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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Plasma Guns for Magnetized Fuel Targets for PJMIF

In plasma jet driven magneto-inertial fusion (PJMIF) an array of discrete supersonic plasma jets is used to form a spherically imploding plasma liner, which then compresses a magnetized plasma target to fusion conditions. With funding from ARPA-E’s ALPHA program and from Strong Atomics LLC, HyperJet Fusion Corp and HyperV Technologies Corp. previously developed the plasma guns required for an experimental demonstration of the plasma liner formation part of the concept. A 36-gun demonstration of an imploding spherical plasma liner is currently underway on the PLX facility at Los Alamos National Laboratory. This present project addresses the next step required for a complete PJMIF concept, i.e. developing the magnetized plasma target. We proposed to form the target by stagnating a number of magnetized plasma jets in the center of the target chamber. This is accomplished by adding a bias field coil to the plasma liner gun to form a magnetized plasma jet. This experimental development took place at HyperJet in a geometry replicating the bias field environment that will be seen on a PLX port so the results are directly transferrable to the PLX experiment. The objective of this effort was the technical development and characterization of a new magnetized plasma jet using a high-performance, high momentum flux, contoured-gap coaxial plasma gun appropriate for use on the next stage of the PLX experiment. Electromagnetic modeling of the coil indicated the coil was best placed around the aluminum tubes of the gun transmission line, rather than around the chamber port as originally proposed. This maximized field strength in the breech and yielded much better flux linkage between gun electrodes. A 30-turn coil was ultimately implemented, allowing long pulses that could diffuse through the metal walls on the timescale of interest. \machtwo modeling predicted that less capacitance in the main PFN could potentially improve plasma jet velocities due to better matching of the current to the smaller plasma mass and the existing electrode contour designed to suppress blowby. This proved true, as testing showed markedly improved performance when the original 600uF bank was reduced to 400uF. A number of diagnostics were built and/or upgraded in order to characterize the plasma jets, inluding laser interferomtery for density, photodiodes for velocity, Bdot probes for magnetic field measurements, a Triple probe for temperature measurements, and spectroscopy for impurity content. A plasma gun with the 30-turn magnet coil installed, a 70% reduction in gas valve plenum volume, and a 33% reduction in main PFN capacitance produced a dense, high velocity, well magnetized plasma jet. We met or exceeded virtually all of the plasma jet parameter goals. Peak velocity of 135 km/s exceeded the 100 km/s goal by 35%, while the peak density of over 1.0x10 15 cm -3 was 3.3 times the goal of >3.0x10 14 cm -3 . Shot-to-shot repeatability is excellent, with a jitter of less than 300ns observed on the arrival fronts of the photodiode signals from one shot to the next. Average plasma jet lengths of 33cm (at 120km/s) were slightly longer than the 20cm goal, but jets as short as 11.8cm were observed at 135km/s. Mass is much higher than the targeted goal, averaging 106ug per shot compared to 20ug. The magnetic field is still a bit lower than desired, with a maximum to date of ~811G, compared to the goal of 1000G. Average values, though, were typically in the 300-450G range, when measured further downstream after some expected in flight decay. Temperature measurements are still a work in progress. Increasing the B field and completing temperature measurements will be continued on into the ongoing BETHE project.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

A Dirac R-matrix calculation for the electron-impact excitation of W +

Abstract Aims : tungsten has been chosen for use as a plasma facing component in the divertor for the ITER experiment, and is currently being used on existing tokamaks such as JET. W + plays an integral role in assessing the impurity influx from plasma facing component of tokamaks and subsequent redeposition. Together with previously calculated a neutral tungsten electron-impact dataset this study allows us to determine neighbouring spectral lines in the same wavelength window of the spectrometer, and detect if there is strong blending of overlapping lines between these two ion stages as well as providing ionisation per photon ratios for both species. The new data is to be used for tungsten erosion/redeposition diagnostics. Methods : a significantly modified version of the GRASP0 atomic structure code in conjunction with DARC (Dirac Atomic R -matrix Code) are used to calculate the Einstein A coefficients and collisional rates used to generate a synthetic W II spectrum. The W II spectrum is compared against tungsten spectral emission experiments. Results : this study is used to model the spectrum of W ii , providing the predictive capability of identifying spectral lines from recent experiments. These results provide an integral part of impurity influx and redeposition determination, as the ionisation rates may be used to calculate S/XB ratios.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2022

