Bulletin of the American Physical Society
62nd Annual Meeting of the APS Division of Plasma Physics
Volume 65, Number 11
Monday–Friday, November 9–13, 2020; Remote; Time Zone: Central Standard Time, USA
Session JP19: Poster Session: Magnetic Confinement: Divertor Physics & Plasma-Material Interactions (2:00pm - 5:00pm)On Demand
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JP19.00001: Overview of the preliminary design of the diagnostic suite for MPEX T.M. Biewer, T.S. Bjorholm, J. Rapp The Material Plasma Exposure eXperiment (MPEX) is a planned steady-state device at ORNL that will be used to study plasma-material interactions to advance the progress of engineered materials for the plasma facing components of fusion reactors. The preliminary design of MPEX will soon be reviewed, including the diagnostic suite of instrumentation. Similar to other fusion-relevant devices, MPEX diagnostics will serve a variety of roles: machine protection, basic operation, advanced plasma control, and scientific utilization. The diagnostic suite at the conceptual design stage includes: Thomson scattering, optical emission spectroscopy, interferometers, visible and infra-red camera imaging, pyrometers, microwave diodes, bolometry, in-vessel and ex-vessel thermocouple arrays, pressure gauges, residual gas analysis, and in vacuo surface analysis techniques. MPEX diagnostics will be implemented in a staged approach; Phase I diagnostics are those necessary to meet key performance parameters, while Phase II diagnostics are those necessary for the initial scientific utilization of MPEX. It is envisioned that Phase II diagnostics will be implemented in collaboration with institutions outside of ORNL. This presentation will give an overview of the planned diagnostic layout for MPEX. [Preview Abstract] |
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JP19.00002: Operation of Proto-MPEX with High Magnetic Field Strength in the Helicon Source Region R. H. Goulding, J. F. Caneses, J. B. Caughman, C. H. Lau, E. H. Martin, T. M. Biewer, T. S. Bigelow, J. Rapp, C. J. Beers Several changes to the operating configuration in Proto MPEX have allowed successful production of high density deuterium plasmas at the highest magnetic field strength yet achieved in the helicon region of ~ 0.15 T. The changes include: combining of power from two 13.56 MHz transmitters, achieving a net RF power into the helicon antenna of up to ~180 kW, increasing the diameter of the hole in the skimmer plate downstream from the helicon source, and ramping of the magnetic field in the helicon region from an initial value during plasma breakdown of ~ 0.05 T. With these changes, record deuterium plasma density up to $1.5 \times 10^{20} m^{-3}$ at the target has been achieved, as measured with a double Langmuir probe, and consistent with Stark broadening measurements that have also been made. The highest central/edge heat fluxes for plasmas heated only by the helicon RF were also achieved of ~ 1.8/2.8 $MW-m^{-2}$, simultaneously with an increased plasma diameter at the target of ~ 4 cm. This high-performance operation has been useful for other experiments, for instance it has allowed the demonstration of successful EBW heating of electrons at a plasma density considerably above the cutoff value. [Preview Abstract] |
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JP19.00003: Two-photon Laser Induced Fluorescence Measurements of Neutral Density in the Prototype Material Plasma Exposure eXperiment T.E. Steinberger, J.W. McLaughlin, T.M. Biewer, C.J. Beers, J.B. Caughman, J.F. Caneses, E.E. Scime The development of remotely situated diagnostics for fusion plasmas is necessary since the radiation and electromagnetic-interference (EMI) environment of such devices is becoming ever more intense. Two-photon absorption laser induced fluorescence (TALIF) measurements of neutral velocity distribution functions (NVDF) provide non-perturbative measurements of neutral deuterium temperature, absolute density, and bulk flow in fusion relevant plasmas. Recently, a TALIF system was installed at ORNL on the Prototype Material Plasma Exposure eXperiment (Proto-MPEX). A maximum of 4 mJ of 205 nm light in 8 ns pulses is transmitted over 20 m to measure deuterium NVDFs. Signal is collected confocally and is fiber-coupled back to the laser room. The confocal design requires only one point of optical access-making this design suitable for fusion devices. However, 205 nm light is difficult to transmit through common vacuum components and fiber optics. To explore more compatible options for transmission, a three-photon absorption laser induced fluorescence (3pLIF) scheme is attempted in krypton as a proof-of-principle measurement to increase the injected wavelength. We will present preliminary TALIF deuterium NVDFs and 3pLIF krypton NVDFs. [Preview Abstract] |
