A series of L-mode plasma discharges was performed in the DIII-D tokamak to assess the impact of outer strike point (OSP) position and toroidal magnetic field direction on erosion and core contamination potential of the recently-installed, tungsten-coated Small Angle Slot (SAS-VW) divertor. In one discharge, in-slot emission spectroscopy measured an < 48 % increase in the W gross erosion rate when the OSP was moved 3 cm outwards, away from the V-shaped vertex of the slot divertor. However, the effective W yield (erosion rate divided by the incident D flux) was, overall, insensitive to changes in OSP location. Consistently low estimates of the effective W yield based on measurements taken a few cm outwards from the vertex suggest potentially significant C surface contamination. No W emission signal was detected when orienting the toroidal magnetic field such that the ion Bx del B drift direction is pointed away from the X-point. However, measurements of W content in the plasma core for both toroidal magnetic field directions suggest the presence of additional, unmeasured sources of erosion. The difference in the measured core W density with OSP position is much greater than the difference in the measured erosion rates, which may suggest that the leakage of eroded impurities out of the divertor is governed primarily through the parallel ion temperature gradient and friction forces.

A novel impurity transport model that approximates SOL turbulence as a fluctuating poloidal electric field is shown to be an acceptable replacement for the traditional approach of assigning an arbitrary radial diffusion coefficient to the impurity ions. The model is implemented in the DIVIMP impurity transport code and applied to an L-Mode tungsten divertor experiment on DIII-D. The poloidal electric field is represented as fluctuating between +/- 1000 V m-1 based on previous measurements. The resulting intermittent vr = E theta x BT transport causes ions to transport both into the core as well as into the far-SOL. Simultaneous agreement with estimates of the W density just inside the separatrix as well as in the far-SOL is obtained (nW similar to 1014 m-3 and nW similar to 1012 m-3, respectively). Prompt re-deposition of the W ions was necessary to obtain agreement (fredep similar to 99%). We conclude that simulating impurity transport using a physics-based approximation for turbulence in the SOL, versus arbitrarily assigning diffusion coefficients, may enable better reactor scale predictions of core impurity contamination.

Fundamental mechanisms governing the erosion and prompt re-deposition of tungsten impurities in tokamak divertors are identified and analyzed to inform the lifetime of tungsten plasma-facing components in ITER and other future devices. Various experiments conducted at DIII-D to benchmark predictive models are presented, leveraging the DiMES removable sample exposure probe capability and the Metal Rings Campaign, in which toroidally symmetric rows of tungsten-coated tiles were installed in the DIII-D divertor. In tokamak divertors, the width of the electric sheath is of the order of the main ion Larmor radius, and a vast majority of sputtered tungsten impurities are typically ionized within the sheath. Therefore, W prompt redeposition is mainly governed by the ratio of the characteristic ionization mean-free path of neutral tungsten to the width of the sheath. In-situ monitoring of the prompt redeposition of tungsten impurities in divertors is demonstrated via the use of WII/WI line ratios and the ionizations/photon (S/XB) method in L-mode discharges. Even with this relatively limited set of emission measurements, net erosion measurements were found to be a consistent upper bound to an analytic scaling based on the ratio of the W ionization length, λ iz , and the width of the magnetic sheath rather than the ratio of λ iz and the W + gyro-radius. In the far-scrape-off layer (SOL) of the ITER divertor, however, it is calculated that the measurement of photon emissions associated with the ionization of tungsten impurities up to W 5 + may be required. Finally, W deposition patterns on DiMES collector probes, interpreted via DIVIMP-WallDYN modelling, reveal the key roles of progressive W erosion/re-deposition staps and E × B drifts in regulating long-range high-Z material migration.

Silicon carbide (SiC) is a promising material for use in a fusion reactor due to its low hydrogenic diffusivity, high temperature strength and resilience under neutron irradiation [1,2]. To assess SiC as a main wall material in DIII-D, simulations with TRIM.SP and DIVIMP are performed on a well-diagnosed L-mode discharge. The effective charge, Zeff, across the separatrix is used as a figure of merit in comparing SiC to the current graphite walls. It is found that SiC is expected to reduce Zeff, potentially by as much as ∼50 %. It is discussed how SiC may be expected to “self-condition” and create wall conditions similar to siliconization, further lowering Zeff due to efficient oxygen gettering. The potential benefits are reviewed and a path towards SiC walls in DIII-D is presented.