High-Energy X-ray Diffraction Microscopy for Nuclear Forensics FY2022 Project Report

Morphological information on nuclear material has been identified using visible light and scanning electron microscopy. These identify qualitative differences in particle morphology. Three-dimensional imaging of materials through alternating scanning electron microscopy imaging and focused ion beam milling has also been used. Unfortunately, these techniques are time- and labor-intensive, with significant sample preparation required and lengthy analysis times. Further, the resulting 3D images are qualitative, require manual identification, and do not capture statistically-representative populations. High energy X-ray 3D imaging using a direct-beam or diffracted-beam (High-Energy Diffraction Microscopy) have been developed at the Advanced Photon Source and can produce quantitative information on grains (phase, location, etc.) and pores (size distribution, sphericity) in a material. These techniques require only minutes to characterize a sample volume and are non-destructive, thus suitable for a wide range of existing samples and for confirmatory analyses to be carried out using conventional microscopy techniques. In this first year of the project, all uranium oxide samples were synthesized and characterized using conventional analyses by the analytical chemistry laboratory. Conventional analysis methods included powder x-ray diffraction, scanning electron microscopy, impurity analysis via inductively coupled plasma mass spectrometry, and infrared spectroscopy. Impurity analysis shows a drop in boron content from UO 3 to the lowest U 3 O 8 calcination temperature, but otherwise no appreciable difference in any sample. Analysis of diffraction data shows a flip of peaks from UO 3 dominated for the 600 °C calcined sample to U 3 O 8 dominated at 700 °C and 800 °C. Analysis of scanning electron microscopy images shows that with increased calcination temperature the size distribution of particles seems to increase and broaden. Both of these last findings are in line with previously published data, though this work used significantly fewer particles to simply show similar trends instead of getting truly quantitative particle analysis. Infrared analysis similarly shows ingrowth of U 3 O 8 as calcination temperature is increased, along with depression of peaks associated with UO 3 and water. Samples were prepared for analysis at the Advanced Photon Source at beamline 1-ID. It is anticipated that analysis will occur in November of 2022. AI/ML techniques to de-noise data coming out of 1-ID during the analyses was also developed during this time using previously gathered data. Preliminary results using a self-supervision technique called Noise2Selfshow good de-noising of data. Once the uranium oxide samples are analyzed, real data will be used to test the de-noising and other AI/ML techniques that may be developed in the second year of the project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Achievement of highly radiating plasma in negative triangularity and effect of reactor-relevant seeded impurities on confinement and transport

The first achievement of highly radiating plasmas in negative triangularity (NT) is shown with an operational space featuring high core radiation at high Greenwald fraction obtained with the injection of reactor-relevant seeded gases. These NT shape diverted discharges reach high values of normalized plasma pressure ( β N > 2) at high radiation fraction with no edge localized modes (ELMs) and a simultaneous reduction of the divertor heat flux. Here, we demonstrate that as long as the impurity level in the core is kept low to avoid excessive fuel dilution and impurity accumulation, integration of NT configuration with high radiation fraction not only is achievable but it can lead to confinement improvement with stabilization effects originating from collisionality, E × B shear and profiles changes due to impurity radiation cooling. The underlying physics mechanism is robust and reproducible and holds for a variety of impurity species. The absence of the requirement to stay in H-mode translates in a higher core radiation fraction potentially allowed in NT shape effectively mitigating the power exhaust issue. The results presented here demonstrate a path to high performance, ELM free and highly radiative regime with reactor-relevant seeding gases making this regime a potential new scenario for reactor design.