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JP19.00004: Integrated model predictions of the impact of He pre-exposures on ITER burning-plasma conditions Ane Lasa, Sophie Blondel, Jon Drobny, Davide Curreli, Jeremy Lore, John M. Canik, Brian D. Wirth The interplay between the most abundant gases (hydrogen isotopes and helium, He) present in fusion plasmas is known to alter fuel retention, especially in tungsten, the leading material candidate for the divertor. However, this interplay is yet to be sufficiently understood or characterized in order to confidently project fuel retention levels to future fusion devices. Here, we present a series of integrated simulations that include surface erosion, ion implantation, and sub-surface gas dynamics and recycling; and apply them to modeling the exposure of a tungsten target pre-damaged by He plasma to 100 MW burning plasma conditions in ITER. We have selected multiple scenarios for the He pre-exposure; including predictions of sub-surface damage resulting from early ITER He-operation as well as calculations of higher-fluence He plasma exposures in linear devices. Initial results, focused on the outcome of gas implantation, dynamics and recycling, indicate that under fluxes sufficient for He bubbles to nucleate in the near-surface, these clusters will locally increase the hydrogenic retention, but decrease the hydrogen species permeation. Our results also predict longer term ITER behavior. [Preview Abstract] |
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JP19.00005: High-performance PIC-BCA code for fusion-relevant plasma-material interactions Jon Drobny, Davide Curreli Plasma-material interactions (PMI) are among the most difficult fusion-relevant phenomena to model. Reduced models of ion reflection, target sputtering, or redeposition are commonly used. However, reduced models neglect details such as the significant high-energy tails of ion energy-angle distributions at plasma facing component surfaces, leading to an underestimation of total sputtering and loss of information concerning the trajectories of emitted particles. Adequately modeling PMI requires kinetic models of the plasma-material interface to capture the behavior of the plasma sheath, the material, and their response to each other. To that end, we have produced a tightly coupled, high-performance particle-in-cell (PIC) and binary collision approximation (BCA) code with two components; hPIC, which has been recently upgraded to include RF effects and a nonuniform mesh to simulate large plasma domains beyond the surface, and Rustbca, which is a newly developed code benchmarked against previous codes but with many advantages, such as modern code design and high performance. Using this form of the PIC-BCA, we have analyzed the equilibrium PMI along an ITER-like target including redeposition fractions, energy-angle distributions, and calculations of net and gross erosion. [Preview Abstract] |
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JP19.00006: Modeling of W-based High Entropy Alloys Proposed as PFM for Fusion Reactors Muhammad Abdelghany, Jean Paul Allain High-Entropy Alloys (HEAs) have great potential to be employed in the extreme environments of nuclear systems. However, we have very little knowledge about the occurrence, structure, and properties of their crystalline phases. One way of filling this gap is by predicting their behavior computationally. In this study, we are employing some methods to computationally develop an interatomic potential for a W-based HEA proposed by our collaborators at LANL as a PFM for fusion reactors. The proposed alloy is composed of W-Ta-Cr-V which is already tested experimentally and showed promising behaviors under irradiation. The first method is based on using a genetic-algorithm that combines the single-element potentials to generate binary potentials which are optimized such that the MD calculations of selected surface properties converge to the DFT calculations of these properties. The second method employs machine learning to connect the first principle calculations with MD calculations to produce a potential that works well for both the bulk and surface. BCA calculations were also performed for this alloy to capture the mixing effect in response to ion bombardment. For that, DYNAMIX code was used to calculate the partial scattering yield, surface, and depth profiles under ion irradiation. [Preview Abstract] |