Experimental results from the 2022 tungsten (W)-coated Small Angle Slot (SAS-VW) divertor campaign at DIII-D coupled with interpretive 3DLIM modelling show opposing trends for core impurity content when compared to impurity deposition on far-Scrape Off Layer (SOL) Collector Probes (CPs) with increasing main ion density. SAS-VW is a closed, W-coated divertor designed to more easily facilitate divertor detachment while reducing impurity leakage. An experiment performed a series of upper-single-null L-mode discharges in each toroidal magnetic field (BT) direction, with increasing main ion density (line-averaged density = 3.15-4.35e19 m- 3) that approaches and slightly exceeds the divertor detachment threshold. The results indicate: a) increased radial W transport with decreasing peak TLP e,t ; and b) negligible change in W content in the far-SOL at the outer mid-plane with the onset of divertor detachment.Preliminary W deposition measurements using double-sided, graphite CPs inserted at the Midplane Materials Evaluation System (MiMES) reveal a 75% decrease over the density scan when operating in the unfavorable (ion Bx backward difference B out of the divertor) BT direction. In contrast, soft X-ray (SXR) radiation from the same discharges is used as a proxy for W core contamination, showing core W content that increases by 77% with increasing line-averaged density. Similar L-mode discharges conducted in the favorable BT direction result in significantly less deposition on CPs.Using an interpretive modeling workflow following Zamperini 2022 [1] for assessing the transport of W sputtered from the SAS-VW divertor, the analysis suggests that W migration to the main chamber surfaces during the campaign may also contribute to far-SOL deposition.

A novel multicode workflow to interpret collector probe deposition patterns in DIII-D has been developed. The components of the workflow consist of a detailed computer-aided-design file of the vessel wall and the scrape-off-layer (SOL) codes MAFOT, OSM, DIVIMP, and 3DLIM. A special-purpose toolkit enables passing the output of these codes among each other to provide a full-SOL picture of impurity transport. A demonstration of the workflow is described to support evidence of near-SOL tungsten parallel accumulation during trace W impurity experiments on DIII-D. Iteration between simulated deposition patterns in 3DLIM and DIVIMP predicts a region of elevated W density near the separatrix about halfway between the outboard midplane and the top of the plasma. This workflow will be used to better interpret collector probe experiments on DIII-D.

Near-separatrix impurity accumulation between the crown and the outer midplane of tokamaks is a common feature in results from codes such as SOLPS-ITER and DIVIMP; however, experimental evidence of accumulation has only recently been obtained and is reported here. The codes find that the poloidal distribution of impurity ions in the scrape-off layer (SOL) depends primarily on toroidal field ( B T )-dependent parallel flow patterns of the background plasma and the parallel ion temperature gradient (∇ ‖ T ion ) force. Experimentally, Mach probes used in L-mode plasmas with favorable (for H-mode access) B T measure fast ( M ∼ 0.3–0.5) inner-target-directed (ITD) background plasma flows at the crown of single-null discharges. This study reports a set of DIVIMP simulations for two similar H-mode discharges from the DIII-D W metal rings campaign differing primarily in B T -direction to assess the effect that fast ITD flows have on the distribution of W ions in the SOL. It is found that for imposed ITD flows of M = 0.3, W ions that otherwise accumulate due to the ∇ ‖ T ion -force are largely flushed out. It is also found that doubling the radial diffusion coefficient from 0.3 to 0.6 m 2 s −1 prevents accumulation due to rapid cross-field transport into the far-SOL, where background plasma flows drain W ions to the divertors. Far-SOL W distributions from DIVIMP are then used to specify input to the impurity transport code 3DLIM, which is used to interpretively model collector probe (CP) deposition patterns measured in the ‘wall-SOL’. It is demonstrated that the deposition patterns are consistent with the DIVIMP predictions of near-SOL accumulation for the unfavorable- B T direction, and little/no accumulation for the favorable- B T direction. The wall-SOL CPs have thus provided the first experimental evidence, albeit indirect, of near-SOL W accumulation—finding it occurs for the unfavorable- B T direction only. For the favorable- B T direction, fast flows can largely prevent accumulation from occurring.

