A coordinated DIII-D campaign exposed and comparatively assessed 44 advanced plasma-facing materials from 12 institutions, including four public-private fusion partnerships, to support fusion pilot plant wall and divertor material down-selection. Samples were exposed using the Divertor Materials Evaluation System (DiMES) under Ohmic, L-mode, and H-mode conditions with edge-localized modes, at 0.2-2.5 MW m$^{-2}$ on flush geometries and 10-15 MW m$^{-2}$ on 10$^{\circ}$ angled geometries. Engineered tungsten architectures retained integrity; long-fiber Wf/W showed the clearest crack-arrest behavior. W-Re and K-doped W showed near-ITER-W-like responses, while additively manufactured W-Ta showed heat-flux-sensitive mass losses of 0.64 mg for the flat sample and 2.19-2.87 mg for angled samples. After irradiation to 0.3 dpa at 550$^{\circ}$C, neutron-irradiated ITER-grade W retained 2.8 times more deuterium than pristine W, while TiB$_2$ showed the lowest D$_2$ release in the Ohmic set. VTaHfMo was the most stable refractory multi-principal-element alloy. NbC and (Nb$_{0.5}$Ta$_{0.5}$)C retained integrity with 0.02-0.03 mg mass loss, whereas ZrC lost 7 mg. CVD SiC retained macroscopic integrity but exhibited an effective Si erosion yield of 0.5, about 5-10 times above prior DIII-D trends. Renewable boron pebble rods underwent controlled recession; 13% of released boron was ionized near the outer strike point and up to 50% was recovered locally. Initial in-situ chromium gross-erosion measurements yielded values of order $10^{-2}$. Together, these results provide cross-material benchmarks for fusion pilot plant down-selection and future AI/ML-assisted plasma-facing-material development.

Silicon carbide (SiC) and SiC fiber-reinforced SiC matrix composites (SiC/SiC) are receiving renewed attention for use in next-generation fusion reactors due to their ability to withstand extreme conditions, including high temperatures, neutron irradiation, and plasma interactions. General Atomics Electromagnetic Systems (GA-EMS) has demonstrated significant progress in scaling up the fabrication of SiC/SiC, achieving high mechanical uniformity and meeting dimensional requirements in components up to 12 feet in length. Key developments are discussed including scale-up of the chemical vapor infiltration (CVI) process from lab-scale to full sized parts, high-dose (100 dpa) irradiation testing, nuclear-grade ceramic joining technologies, and production-focused quality control with the collective aim to establish SiC/SiC as a reliable solution for structural and functional components in fusion systems. Beyond manufacturing, the paper addresses supply chain barriers, particularly the limited availability and high cost of nuclear-grade SiC fiber. GA-EMS is developing a novel SiC fiber production method based on a thermochemical cure step that is anticipated to reduce costs compared to traditional approaches. Additionally, advancements in engineered SiC materials, such as SiC foams and tungsten-graded SiC composites, are discussed as promising solutions for specific fusion reactor components.

Characterization and testing of advanced plasma-facing materials (PFMs) for Fusion Pilot Plants (FPP) is being conducted at the DIII-D National Fusion Facility through the ongoing two-year FPP Candidate Materials Thrust. Year one tested 17 novel materials utilizing the Divertor Materials Evaluation System (DiMES), with samples analyzed pre- and post-experiment via SEM, EDS, and confocal microscopy. Repeatable reference discharges were developed to ensure uniformity between experiments, including a new strike-point rastering scenario to provide more uniform heat/particle flux across DiMES during ELMing H-mode discharges. Various sample geometries and temperatures were used to achieve FPP-relevant conditions, including samples angled 10 degrees towards the incident plasma flux and pre-heating up to 500 degrees C. The first exposure of liquid lithium (Li) capillary porous structures in a tokamak demonstrated uniform emission of Li vapor and suppression of Li droplets in H-mode when preheated to 350 degrees C. Dispersoidstrengthened W with 1 wt% TaC, TiC, and ZrC exposed to H-mode showed cracking and dispersoid ejection for all varieties except TiC, providing a clear down-selection. Ultra-high temperature ceramic materials TiB2 and ZrB2 showed minimal degradation under L-mode exposure. Silicon carbide (SiC) fiber composites showed arcing along edges, while CVD SiC remained pristine. Atmospheric plasma-sprayed W and SiC coatings endured H-mode exposure without macroscopic delamination; SiC exhibited granular ejection, while W showed increased outgassing. Additional W-based alloys were stress tested in H-mode, including Ni-based W heavy alloys, WfSiCf/W composites, W multi-principle element alloys, and functionally-graded W/SiC, to varying degrees of success.