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JP19.00007: Analysis of Plasma Contamination via Plasma-Surface Interactions from Neutral Beam Shine-Through Matthew Parsons, Hanna Schamis When high-energy neutral beam particles are not fully absorbed by the plasma, they will collide with plasma-facing components (PFCs). This beam shine-through occurs strongly at low plasma densities, which is particularly relevant to the development of non-inductive start-up scenarios, but will also occur to a smaller degree even during normal plasma operation. Typical analysis of the beam shine-through looks only at the thermomechanical impact of the power deposited on the PFCs. In this work, we instead examine the interaction between the beam and the surface of the wall from a particle perspective by using the well-known TRIM software. In particular, we calculate sputtering and reflection coefficients for a range of deuterium beam energies, beam geometries, and PFC materials, and assess the extent to which these processes result in (1) contamination of the main plasma by sputtered impurities and (2) modification of the heating profile due to particle reflection in a variety of realistic operating scenarios. [Preview Abstract] |
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JP19.00008: Hybrid Illinois Device for Research and Applications Material Analysis Test-stand (HIDRA-MAT) Andrew Shone, Alfonso De Castro, Zachariah Koyn, Aveek Kapat, Rabel Rizkallah, Daniel O'Dea, Jean-Paul Allain, Daniel Andruczyk HIDRA is a toroidal plasma fusion device located at UIUC. Steady state plasmas and high particle fluxes on the order of 10$^{\mathrm{22}}$ m$^{\mathrm{-2}}$s$^{\mathrm{-1}}$ provide a unique test-bed for materials research. HIDRA-MAT is an extension of HIDRA that acts as a material test-stand capable of performing plasma exposure and surface characterization \textit{in-vacuo}. Removing the material surface degradation that occurs from oxidation when exposed to atmosphere produces data that more accurately depicts the PMI mechanisms taking place. HIDRA-MAT's material preparation systems include a rotatable UHV sample heater and liquid metal (LM) droplet injector. Sample surface characterization in-between and after plasma exposures is achieved by dual-pulsed LIBS and TDS. In addition to the surface characterization diagnostics, HIDRA-MAT's dual RGA system allows for deuterium and helium differentiation. Preliminary testing of each of HIDRA-MAT's sample preparation systems and diagnostics have been conducted. LM droplet size and sample temperature can be controlled during application. Results will be shown that highlight HIDRA-MAT's ability to investigate PMI mechanisms associated with LM PFCs retention of deuterium and helium. [Preview Abstract] |
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JP19.00009: Design of an Experiment to Investigate the MHD Response of Liquid Metal to Pulsed Currents and Magnetic Fields Daniel P. Weber, Colin S. Adams We present plans for an experimental campaign to examine the MHD response of a liquid metal free surface in scenarios where large-amplitude waves are induced by a nonuniform magnetic field ($<$40~T) parallel to the free surface. An apparatus has been designed in which electrical energy stored in an LC pulse-forming network (PFN) is passed through coaxial conductors to a wire suspended between a solid electrode and a pool of liquid metal. The PFN is designed to deliver a current pulse of up to 100~kA through suspended wires with radii ranging from 1--3~mm for a duration of roughly 5~$\mu$s. A vacuum power feedthrough transmits the pulse to the suspended wire. The PFN circuit was simulated using LTSpice and design iteration resulted in a waveform with less than 55\% predicted load current reversal. The first experiments will be performed with liquid tin due to its low vapor pressure and melting point. [Preview Abstract] |
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JP19.00010: In situ mass spectrometry measurements of erosion and hydrogen scattering from lithium surfaces during irradiation with ultralow-energy ion beams Shota Abe, Bruce E. Koel We have conducted experimental in situ measurements of the sputtered and reflected species from thin Li film targets during irradiation by a mass-selected deuterium ion beam to study erosion and hydrogen retention behavior. The experiments include ions incident at ultralow energies (\textless 100 eV), which is relevant to tokamak divertor environments. A quadrupole mass spectrometer (QMS) is used to detect positive and negative ions and neutral species, and is calibrated by referencing available sputtering and secondary ion yield data. Identifying and quantifying both the neutral and ion species produced by sputtering and reflection is necessary to advance our understanding of plasma and materials physics in tokamaks that employ Li-coated first wall concepts such as in LTX-$\beta $ and NSTX-U at PPPL. In preliminary results, we measured Li species from Li films irradiated by D$^{\mathrm{+}}$ or D$_{\mathrm{2}}^{\mathrm{+}}$ beams with a flux of \textasciitilde 10$^{\mathrm{9}}$ cm$^{\mathrm{-2}}$s$^{\mathrm{-1}}$ for incident ion energies of 5-400 eV/D while keeping the chamber background pressure low so that volumetric reactions do not occur. [Preview Abstract] |