confinement improvement↗
DOE OSTI2020

Thermal stability of oxidized ultrathin Li films on TZM for plasma facing components

For improved understanding of the behavior of Li plasma facing components (PFCs) in the presence of oxygen impurities, we report on the thermal stability of ultrathin (up to 1.0 nm) LiO x films on a titanium-zirconium-molybdenum (TZM) alloy substrate over the temperature range of 310–1400 K. LiO x films were prepared by either post-oxidation of Li films or Li films deposited on an oxygen-precovered TZM surface. Thermal stability of these films was studied using a combination of temperature programmed desorption (TPD), X-ray photoelectron spectroscopy (XPS), and low-energy ion scattering (LEIS). For post-oxidized Li films, TPD showed that no Li desorption occurred until temperatures above 620 K, and then Li desorbed from the surface via at least three desorption peaks. The O 1s XPS spectra on the post-oxidized films at 310 K indicate the formation of lithium oxide (Li2O) and peroxide/hydroxide (Li 2 O 2 /LiOH). The peroxide/hydroxide converts to oxide after heating to 680 K with no desorption of Li or O 2 , and then this film decomposes to liberate Li into the gas phase while leaving oxygen at the TZM surface. Heating the LiO x films to 1070 K led to a solid-state reaction with surface Mo atoms of the substrate to form a condensed binary lithium molybdenum oxide (LixMoOy) phase, which upon further heating decomposed to liberate gas phase binary lithium molybdenum oxide species. Li deposition on an oxygen-precovered TZM surface formed a complex, highly stable, oxygen rich, Li-O-Mo interfacial oxide. Li desorption from multilayer Li films on oxygen-precovered TZM surfaces occurred in a metallic Li multilayer peak and three other oxide-derived peaks. These results support the improved thermal stability of oxidized lithium as compared to metallic Li and have implications for Mo impurity transport into the plasma at high temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Self-consistent modeling of tokamak edge plasma transport with lithium sources

Magnetic confinement fusion devices require effective heat and particle exhaust solutions on the divertor plates to operate sustainably, especially under reactor-relevant conditions. Liquid lithium divertors have been proposed to address two major challenges: control of excessive heat flux to plasma-facing components through vapor shielding and minimization of core plasma contamination from impurities. The National Spherical Torus Experiment-Upgrade (NSTX-U) will explore lithium as a divertor material due to its potential to meet both objectives. We present a self-consistent coupling framework between the plasma boundary transport code UEDGE and the lithium wall transport code Wall–Li to evaluate the feasibility and operational limits of lithium-based divertors. The model aims to optimize lithium sourcing levels to prevent core plasma contamination via fuel dilution while ensuring divertor protection through vapor shielding. This integrated framework, applicable to any tokamak with lithium sources, dynamically adjusts lithium sourcing based on local plasma conditions and surface temperature. The coupled model is tested using NSTX-like geometry and plasma conditions to assess its performance and reliability. Wall–Li calculates lithium fluxes from plasma-facing components, incorporating physical sputtering, thermally enhanced sputtering, and evaporation driven by surface temperature and ion flux. These fluxes are reintroduced into UEDGE as neutral lithium atoms, enabling simulation of their transport and distribution within the plasma. UEDGE computes plasma and neutral transport, surface heat flux, and iteratively feeds this information back to Wall–Li. A small time step is employed to ensure numerical stability and convergence, enabling accurate simulations over typical tokamak discharge durations. This integrated modeling approach provides a robust tool for identifying operational regimes that balance effective lithium sourcing with minimal core plasma contamination, offering critical insights for optimizing lithium-based divertor systems in current and future fusion devices.

