The SPARC tokamak project, currently in engineering design, aims to achieve breakeven and burning plasma conditions in a compact device, thanks to new developments in high-temperature superconductor technology. With a magnetic field of 12.2 T on axis and 8.7 MA of plasma current, SPARC is predicted to produce 140 MW of fusion power with a plasma gain of Q ≈ 11, providing ample margin with respect to its mission of Q > 2. All tokamak systems are being designed to produce this landmark plasma discharge, thus enabling the study of burning plasma physics and tokamak operations in reactor relevant conditions to pave the way for the design and construction of a compact, high-field fusion power plant. Construction of SPARC is planned to begin by mid-2021.

The presence of magnetic dust can be an important issue for future fusion reactors where plasma breakdown is critical. Magnetic dust has been collected from contemporary fusion devices (FTU, Alcator C-Mod, COMPASS and DIII-D) that feature different plasma facing components. The results of morphological and elemental analysis are presented. Magnetic dust is based on steel or nickel alloys and its magnetism is generated by intense plasma-material interactions. In spite of the strong similarities in terms of morphology and composition, X-ray diffraction analysis revealed differences in the structural evolution that leads to non-trivial magnetic responses.

The erosion and redeposition of first-wall armor materials is a problem in nuclear fusion devices with carbon walls, where deuterium, tritium, and (eroded) carbon present in the plasma are deposited on the walls of the device, trapping the expensive and radiologically hazardous tritium. Thermo-oxidation, in which vessel surfaces are heated and oxygen containing gas is injected, is a possible solution. It results in the production of carbon oxides and tritiated water vapor, which can be pumped out by the vacuum pumps and recycled in a tritium recycling facility. In the present study, thermogravimetric analysis was used to measure the mass loss (or gain) of codeposited specimens from the General Atomics DIII-D National Fusion Facility under thermo-oxidation, in addition to laser thermal desorption spectroscopy. X-ray photo-electron spectroscopy was also used in this work to examine the tile's surface composition pre and post oxidation. Dust scraped from the specimen was also studied, as this is a surrogate for dust that naturally falls from the tile codeposits and builds up in the tile gaps. One key conclusion is that boron oxides form where boron is present in the codeposit as an impurity, and these oxides dominate the weight-change behavior of the codeposit specimens for long exposures.

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.

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.

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.

Experiments with the lower divertor of DIII-D during the Metal Rings Campaign (MRC) show that the fraction F of atomic D in the total recycling flux is material-dependent and varies through the ELM cycle, which may affect divertor fueling. Between ELMs, F-C similar to 10% and F-W similar to 40%, consistent with expectations if all atomic recycling is due to reflections. During ELMs, FC increases to 50% and F-W to 60%. In contrast, the total D recycling coefficient including atoms and molecules R stays close to unity near the strike point where the surface is saturated with D. During ELMs, R can deviate from unity, increasing during high energy ELM-ion deposition (net D release) and decreasing at the end of the ELM which leads to ability of the target to trap the ELM-deposited D. The increase of R >1 in response to an increase in ion impact energy E-i has been studied with small divertor target samples using Divertor Materials Evaluation System (DiMES). An electrostatic bias was applied to DiMES to change E-i by 90 eV. On all studied materials including C, Mo, uncoated and W-coated TZM (>99% Mo, Ti, and Zr alloy), W, and W fuzz, an increase of E-i transiently increased the D yield (and R) by similar to 10%. On C there was also an increase in the molecular D-2 yield, probably due to ion-induced D-2 desorption. Despite the measured increase in F on W compared to C, attached H-mode shots with OSP on W during MRC did not demonstrate a higher pedestal density. About 8% increase in the edge density could be seen only in attached L-mode scenarios. The difference can be explained by higher D trapping in the divertor and lower divertor fueling efficiency in H- versus L-mode.

on core impurity accumulation in DIII-D high performance discharges * D.M. Thomas, T. Abrams, J. Barton, J.A. Boedo, A.R. Briesemeister, D. Buchenhauer, I. Bykov, C.P. Chrobak, R. Ding, D. Donovan, J.D. Elder, B.A. Grierson, H.Y Guo, J. Guterl, E.T. Hinson, E.M Hollmann, C.J. Lasnier, A.W. Leonard, M.A. Makowski, A.G. McLean, R. Nygren, T.W. Petrie, D.L. Rudakov, P.C. Stangeby, E.A. Unterberg, B. Victor, W.R. Wampler, H.Q. Wang, J.G. Watkins, M. Zach

We present analysis and modeling of Al sputtering and ionization in attached, low-power L-mode plasmas near the outer divertor strike point of the DIII-D tokamak. Al serves as a useful proxy for Be, the low-Z main wall material for ITER and JET that will undergo significant divertor plasma contact upon migrating from the first wall to the divertor. Al is easily distinguishable from background sources in DIII-D (namely C and B), has a high physical sputtering yield similar to Be, and has a long ionization mean free path compared to its gyro radius (& Unknown;lambda & Unknown;) /r(gyro) similar to 2.5), Using neutral Al emission imaging techniques, we measured a toroidal and radial asymmetry in the shape of the photo-emission plumes of sputtered neutral Al that was consistent with previously observed asymmetry in the distribution of redeposited Al in these experiments. We propose that the main cause of the emission and redeposition asymmetry is due to a sputtering anisotropy caused by near-grazing angle incident ions. The observed emission asymmetry was reproduced using a simple emissioil/ionization model that included full angular distributions of sputtering yield and energy calculated by SDTRIM.SP, but not when symmetric, mono-energetic cosine sputtering distributions were assumed. We used an ion orbit tracking model to calculate the distributions of ion impact energies through the potential gradient in the magnetic pre-sheath and Debye sheath. We found that with the magnetic field pitch angle (1.5 degrees-2 degrees with respect to the surface plane), the majority of ions strike the surface at <15 degrees with respect to the surface plane, leading to angular sputtering yield and energy distributions with significant forward-scattering bias. We also observed surface microstructure consistent with directional sputtering and ion flux shadowing expected from the calculated ion incidence angles.