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JP19.00011: Design and Initial Results from the Dynamic Lithium Corrosion Test Bed Cody Moynihan, Steven Stemmley, Alfonso de Castro, Joerg Zimmermann, David Ruzic Liquid metals as plasma-facing components (PFCs) present the opportunity to overcome some of the drawbacks of solid high-Z refractory PFCs, such as irreversible erosion and melting during plasma disruptions. However, it is well know that liquid metals can be highly corrosive to many common structural materials. While static corrosion studies have been performed in the past, there is little known about the effects of surface shear and tensile stress on the corrosion of structural materials submerged in liquid metals. With the support of General Fusion, a dynamic liquid metal corrosion test bed has been developed at the Center for Plasma Material Interactions, allowing rotation and tensile stress to be applied to submerged metallic samples. The design and testing of the experimental setup are presented, along with initial results of the corrosion studies. Initial testing involves rotation of metallic samples in a bath of liquid lithium at 300$^\circ C$ for 100 hours at a rotation speed of 50 Hz. Mechanical and surface properties of the materials are characterized through the use of tensile strength testing and surface imaging and compared to samples exposed to rotation in the liquid metal allowing observation of both macroscopic and microscopic changes. [Preview Abstract] |
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JP19.00012: Advanced Flowing Liquid Lithium Divertor Plate Design Steven Stemmley, Cody Moynihan, Matt Szott, Alfonso de Castro, Patrick Bunting, Peter F. Buxton, Daniel Iglesias, David Ruzic Liquid lithium has shown promise as a plasma facing material because of its ability to handle high heat loads and particle fluxes as well as improve plasma performance. In the past, the Liquid Metal Infused Trench (LiMIT) concept, working solely via thermoelectric magnetohydrodynamics (TEMHD), has been shown to work well under fusion relevant conditions. However, under intense, local heat flux, lithium can be accelerated quickly and expose the underlying substructure, called dryout. Recently, alternative geometries such as posts and large pore foam geometries have been shown to mitigate liquid lithium dryout at high heat fluxes. For the large pore foam, no depression in the lithium surface was observed for an electron beam heat flux of 6.8 MW/m$^2$. These new structures are being incorporated into a divertor plate design with a fully flowing liquid lithium loop for the ST40 tokamak. This design involves the modification of an existing divertor tile to include these advanced geometries as well as the associated liquid metal pumps, piping, and reservoir system. Results and discussion of the advanced geometry testing and the preliminary design for the liquid lithium loop will be presented. [Preview Abstract] |
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JP19.00013: Abstract Withdrawn
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JP19.00014: Experimental Understanding of Heat Flux Dissipation During Plasma Detachment in the Small Angle Slot Divertor of DIII-D Jun Ren, David Donovan, Jon Watkins, Huiqian Wang, Dan Thomas, Rejean Boivin In DIII-D, an array of surface eroding thermocouples (SETCs) were installed in Small Angle Slot (SAS) divertor to measure the heat flux in a slot-like divertor during detachment. In both toroidal field directions, the peak heat flux measured at the outer strike point was reduced by \textasciitilde 50{\%} when the plasma reached the detachment regime (obtained by ramping up the plasma density). When the ion Bx$\nabla $B drift direction was toward the SAS divertor, the heat flux measured by SETCs first increased as plasma density increased, then started to roll over when the line-averaged density approached 6x10$^{\mathrm{19}}$m$^{\mathrm{-3}}$. In contrast, when the ion Bx$\nabla $B drift direction was away from SAS, the heat flux began to decrease at a lower plasma density (\textasciitilde 4x10$^{\mathrm{19}}$m$^{\mathrm{-3}})$, indicating the onset of plasma detachment, and continued to decrease with further increases of plasma density. These significant differences in the onset of detachment in different B$_{\mathrm{T}}$ directions are believed to be largely determined by the ExB drift. Because ITER's divertor must be operated with some degree of plasma detachment to radiate most of the power arriving in the scrape-off-layer, it is crucial to further understand of the impact of magnetic drifts on plasma detachment behavior. [Preview Abstract] |
