-ablation shell surrounding a dispersive payload, giving a greatly increased impurity ablation (and radiation) rate when the payload is released in the plasma core. This principle was demonstrated successfully using 3.6 mm outer diameter, $40\text{ }\text{ }\ensuremath{\mu}\mathrm{m}$ thickness diamond shells holding boron powder. The pellets caused rapid ($l10\text{ }\text{ }\mathrm{ms}$) discharge shutdown with low conducted divertor heat fluence ($\ensuremath{\sim}0.1\text{ }\text{ }\mathrm{MJ}/{\mathrm{m}}^{2}$). Confirmation of massive release of the boron powder payload into the plasma core was obtained spectroscopically. Some evidence for the formation of a hollow temperature profile during the shutdown was observed. These first results open a new avenue for disruption mitigation research, hopefully enabling development of highly effective methods of avoiding disruption wall damage in future reactor-scale tokamaks.

Integration of divertor detachment with a high-performance (βN ∼ 3, βp > 2, H98 ∼ 1.5) core plasma has been demonstrated in DIII-D high-βp (poloidal beta) plasmas associated with a sustained core internal transport barrier (ITB) and an H-mode edge transport barrier (ETB). Such good core-edge integration has been achieved for both neon and nitrogen seeding, for both favorable and unfavorable B-field directions, independently from the impurity puffing locations, though these variations play important roles on divertor characteristics. Compared to the standard H-mode plasmas, the high-βp plasma exhibits a much wider window of detachment compatible with high confinement core. Fully detached divertor plasmas with low plasma temperature (Te

One of the challenges of the snowflake divertor (SFD) configuration is finding a reliable means of reconstructing the magnetic field geometry in the divertor. Since the SFD (and other advanced divertors) have multiple field nulls, there is a large region with shallow flux gradients that is difficult to resolve accurately using external diagnostics. In this work we present a technique that uses heat flux measured by the infrared television (IRTV) camera to improve SFD reconstruction. This is relevant for purposes of control, since the SFD is topologically unstable and requires active feedback on the shape [E. Kolemen, et. al., Nucl. Fusion, 58, 6 (2018)], and analysis, since reconstructions provided by other algorithms such as EFIT [L. Lao, et. al., Nucl. Fusion, 25, 11 (1985)] can mis-characterize the shape and even the snowflake type (plus or minus). The technique identifies the spatial position of the two x-points located in the SFD based on characteristics of the heat flux such as the strike point location and power distribution. The inferred x-point positions are then used as a constraint in fitting new equilibria using the TokSys suite of software. This procedure is applied to ~ 800 DIII-D SFD timeslices and reduces the summed strike point errors from an average 9.4 cm to 0.9 cm. The newly-created x-point constrained equilibria are compared to kinetic reconstructions and an average 16% reduction in the edge current is observed. This is correlated via a simple linear relationship to the shape constraints. Other changes in the pedestal structure are observed, but more work must be done to incorporate the IRTV constraint directly into kinetic solvers to obtain integrated solutions.

Coherence Imaging Spectroscopy (CIS) has emerged as a powerful tool for investigating complex ion phenomena in the boundary of magnetically confined plasma devices. The combination of Fourier-transform interferometry and high-resolution fast-framing cameras has made it possible to make sensitive velocity measurements that are also spatially resolved. However, this sensitivity makes the diagnostic vulnerable to environmental effects including thermal drifts, vibration, and magnetic fields that can influence the velocity measurement. Additionally, the ability to provide an absolute calibration for these geometries can be impacted by differences in the light-collection geometry between the plasma and reference light source, spectral impurities, and the presence of thin-films on in-vessel optics. This paper discusses the mitigation of these effects and demonstration that environmental effects result in less than 0.5 km/s error on the DIII-D CIS systems. A diagnostic comparison is used to demonstrate agreement between CIS and traditional spectroscopy once tomographic artifacts are accounted for.