Magnetic confinement fusion↗
DOE OSTI2020

FY2018 Fusion Theory and Simulation Milestone on Plasma Materials Interaction (PMI) (Final Report)

The interaction of the boundary plasma with the material surfaces in magnetically confined plasmas is among the most critical problems in fusion energy science. In FY2018, perform high performance computing simulations with coupled boundary plasma physics and materials surface models to predict the fuel recycling and tritium retention of the ITER divertor for D-T burning plasma conditions, accounting for erosion, re-deposition and impurity transport in the plasma boundary, and an initial evaluation of the influence of material deposition on the recycling and retention.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Impact of gas injection location and divertor surface material on ITER fusion power operation phase divertor performance assessed with SOLPS-ITER *

Abstract The ITER divertor design and performance assessment, primarily based on the SOLPS-4.3 burning plasma database (Pitts R. et al 2019 Nucl. Mater. Energy 20 100696), assumes the use of beryllium (Be) as the divertor surface material and the injection of gas from the main chamber top. However, the current ITER baseline favors gas injection from the more toroidally symmetric sub-divertor region. This paper evaluates the implications of these assumptions for divertor performance in the ITER fusion power operation phase. The impact of the divertor surface material and the gas injection location on the main ions mirrors the hydrogen only low power phase scenario shown in Park J.-S. et al (2020 Nucl. Fusion 61 016021). However, during burning plasma operation, extrinsic impurity seeding will be required. In the case of neon (Ne), studied here, impurity retention is influenced by both the divertor surface material and the fueling location. Neon leakage increases due to more energetic reflection from tungsten than beryllium, but equivalent divertor performance can be achieved by adjusting the neon seeding rate. While the impurity seeding location does not affect the distributions of impurity or radiation, the fueling location does. Top fueling provides local ionization sources mainly in the mid-SOL under detached conditions, enhancing divergences of the flux there (source-driven flow), bringing stagnation points close to the fueling location, and equilibrating flows towards both targets. In contrast, the global flow pattern (in the absence of fluid drifts) in the case of sub-divertor fueling is biased towards the inner target. Impurity flows, driven by force balance, largely mirror those of the main ion flow, including the stagnation point. The case with top fueling enhances Ne retention and corresponding radiation in the outer divertor, effectively reducing the total and peak target heat fluxes by 20%–40%, compared to the case with divertor fueling. Meanwhile, the case with outer target fueling also achieves similar reductions by enhancing plasma-neutral interactions. These results suggest the possibility that the selection of the fueling location and throughput can be used as an actuator to control impurity divertor retention and divertor radiation asymmetry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Determination of impurities in cubic boron nitride (cBN) by inductively coupled plasma mass spectrometry (ICPMS)

The phase stability of cBN is critically dependent on the content and amount of impurities present. While studies have shown qualitatively that the addition of certain impurities can either increase or decrease the cBN-hBN phase transition temperature, studies have yet to show this effect quantitatively. This is because analytical methods to characterize minor and trace impurities in cBN are currently lacking. To this end, we present flux fusion methods capable of fully digesting cBN samples for ICPMS analyses. The materials used in this study are commercially available cBN powders in three different sizes. Two types of fluxes (LiOH and NaOH·H 2 O) are utilized so that both Li and Na impurities can be determined. To ensure the accuracy of Si measurements, a column chromatography method has been developed to purify Si from flux materials in order to increase the amount of dissolved sample sent to the ICP. Detection limits have been determined using procedural blanks and are found to be below 20 μg/g for all 58 elements analyzed except for Si, for which the detection limit is 312 μg/g. Nonetheless, all cBN powders studied are found to have high Si impurity levels that are above detection limit. This study provides a much-needed analytical protocol to enable comprehensive screening of impurities in cBN.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Reply to Comment on ‘Deuterium supersaturated surface layer in tungsten: ion energy dependence’

Abstract We reply to the comment by Li et al (submitted to Nucl. Fusion with this response) on our recent paper Nishijima et al (2023 Nucl. Fusion 63 126003). In this response, we address the existence of an incident ion energy, E i , threshold for the deuterium (D) supersaturated surface layer (DSSL) formation with a newly conducted D plasma exposure experiment at E i ∼ 20 eV. It is also further demonstrated, based on new experiments where the ion flux, fluence, and sample temperature are scanned, that non-kinetic (ballistic) processes play a role in the DSSL formation and growth. In addition, the effect of impurities in our plasma is discussed also with a new analysis of the surface composition made after a D plasma exposure.