Wide Pedestal QH-mode (WPQH) plasmas in the DIII-D tokamak show a sheath limited SOL, where electron density and temperature remain nearly constant along field lines from the midplane to the divertor. Consequently, parallel gradients are weak. The first Langmuir probe measurements in this regime point to a high sheath temperature with target electron temperature Te up to similar to 150 eV, leading to high carbon self-sputtering. Midplane carbon densities from SOLPS-ITER modeling of these plasmas in a double-null configuration fall far below experimental measurements unless full drifts are activated. In the drift-dependent SOLPS-ITER modeling, anomalous poloidally uniform, radially varying transport is adjusted to match measured radial electron temperature and density profiles in the pedestal and SOL region. The resulting carbon density (C6+) matches measured carbon densities just inside the separatrix. The C2+ density near the outer strike point is also consistent with spectroscopic imaging. With the ion B x del B drift towards the X-point, carbon in the lower divertors is redistributed from the private flux region to the high field side (HFS) and pushed upstream in the SOL by poloidal E xB drifts. The Bx del B drift dominates the radial flow of carbon as it moves upstream. Simulations with reversed toroidal field (ion Bx del B drift away from the X-point) show a radically different behavior, where carbon accumulates on the low field side in lower divertors, and is pushed towards the HFS in the upper divertors, indicating a strong effect of particle drifts on impurity distribution. In double-null configurations which are usually used in WPQH plasmas, these drift effects tend to counterbalance, so that the carbon density and Zeff are reduced at the outer mid-plane only modestly, by around 7% in the modeling. Further predictivemodeling indicates that the carbon density can be significantly reduced by an order of magnitude by implementing a single-null shape with reversed toroidal field.

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.

ITER-grade tungsten and dispersoid-strengthened tungsten samples with the top surface angled at similar to 15 deg toward the incident plasma flux were exposed to nine H-mode discharges with edge-localized modes (ELMs) in the lower divertor of the DIII-D tokamak using the Divertor Materials Evaluation System (DiMES). Surface damage included cracking and flaking of material on the two samples farthest away from the plasma strike point (SP) and significant melting of the two samples closest to the SP. Heat flux and thermal analysis tools new to DIII-D have been applied to better understand this material response and to help optimize the exposure conditions for future experi-ments. SMITER field-line tracing simulations based on IRTV data and EFIT equilibria estimate an average inter-ELM perpendicular heat flux q(perpendicular to,inter-ELM) on the angled surfaces of 10.1 to 19.6 MW/m(2) for a majority of the nine discharges, increasing to 15.6 to 24.5 MW/m(2) for the single, higher-power shot where samples melted. Fast camera data showed shallow intra-ELM melting and resolidification, which transitioned to bulk inter-ELM melting with melt motion in the (J) over right arrow x (B) over right arrow direction. About 50% of the protruding volume of the most affected sample was displaced via melt motion. SIERRA thermal modeling software was able to reproduce an onset time of melting consistent with fast camera data and final sample conditions, within <200 ms. Maximum surface temperatures of 3122 and 2787 K are estimated for the samples farthest away from the SP, while the closest samples achieve melting at 4067 and 4750 ms into the similar to 5000-ms plasma exposure. A +10% increase in both the SMITER q(perpendicular to,inter-ELM) calculations and the estimated ELM heat loads q(perpendicular to,ELM) was required to achieve this result, which is within the uncertainty of the diagnostic data but likely accounts for non-ideal geometry effects plus other physics uncertainties not included in this first iteration of modeling. This work provided valuable estimates of the three-dimensional temperature evolution to help better understand the observed surface morphology and internal recrystallization of samples, which are discussed in detail in a complementary paper (R. D. Kolasinski et al. "Recrystallization, Melting, and Erosion of Dispersoid- Strengthened Tungsten Materials During Exposure to Divertor Plasmas"). Benchmarking efforts with more diag-nosed DIII-D experiments are underway to further refine the SMITER and SIERRA models for DiMES. Future use of these tools will enable researchers to precisely target heat flux exposure conditions in DIII-D to test-but not exceed- the thermomechanical limitations of novel plasma-facing materials.