Mixed-material DIVIMP–WallDYN modeling, now incorporating ExB drifts, is presented that simultaneously reproduces tungsten (W) erosion and deposition patterns observed during the DIII-D metal rings campaign, in which a toroidally symmetric set of W-coated tiles were installed in the carbon (C) DIII-D divertor. Since most reactor plasma facing component (PFC) designs call for mixed-material environments, including ITER's W/Be environment, the divertor targets will quickly evolve into reconstituted surfaces of multiple elements. This work identifies controlling physics that affects material migration patterns in the divertor, which impact PFC lifetimes and impurity leakage from the divertor to the core. These simulations indicate that radial and poloidal ExB transport dominates over parallel force balance for high-Z impurities such as W in the divertor region of DIII-D. It is demonstrated that ExB drifts are required to reproduce the experimental observation of non-local W and C co-accumulation in a band ∼7–9 cm outboard of the outer-strike-point (OSP) W source, for attached L-mode conditions in the unfavorable ion grad-B drift direction. In addition, W gross erosion is localized to the region outboard of the OSP, as the formation of C co-deposits suppresses W erosion at the strike point. Time-dependent simulations with scaled ExB impurity drifts (60% of the OEDGE-calculated drift velocity) and W re-erosion quantitatively reproduce these features, including depth-resolved W/C ratios, within a factor of 2 over ∼115 s of accumulated plasma exposure. The location of co-deposition regions is shown to be well-represented by an analytic leakage model, driven largely by poloidal ExB drifts. Qualitative agreement is also found between campaign-integrated W deposition measurements and simulations for the favorable ion grad-B drift direction, the standard mode of operation for most tokamaks. These results imply that a long-term inward radial migration of material from the outer divertor through the private flux region may occur in future devices.

A set of experiments are planned to exploit the high SOL collisionality enabled by a tightly baffled slot divertor geometry to suppress tungsten leakage in DIII-D. A toroidal row of graphite tiles from the Small Angle Slot (SAS) divertor is being coated with 10–15 μ m of tungsten. New spectroscopic viewing chords with in-vacuo optics will measure the W gross erosion source from the divertor surface with high spatial and temporal resolution. In parallel, the bottom of the SAS divertor is changed from a flat to a ‘V’ shape. New SOLPS-ITER/DIVIMP simulations conducted with drifts using the planned ‘V’ shape predict a substantial reduction in W sourcing and SOL accumulation in either B × ∇B direction relative to either the old SAS divertor shape or the open, lower divertor. Dedicated studies are planned to carefully characterize the level of W sourcing, leakage, and scrape-off-layer (SOL) accumulation in DIII-D over a wide range of plasma scenarios. Various actuators will be assessed for their efficacy in further reducing high-Z impurity sources and leakage from the slot divertor geometry. This coupled code-experiment validation effort will be used to stress-test physics models and build confidence in extrapolations to advanced, high-Z divertor geometries for next-step devices.

Gross and net erosion rates of silicon from silicon carbide (SiC) coatings were measured in the divertor of DIII-D under well diagnosed reactor-relevant plasma conditions. Amorphous and crystalline SiC coatings on graphite with thickness of similar to 80 nm and similar to 250 mu m, respectively, were exposed near an attached outer strike point of lower single null L-mode plasmas using the Divertor Material Evaluation System (DiMES). Plasma density and electron temperature near the center of the coatings were n(e) similar to 4 x 10(19) m(-3) and T-e similar to 23 eV. Gross erosion of Si from all samples was measured spectroscopically using the Si II 636 nm line. It was found to be a factor of similar to 4 higher for the amorphous coatings compared to the crystalline one. The thin amorphous coatings allowed measurements of net Si erosion with Rutherford backscattering. Net average Si erosion rate of similar to 3 x 10(16) cm(-2) s(-1) was measured on the amorphous coatings with toroidal extent of 1 mm, where, according to ERO code modeling, the local redeposition of Si was about 30%. Using this rate, spectroscopic measurements, measured D+ ion fluxes, and corrections from ERO-OEDGE modeling, effective SXB coefficient for the Si II 636 nm line of similar to 52 and Si sputtering yield of similar to 0.017 Si/D were calculated. Deuterium retention on SiC coatings was measured by 2.5 MeV He-3 nuclear reaction analysis at 0.5-2.5 x 10(17) atoms cm(-2), consistent with retention due to implantation into a surface undergoing net erosion.