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JP19.00015: Analysis of the Impact of Divertor Closure on Detachment in DIII-D Jerome Guterl Several experiments have been recently conducted in DIII-D to investigate effects of divertor closure on plasma detachment to improve power dissipation capabilities of tokamak divertors [1]. We present here the modeling and analysis of plasma detachment in closed divertor geometry in DIII-D. Effects of divertor closure on plasma detachment are quantified through the critical ratio of the upstream plasma pressure over the heat flux entering the recycling region which determines the threshold for plasma detachment [2]. To that end, DIII-D boundary plasma are modeled using the 2D boundary plasma transport code UEDGE. ExB drifts have been recently shown to play a critical role in detachment of divertor plasma [3]. Synergistic effects of ExB drifts and divertor closure on plasma detachment are thus examined, taking advantage of the numerical robustness of UEDGE against ExB drifts. In this framework, neutral recycling is modeled with the neutral fluid model embedded in UEDGE to exploit fast convergence of UEDGE simulations permitted by a fully implicit time scheme. The range of applicability of this neutral fluid model in simulations of plasma detachment in DIII-D divertor is estimated using neutral distributions obtained from the kinetic neutral code EIRENE. [1] Guo, H. Y., et al. \textit{Nuclear Fusion} 59.8 (2019): 086054. [2] Pshenov, A. A., et al. \textit{Nuclear Fusion}, \textit{59}(10), 106025 [3] Jaervinen, A. E., et al. \textit{Physical review letters} 121.7 (2018): 075001. [Preview Abstract] |
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JP19.00016: Infrared Constrained Equilibrium Reconstruction and Application to Snowflake Divertor Studies J.T. Wai, P.J. Vail, A.O. Nelson, Z.A. Xing, C. Lasnier, T.K. Gray, E. Kolemen One of the challenges of the snowflake divertor (SFD) is a reliable means of reconstructing the magnetic field geometry in the divertor, due to the shallow flux gradients associated with multiple field nulls. We have developed a technique to improve SFD reconstruction using heat flux measured by the divertor InfraRed (IR)TV diagnostic. This technique identifies the spatial position of the two SFD X-points using characteristics of the heat flux, such as power distribution among the peaks and the fitted strike point positions. The algorithm to find X-points converges quickly and is amenable to real-time control. Using this method, a set of inferred X-point positions are used as a constraint to create new equilibria. Self-consistency is verified by using an analytic SFD heat flux model [1] to show that the X-point-constrained equilibria do indeed match the measured heat flux. The set of X-point-constrained equilibria are compared to kinetic equilibria, and we discuss differences in the pedestal and scrape-off layer. In particular, preliminary analysis shows that X-point-constrained equilibria predict 10{\%} variation in the edge current levels versus kinetic equilibria without X-point constraints. [1] Vail, NME. 516-523 (2019) 9 [Preview Abstract] |
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JP19.00017: Simulations of divertor plasmas with inverse sheaths Rebecca Masline, Roman Smirnov, Sergei Krasheninnikov The effect of strong electron emission from material surfaces has been proposed to form an “inverse sheath”: a region with positive potential relative to the plasma edge which prevents the flow of ions to the wall. We assess the viability of this regime in a tokamak device using the 2D edge plasma transport code UEDGE. Since the UEDGE code does not consider the sheath region directly, we apply boundary conditions at the divertor targets which emulate the physics of both "standard" and "inverse" sheath regimes. Using these boundary conditions, we perform scoping studies to assess plasma parameters near the target by varying the density at the core-edge interface. We observe a smooth transition in the resultant profiles of plasma parameters for the standard sheath, and a bifurcation across the simulation set for plasmas with an inverse sheath. The cause of this bifurcation is assessed by performing the parameter scan both with and without impurity radiation; we observe that the bifurcation persists in both cases, indicating that this bifurcation is caused by plasma recombination. [Preview Abstract] |