Group 5 metals are promising candidates for hydrogen (H) separation due to their exceptionally high permeability. However, they are prone to fracture-failure caused by H-induced embrittlement which limits their application. First-principles calculations were used to assess the effect of alloying element density in niobium-tungsten (Nb-W), niobium-nickel (Nb-Ni), and niobium-iron (Nb-Fe) alloys on H solubility, diffusivity, and permeability. Ground state energy calculations were performed to assess the average heat of solution and nudged elastic band calculations were performed to assess migration barriers between adjacent H interstitial positions. Migration barriers were used to parameterize a kinetic Monte Carlo diffusion model. Results show diminished H solubility and diffusivity with increasing dopant concentration. H delocalization and enhanced trapping are observed for Ni and Fe dopants, resulting in dramatic reduction in diffusivity and permeability. Nb-W alloys show high permeability and no enhanced trapping, suggesting Nb-W alloys may reduce H-induced embrittlement of membranes in metal foil pumps (MFP).

Boron coating of plasma-facing components (PFCs) is widely used in fusion devices to form surface coatings and chemical compounds on the PFCs, thus suppressing plasma contamination by impurities such as oxygen, carbon, and tungsten. Understanding the processes of hydrogenic species retention is crucial to assessing the viability of boron coating techniques in ITER and beyond. In this work, we deposited boronization films on silicon crystal and ITER-grade tungsten samples using a glow discharge boronization plasma of helium (He, 20 %) and diborane (B2D6, 80 %) in the DIII-D tokamak at the heated vessel temperature of 600 K. We performed post-mortem analysis of DIII-D boronization films for thickness utilizing a focused ion beam, deuterium retention utilizing thermal desorption spectroscopy (TDS), and chemical characterization utilizing X-ray photoelectron spectroscopy and Raman spectroscopy. The DIII-D boronization films were exposed to DIII-D tokamak L-, H-mode, and/or PISCES-E linear device D plasmas. We observed a D uptake for B:D = 5:2 in the DIII-D boronization film before D plasma exposures. The observed D retention was mainly in the B-C-D bond with a desorption peak of 1000 K. No B-D bond with a desorption peak around 700 K was found in the TDS spectrum as the DIII-D vessel was heated to 600 K. Exposures to both DIII-D L- and H-mode D plasma, which includes 1-3 % of C ion contamination, on the DIII-D boronization film at similar to 400 K gave a D uptake increase of similar to 20 %. Contrarily, D removal by similar to 30 % was observed after exposing the DIII-D boronization film to a carbon-free PISCES-E D plasma. This implies that future fusion devices should be careful to minimize C impurity contamination to reduce D retention.