First results are reported from the 3D Monte Carlo far-SOL impurity transport code 3DLIM. Tungsten deposition profiles measured on a Collector Probe (CP) located in the far-SOL near the outer midplane, OMP, during W tracer experiments in DIII-D are reproduced by 3DLIM. Radial deposition profiles are replicated showing the effect that a decrease in connection length from the CP to the nearest wall contact point has on impurity transport to the probe, as well as the effect of assuming purely diffusive vs convective radial transport. For purely diffusive radial transport, a diffusion coefficient of 10 m2/s best reproduces deposition patterns on both sides of the CP, but for purely convective radial transport a speed of 125 m/s is shown to have better agreement with the ITF deposition profile. Deposition profiles show peaking in W content along the length of the CP edges that is also reproduced in 3DLIM, but only when assuming a convection-dominated SOL plasma parallel transport prescription for the background plasma. The degree of the peaking is shown to be a secondary indicator of the effective location of the W source in the near-SOL OMP relative to the far-SOL (near/far from the separatrix). Identifying the location of the effective source provides insight into near-SOL impurity dynamics, including the existence and location of impurity accumulation near the OMP separatrix. Such accumulation typically occurs in SOLPS and other edge code modeling, but has hitherto been difficult to confirm experimentally. The impurity density at the edge is the boundary condition for impurity levels in the confined plasma.

Predictive modeling of the closed Small Angle Slot (SAS) divertor with toroidal tungsten (W) rings in different poloidal locations is conducted to evaluate the impact of divertor closure on high-Z impurity sourcing and transport. These simulations utilize the DIVIMP code for W erosion, transport and deposition/redeposition, with background plasma solutions provided by SOLPS5.1. It is found that the level of W leakage is mostly determined by the divertor plasma conditions, the location of the W source and redeposition processes. High density plasmas in the SAS slot result in shorter ionization mean free paths than in the far-target region, resulting in higher redeposition rate. The induced friction force dominates in the near-target region, driving W impurities downstream towards the target surface. As a result, a W source at the strike point in the SAS divertor demonstrates lower net erosion and divertor leakage, despite higher gross erosion, compared to W sources in the far-target region.

Experiments carried out on DIII-D using a novel setup of isotopic tungsten (W) sources in the outer divertor have characterized how the W leakage from this region depends on both the exact source location and edge-localized mode (ELM) behavior. The sources are toroidally-symmetric and poloidally-localized to two regions: (1) the outer strike point (OSP) with natural abundance of W isotopes; and (2) the far-target with highly-enriched W-182 isotopes. With the use of a dual-faced collector probe (CP) in the main scrape-off layer (SOL) near the outside midplane and source-rate spectroscopy, a proxy for divertor impurity leakage is developed. Using this proxy, it is found that for the OSP W location, there is a nearly linear increase of leakage with the power across the separatrix (), which is consistent with the effect of an increased upstream ion temperature parallel gradient force in the near-SOL; trends in the pedestal density and collisionality are also seen. Conversely, it is found that for the far-target W location leakage falls off rapidly as increases and ELM size decreases, which is suggestive that ELM size plays a role in the leakage from this location. Indications for main SOL W contamination is evidenced by the measurement of large deposition asymmetries on the two opposite CP faces. These measurements are coupled with interpretive modeling showing SOL W accumulation near the separatrix furthest from both targets driven by forces parallel to the magnetic field. This experimental setup, together with the target and upstream W measurements, provides information on the transport from different divertor W source locations and leakage. These studies help to elucidate the physics driving divertor impurity source rates and leakage, with and without ELMs, and provide better insight on the link in the chain connecting wall impurity sources to core impurity levels in magnetic fusion devices.

We analyzed recent DIII-D tokamak tungsten divertor probe experiments using advanced, coupled, sputter erosion/redeposition, plasma, and surface response code packages. Modeling is done for ELMing H-mode, and L-mode plasmas, impinging on various size tungsten deposits on Divertor Material Evaluation System (DiMES) carbon probes. The simulations compute 3D, full kinetic, sub-gyromotion, impurity sputtering and transport, including changes in tungsten surface composition and response due to mixed deuterium and carbon ions irradiation. Per our analysis, ELM (edge localized mode) plasma sputtering in DIII-D mostly involves free-streaming high energy (∼500–1000 eV) D + and C +6 ions, with high near-surface plasma density. L-Mode sputtering is due to impurity sputtering (C, W) only, with lower density. All cases show complete redeposition of tungsten on the divertor, with significant redeposition on the tungsten spots themselves, and low self-sputtering. Comparison of ELM plasma gross tungsten erosion simulation results with in-situ spectroscopic data is good, as are code/data comparisons of net erosion using post-exposure Rutherford backscattering (RBS) data for the L-mode probes. The analysis, extrapolated to a full tungsten divertor, implies low net erosion and negligible plasma contamination from sputtering. These results support the use of high-Z plasma facing surfaces in ITER and beyond.