Nishijima, D. (ORCID:0000000231194827)↗
DOE OSTI2024

The impact of collisionality on the runaway electron avalanche during a tokamak disruption

The exponential growth (avalanching) of runaway electrons (REs) during a tokamak disruption continues to be a large uncertainty in RE modeling. The present work investigates the impact of tokamak geometry on the efficiency of the avalanche mechanism across a broad range of disruption scenarios. It is found that the parameter ν *,crit ⁠, describing the collisionality at the critical energy to run away, delineates how toroidal geometry impacts RE formation. In particular, utilizing a reduced but self-consistent description of plasma power balance, it is shown that for a high-density deuterium-dominated plasma, ν *,crit is robustly less than one, resulting in a substantial decrease in the efficiency of the RE avalanche compared to predictions from slab geometry. In contrast, for plasmas containing a substantial quantity of neon or argon, ν *,crit ≳1, no reduction of the avalanche is observed due to toroidal geometry. This sharp contrast in the impact of low- vs high-Z material results primarily from the relatively strong radiative cooling from high-Z impurities, enabling the plasma to be radiatively pinned at low temperatures and thus large electric fields, even for modest quantities of high-Z material.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Analysis of the inter-species power balance in JET plasmas

Most auxiliary heating methods provide heating to more than one particle species (electrons, ions, impurities) in a fusion plasma. This can lead to substantial temperature differences between species, depending on conditions such as heating power to the different species and collisionality, with temperature differences between species limited by inter-species thermal equipartition and transport. The analysis of the steady-state electron-ion and impurity-ion power balances presented in this paper are used for consistency-checking experimental ion and electron temperature measurements and for inferring the main ion temperature from measured impurity temperatures. As ion temperature measurements by charge exchange spectroscopy (CXS) based on impurity ions have become more difficult and time-consuming since the installation of the ITER-like wall (ILW) with Be and W PFC's, knowing the maximum sustainable temperature difference between ions and electrons, |T i -T e | allows rejecting erroneous measurements. It also obviates the need for an ion temperature measurement, if an electron temperature measurement is available and |T i - T e | cannot be larger than the combined errors of the underlying measurements. A power balance analysis is also required for estimating the errors of the ion and electron heat fluxes prior to any species-resolved transport analysis. The ion-impurity temperature differences are usually found to be small due to strong thermal equipartition between ion species. However, they can approach 10% in JET-ILW low density, high power discharges, such the ones under development for a future JET deuterium–tritium campaign (Joffrin et al 2019 Nucl. Fusion). This has a generally small, but not always negligible effect on the calculation of fusion reaction rates, which depend on main ion temperatures. An important outcome of this analysis is that temperature differences between impurity species are always much smaller than between the impurities and hydrogenic species and can usually be neglected. The article presents two methods for calculating the impurity-to-main ion temperature ratio. Finally, this analysis leads to a method for the reconstruction ion temperature profiles from ion temperature data available at only one or a small number of spatial locations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Energy-angle distribution of the ions in the RF sheath of ICRH antennas

Radio-frequency sheaths forming at the surface of ion cyclotron resonance heating (ICRH) actuators in fusion experiments are associated with enhanced impurity sputtering from ICRH plasma-facing components (PFCs). The minimization of impurity fluxes from the ICRH PFC is a vital task for the usage of ICRH systems. Capturing the ion kinetics is important for an accurate description of plasma–material interactions, because the ion dynamics plays a crucial role in RF sheaths. Here, we present a hybrid particle-in-cell (hPIC) model able to capture the kinetic behavior of the ions. We analyze the kinetic ion energy-angle distributions (IEADs) impacting the RF antenna and its dependence on different plasma and RF sheath parameters. In particular, the IEAD dependence on RF frequency and magnetic field alignment is investigated. Using hPIC, we simulated a case emulating the latest experimental campaign from JET. The simulation showed that under specific plasma and RF parameters, the kinetic motion of ions results in a cusp formation in the IEAD.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2022