Recent DIII-D experiments on Small Angle Slot (SAS) divertors have confirmed that a combination of divertor closure and target shaping can enhance cooling across the divertor target and increase energy dissipation, but with significant dependence on BT (toroidal magnetic field) direction. In these novel divertors, the roles of closure, target shaping, drifts, and scale lengths are all interconnected in optimizing dissipation, with the separatrix electron density neSEP being the key parameter associated with the level of dissipation/detachment. After modifying the original flat-targeted graphite SAS to include a V shape with a tungsten coating on the outer side of the divertor (SAS-VW), matched series of discharges were run to compare to detailed SOLPS-ITER modeling. Experimentally, when run as designed with the outer strike point at the slot vertex, SAS-VW requires nearly identical neSEP for detachment as the original SAS, with little difference in dissipation for the new geometry. This is in contrast to (1) earlier modeling predictions that a small change of the SAS geometry to a V shape should enhance dissipation at the same neSEP for magnetic configurations having better H-mode access (ion B x del B drift directed into the divertor), and (2) despite the achievement of significantly higher (2-7x) neutral pressures and compression in the SAS-VW slot. Comparisons of experimental density scans to the most recent SOLPS-ITER modeling with ExB drifts show reasonable agreement for dissipation/detachment onset when using separatrix density as the independent parameter. In order to help understand the discrepancy in modeled vs actual performance for the new configuration, additional measurements varying gas injection location and impurity injection were undertaken. In-slot D2 gas fueling is more effective (5-22 %) in promoting detachment, in accord with modeling. In-slot impurity injection (N2 or Ne) can yield 30 % lower core Zeff and 15 % less confinement degradation after detachment compared to main chamber puffing, as well as relatively lower tungsten leakage from the divertor. Modeling can also reproduce the improved detachment seen as the strike point moves inboard of the slot vertex. While we can explain the effects of the most important parameters causing energy dissipation in these slot divertors, it remains that many aspects of their behavior cannot be accurately modeled using state-of-art codes such as SOLPS-ITER. This is of concern for future model-driven designs utilizing similar V-shaped geometries.

The understanding and prediction of silicon carbide (SiC) material evolution exposed to SOL plasma conditions is of prime interest because SiC represents a promising main chamber wall plasma-facing material for next-step fusion devices (low hydrogenic diffusion, good mechanical and thermal properties under neutron irradiation). Gross and net Si erosion rates from SiC surfaces in contact with a well-diagnosed L-mode plasma in the DIII-D tokamak have been simulated and the surface concentrations of impurities have been tracked as a function of time. Coupled simulation of surface model and impurity transport demonstrates amorphization of crystalline SiC exposed to L-mode plasma due to the accumulation of displacement damages under ion irradiation. This affects the lifetime of SiC plasma facing components. Surface evolution is tightly coupled to impurity transport in the plasma and therefore needs to be integrated with impurity transport simulations to effectively predict Si erosion rates and sub-surface concentrations as a function of time. The simulation workflow couples a semi-analytical surface model to the impurity transport code GITR. The surface model is a homogeneous mixed-material model that tracks physical & chemical sputtering and reflection of impurities. Gross erosion is primarily influenced by the background plasma parameters and redeposition patterns are mainly influenced by the prompt redeposition due to the gyro-orbits of impurity ions. Although crystalline form of SiC is preferable for fusion wall applications because of resistance to neutron irradiation, this work indicates that crystalline SiC will undergo amorphization under D plasma contact with implications of higher sputtering and fuel retention. These results direct us to explore the effects of amorphization on crystalline SiC and further the physics basis of SiC usage as first wall material for fusion environments.

The performance of silicon carbide as an alternative plasma facing material (PFM) was studied at various irradiation conditions relevant to ion energies and fluxes of a fusion reactor. This analysis involves detailed modeling of subsurface plasma/material interactions, sputtered particle transport above the surface and redeposition, and related changes in material composition and microstructure induced by steady-state and Edge Localized Mode ion fluxes. Transition of a crystalline SiC surface to semi-crystalline and amorphous phases was analyzed based on advanced modeling of DIII-D tokamak experiments where SiC was irradiated in single- and multiple- L-mode and H-mode discharges. This analysis shows that displacement damage, particle deposition/redeposition, and D accumulation on the SiC divertor surface can lead to significant microstructural changes that result in enhanced sputtering erosion in comparison with the original crystalline material. However, the resulting total net erosion rate for a full-coverage, advanced tokamak, SiC coated divertor may well be acceptably low. Moreover, the C sputtering yield from the evolved SiC surface can be seven times lower than from a pure graphite surface; this would imply significantly reduced tritium co-deposition rates in a D-T tokamak reactor, compared with a pure carbon surface. It was also determined that chemical sputtering of both C and Si should not result in any noticeable effect on the net erosion, for attached plasma regimes. Our results thus show encouraging results overall for use of SiC as a PFM in tokamaks.