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JP19.00018: Computational Analysis of Heat Deposition Pattern Dependence on RMP Phasing for KSTAR Jonathan Van Blarcum, Heinke Frerichs, Oliver Schmitz, Jong-Kyu Park, SeongMoo Yang Resonant Magnetic Perturbations (RMP) will be used on ITER in the effort of suppressing Edge Localized Modes (ELM). A predictive model for ELM control was validated on the Korean Superconducting Tokamak for Advanced Research (KSTAR) defining a 'window' of RMP configurations that offer suppression, [J.-K. Park et al., Nature Physics 14 (2018)]. This work explores variation of the power deposition onto diverter targets relative to RMP phasing throughout the ELM suppression window. A magnetic footprint analysis was performed using Field Line Analysis and Reconstruction Environment (FLARE) which showed significant variation in the diverter's connection to inner plasma regions relative to RMP phasing, suggesting similar variation in the associated heat deposition. EMC3-EIRENE, a Monte Carlo fluid plasma edge model, is then used to model the heat and particle loads on the diverter, including effects from plasma response, modeled by Generalized Perturbed Equilibrium Code (GPEC). In this way the variation of the diverter heat flux due to RMP phasing along the KSTAR ELM suppression window is computationally determined and a minimal heat flux RMP configuration is predicted. [Preview Abstract] |
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JP19.00019: Modeling of dynamic plasma-wall coupling during series of ELM pulses. Roman Smirnov, Sergei Krasheninnikov, Maxim Umansky Plasma recycling on material walls plays a crucial role in edge plasma transport in tokamaks. Recycling depends on both plasma and wall conditions, which are mutually dependent. However, in most edge plasma modeling studies a fixed recycling coefficient is used. This simplified approach can be insufficient when transient plasma processes, e.g. ELMs, are considered. These processes can strongly perturb plasma and wall states triggering complex plasma-wall dynamics. In this work we model dynamics of edge plasma and hydrogen retention/outgassing from divertor targets during series of ELM pulses. The modeling is performed using edge plasma transport code UEDGE and wall code FACE coupled in a fully time-dependent manner. Various initial divertor plasma conditions and ELM characteristics are simulated. The obtained results demonstrate that ELMs can cause release of significant amounts of hydrogen from the target material, triggering plasma detachment in the divertor. The simulated recycling dynamics demonstrates complex response of wall outgassing to plasma perturbations, which depends on the hydrogen density profiles in the material, with longer response times corresponding to deeper material layers. [Preview Abstract] |
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JP19.00020: Role of ion temperature anisotropy in 2D edge plasma transport. Menglong Zhao, T.D. Rognlien, Ilon Joseph A model of ion temperature anisotropy for transport in the tokamak edge/scrape-off-layer (SOL) is implemented in the 2D UEDGE fluid code. This is a generalization of the 1D work reported in [1-3]. The ion temperature is decomposed into two components: $T_{i\|}$ for the temperature parallel to the magnetic field, ${\bf B}$, and $T_{i\perp}$ for that perpendicular to ${\bf B}$. This anisotropy modifies the parallel viscosity expression, gives an additional force to the ion parallel momentum equation, and changes ion heat flux in the two directions. From previous kinetic simulations, e.g. using XG1a and COGENT codes, evidence for a strong ion temperature anisotropy are shown in hot SOL plasmas when collisional equilibration between $T_{i\|}$ and $T_{i\perp}$ is slow. Detailed impact of this anisotropy on 2D plasma profiles and flows is presented for single-null divertor geometry, including the effect of cross-field drifts and recycled neutral gas. Refs: [1] Z. Guo, X-Z. Tang, Phys. Plasmas 19 (2012) 082310; [2] S. Togo et al., J. Comp. Physics 310 (2016) 109; [3] T.D. Rognlien, T.A. Brengle, Phys. Fluids 24 (1981) 871. [Preview Abstract] |
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