Tokamak plasma facing components have surface roughness that can cause microscopic spatial variations in erosion and deposition and hence influence material migration, erosion lifetime, dust and tritium accumulation, and plasma contamination. However, high spatial resolution measurements of deposition on a sub-pm scale of surface roughness have been lacking to date. We will present elemental images of graphite samples from NSTX-U and DIII-D DiMES experiments performed with a Scanning Auger Microprobe at sub-micron resolution that show strong microscopic variations in deposition and correlate this with 3D topographical maps of surface irregularities. The NSTX-U samples were boronized and exposed to deuterium plasmas. The DiMES samples had localized Al and W films and were exposed to dedicated helium plasmas. Topographical maps of the samples were performed with a 3D confocal optical microscope and compared to the elemental deposition pattern. The results revealed localized deposition concentrated in areas shadowed from the ion flux, incident in a direction calculated by taking account of the magnetic sheath.

It is important to develop a predictive capability for the tungsten source rate near the strike points during H-mode operation in ITER and beyond. H-mode deuterium plasma exposures were performed on W-coated graphite and molybdenum substrates in the DIII-D divertor using DiMES. The W-I 400.9 nm spectral line was monitored by fast filtered diagnostics cross calibrated via a high-resolution spectrometer to resolve inter-ELM W erosion. The effective ionization/photon (S/XB) was calibrated using a unique method developed on DIII-D based on surface analysis. Inferred S/XB values agree with an existing empirical scaling at low electron density (ne) but diverge at higher densities, consistent with recent ADAS atomic physics modeling results. Edge modeling of the inter-ELM phase is conducted via OEDGE utilizing the new capability for charge-state resolved carbon impurity fluxes. ERO modeling is performed with the calculated main ion and impurity plasma background from OEDGE. ERO results demonstrate the importance a mixed-material surface model in the interpretation of W sourcing measurements. It is demonstrated that measured inter-ELM W erosion rates can be well explained by C -> W sputtering only if a realistic mixed material model is incorporated.

The angular distribution and sputtering yield of beryllium exposed to helium plasma are estimated from analysis of line-integrated 2D imaging of Be-I line emission in a steady-state linear plasma device. As the surface nanostructure forms during plasma exposure on a ∼100 s timescale (corresponding to a fluence of order 1020/cm2) from nearly mono-energetic ion bombardment, a narrowing of the beryllium sputtering angle and a significant (∼5×) drop in sputtering yield are observed. These trends are found to be qualitatively consistent with modeling taking into account the effect of the surface morphology on sputtering yield and angular distribution.

Dedicated DIII-D experiments coupled with modeling reveal that the net erosion rate of high-Z materials, i.e. Mo and W, is strongly affected by carbon concentration in the plasma and the magnetic pre-sheath properties. Different methods such as electrical biasing and local gas injection have been investigated to control high-Z material erosion. The net erosion rate of high-Z materials is significantly reduced due to the high local re-deposition ratio. The ERO modeling shows that the local re-deposition ratio is mainly controlled by the electric field and plasma density within the magnetic pre-sheath. The net erosion can be significantly suppressed by reducing the sheath potential drop. A high carbon impurity concentration in the background plasma is also found to reduce the net erosion rate of high-Z materials. Both DIII-D experiments and modeling show that local (CH4)-C-13 injection can create a carbon coating on the metal surface. The profile of C-13 deposition provides quantitative information on radial transport due to E x B drift and the cross-field diffusion. The deuterium gas injection upstream of the W sample can reduce W net erosion rate by plasma perturbation. In H-mode plasmas, the measured inter-ELM W erosion rates at different radial locations are well reproduced by ERO modeling taking into account charge-state-resolved carbon ion flux in the background plasma calculated using the OEDGE code.

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 similar to 100 nm 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 similar to 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. (C) 2016 Elsevier Ltd. This is an open access article under the CC BY-NC-ND license.

Unipolar arcing was an important mechanism of metal surface erosion during the recently conducted Metal Rings Campaign in DIII-D when two toroidally continuous tile rings with 5 cm wide W-coated TZM inserts were installed in graphite tiles in the lower divertor, one on the floor and one on the shelf. Most of the arc damage occurred on the shelf ring. The total amount of W removed by arcing from the affected similar to 4% of the shelf ring area was estimated similar to 0.8. x. 10(21). at., about half of the total amount of W eroded and redeposited outside the inserts (1.8 +/- 0.9) x 10(21). at. The rings were exposed for a total of similar to 480 discharges, an equivalent of plasma time on W surfaces (with I-p > 0.5 MA) similar to 10(3) s. Arcing was monitored in situ with WI (400.9 nm) filtered camera and photomultipliers and showed that: (i) arcing only occurred during ELMs and disruptions, (ii) arcing rate was much lower on the floor than on the shelf ring, and (iii) arcing had a low cut off power flux density about 2. MWm(-2). About half of arc tracks had large 10. pitch angle and probably were produced during disruptions. Such tracks were only found on the shelf. Moderate toroidal variation of the arc track density and W erosion with nearly n = 1 pattern has been measured.

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.