Experiments using the V-shaped closed slot tungsten (W) coated SAS-VW divertor in DIII-D studied the effects of the B-T direction on core contamination of eroded tungsten from a closed slot divertor configuration. Core W content is inferred using soft-X ray tomography (SXR) and vacuum ultraviolet spectroscopy (SPRED), while W divertor erosion is inferred from visible spectroscopy of W emission (400.9 nm) measured by in-slot filterscopes (filtered photo-multipliers). Post-mortem analysis from the campaign discovered tile misalignment leading to suspected pronounced leading-edge erosion in the unfavorable B-T direction (ion B -> x del B -> drift away from divertor) likely not captured by diagnostics. However, empirical findings show up to similar to 2-3x larger core contamination in the favorable B-T direction even considering no additional W erosion from leading edges. A "source-to-core efficiency factor" is derived to estimate the effects of leading-edge erosion and compare W contamination for two pairs of H-mode discharges in opposite B-T directions. While having differing absolute parameters, similar core impurity density gradients suggest comparable core impurity transport. These results show that favorable B-T may have stronger source-to-core pathways for W impurities sourced from the outer divertor region. Possible explanations could include the effects of E -> xB -> drifts on W transport in the scrape-off-layer (SOL) as well as previously determined fast SOL inner target directed flows in favorable B-T.

Erosion patterns in a 3D magnetic topology are significantly altered when compared to an axisymmetric scenario due to differences in the impurity transport in a plasma environment. When resonant magnetic perturbations are applied in L-Mode discharges, eroded impurities can buildup in regions where an axisymmetric plasma would otherwise lead to zones of net erosion across the full toroidal length of the divertor. The reduction on the local and integrated erosion observed across a parametric study of the anomalous diffusion and impurity content in a plasma, can lead to the extension of divertor lifetimes in low power scenarios when 3D fields are applied. By contrast, in axisymmetric scenarios, most of the carbon eroded from the divertor are carried away to the far scrape-off layer.

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 B -> x del 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.

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.

W-SiC composite material is a promising plasma-facing material candidate alternative to pure W due to the low neutron activation, low impurity radiation, and low tritium diffusivity of SiC while leveraging the high erosion resistance of the W armor. Additionally, W and SiC have high thermomechanical compatibility given their similar thermal expansion rates. The present study addresses the synthesis and performance of compositionally graded W-SiC films fabricated by pulsed-DC magnetron sputtering. Compositional gradients were characterized using transmission electron microscopy (TEM) and energy-dispersive X-ray spectroscopy (EDS), and crystallographic information was obtained using electron diffraction and X-ray diffraction (XRD). Samples were exposed to L-mode deuterium plasma discharges in the DIII-D tokamak using the Divertor Material Evaluation System (DiMES). Post-mortem characterizations were performed using scanning electron microscopy (SEM) and XRD. Electron diffraction and XRD showed that the compositionally graded W-SiC films were composed of polycrystalline W and amorphous SiC with amorphous W+SiC interlayers. No macroscopic delamination or microstructural changes were observed under mild exposure conditions. This study serves as a preliminary examination of W-SiC compositionally graded composites as a potential candidate divertor material in future tokamak devices.

Many existing and future tokamaks with tungsten divertors operate, or will operate, with low- Z impurity seeding, but the direct effect of these seeded impurities on tungsten Scrape-off-Layer (SOL) transport has not been explored in detail. This paper reports on a DIII-D experiment designed to test how tungsten divertor leakage from the Small-Angle Slot V-Shaped, tungsten-coated divertor is impacted by neon seeding at a variety of injection rates and poloidal injection locations. Measurements from the experiment show an inverse relationship between the neon injection rate and the tungsten core penetration factor. Interpretive modeling is performed with a combination of the SOLPS-ITER and DIVIMP codes to assess the underlying tungsten behavior. The modeling results show that the reduction in tungsten divertor leakage is driven by both an increase in the divertor collisionality as well as a reduction in the ion temperature gradient near the divertor target. Collisions between low- Z impurities and tungsten impurities are found to have a significant impact on the tungsten SOL transport, such that ignoring the low- Z impurity collisional effects on the tungsten transport can result in an overestimate of the divertor leakage by an order-of-magnitude. Given the importance of these localized interactions, neon seeding from the closed, slot-like divertor has a clear advantage in being able to reduce tungsten divertor leakage without the high levels of neon core contamination that occur when seeding from other poloidal locations.