Investigation of core impurity transport in DIII-D diverted negative triangularity plasmas

Abstract Tokamak operation at negative triangularity has been shown to offer high energy confinement without the typical disadvantages of edge pedestals (Marinoni et al 2021 Nucl. Fusion 61 116010). In this paper, we examine impurity transport in DIII-D diverted negative triangularity experiments. Analysis of charge exchange recombination spectroscopy reveals flat or hollow carbon density profiles in the core, and impurity confinement times consistently shorter than energy confinement times. Bayesian inferences of impurity transport coefficients based on laser blow-off injections and forward modeling via the Aurora package (Sciortino et al 2021 Plasma Phys. Control. Fusion 63 112001) show core cross-field diffusion to be higher in L-mode than in H-mode. Impurity profile shapes remain flat or hollow in all cases. Inferred radial profiles of diffusion and convection are compared to neoclassical, quasilinear gyrofluid, and nonlinear gyrokinetic simulations. Heat transport is observed to be better captured by reduced turbulence models with respect to particle transport. State-of-the-art gyrokinetic modeling compares favorably with measurements across multiple transport channels. Overall, these results suggest that diverted negative triangularity discharges may offer a path to a highly-radiative L-mode scenario with high core performance.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Erosion and impurity transport for the edge localized mode suppression window in KSTAR

A 3D analysis of plasma wall interactions and global impurity transport for the edge localized mode suppression window in KSTAR during H-Mode discharges has shown that carbon erosion at the divertor plates is a strong function of the resonant magnetic perturbation (n = 1) coil current and relative phasing. The Generalized Perturbation Equilibrium Code was used to determine a realistic initial perturbed plasma equilibrium, and EMC3-EIRENE was used to calculate the resulting scrape-off layer plasma used in this study as a fixed background for the ERO2.0 plasma–material interaction model. The resulting transport leads to deposition of impurities along the targets positioned at the high-field side of the device. An attempt at calculating the resulting effective charge state has demonstrated a similar dependence on the perturbation coil current and has been able to determine a window for the experimentally observed values of Zeff by including contributions of all ionized carbon charge states and deuterium.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Correction and verification of x-ray imaging crystal spectrometer analysis on Wendelstein 7-X through x-ray ray tracing

X-ray ray tracing is used to develop ion-temperature corrections for the analysis of the X-ray Imaging Crystal Spectrometer (XICS) used at Wendelstein 7-X (W7-X) and perform verification on the analysis methods. The XICS is a powerful diagnostic able to measure ion-temperature, electron-temperature, plasma flow, and impurity charge state densities. While these systems are relatively simple in design, accurate characterization of the instrumental response and validation of analysis techniques are difficult to perform experimentally due to the requirement of extended x-ray sources. For this reason, a ray tracing model has been developed that allows characterization of the spectrometer and verification of the analysis methods while fully considering the real geometry of the XICS system and W7-X plasma. Through the use of ray tracing, several important corrections have been found that must be accounted for in order to accurately reconstruct the ion-temperature profiles. The sources of these corrections are described along with their effect on the analyzed profiles. The implemented corrections stem from three effects: (1) effect of sub-pixel intensity distribution during de-curving and spatial binning, (2) effect of sub-pixel intensity distribution during forward model evaluation and generation of residuals, and (3) effect of defocus and spherical aberrations on the instrumental response. Possible improvements to the forward model and analysis procedures are explored, along with a discussion of trade-offs in terms of computational complexity. Finally, the accuracy of the tomographic inversion technique in stellarator geometry is investigated, providing for the first time a verification exercise for inversion accuracy in stellarator geometry and a complete XICS analysis tool-chain.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

The role of B T -dependent flows on W accumulation at the edge of the confined plasma

Abstract 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.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