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 GO13: Magnetic Confinement: Plasma-Material InteractionsLive
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Chair: Tyler Abrams, GA |
Tuesday, November 10, 2020 9:30AM - 9:42AM Live |
GO13.00001: Understanding near and far-SOL impurity spatial distributions in DIII-D by reproduction of collector probe deposition patterns with DIVIMP and 3DLIM S. A. Zamperini, D. C. Donovan, J. D. Duran, J. D. Elder, J. N. Nichols, P. C. Stangeby, E. A. Unterberg Replication of tungsten deposition patterns on a collector probe near the outboard midplane (OMP) of DIII-D in 3DLIM simulations involves W entering the far-SOL from the inner target direction in unfavorable Bt direction (grad-B drift away from divertor), despite the W source originating from the outer target. This is explicable by a long-hypothesized near-SOL impurity accumulation effect via the Ti gradient force. Mach probes on several tokamaks show main ion flow (vi) stagnation near the plasma crown in unfavorable Bt. However, stagnation between the OMP and outer target is commonly observed in favorable Bt, with fast inner-target-directed flow in the crown. In DIVIMP, the ad hoc addition of fast inner-target vi "flushes out" W ions that otherwise accumulate, implying that accumulation may only occur in unfavorable Bt. The impurity distribution along a far-SOL flux tube is input to 3DLIM, a new far-SOL Monte Carlo 3D impurity transport code. All probes inserted for unfavorable Bt had more W on their inner target facing (ITF) sides compared to their outer target facing (OTF) sides (ITF/OTF\textgreater 1), which is reproduced in 3DLIM only when W enters the far-SOL from the ITF direction, as expected for near-SOL accumulation, and vice-versa for favorable Bt. These results give insight into what SOL phenomena may be setting the boundary condition on core W content. [Preview Abstract] |
Tuesday, November 10, 2020 9:42AM - 10:06AM Live |
GO13.00002: Ex Situ Characterization of Digital Holography in Preparation for Demonstration of In Situ Plasma-Induced Erosion Measurement (PhD Oral-24) Cary Smith, Theodore Biewer, Trey Gebhart, Elizabeth Lindquist, C. E. Thomas Digital holography has been proposed for \textit{in situ} measurement of plasma facing component erosion. Progress on the development of a dual laser digital holography erosion diagnostic will be presented. The diagnostic images a surface region 1 centimeter in diameter and is capable of single or dual wavelength operation, with the ability to measure surface height changes of up to 4.5 microns in single laser mode and up to 2 millimeters in dual laser mode. Characterization results to be presented include \textit{ex situ} measurements of plasma-eroded targets exposed to an electrothermal (ET) arc source with detected erosion of 150 nanometers/exposure. Measurements of a moving target, intended to simulate dynamic surface change detection, will also be shown; translations from 60 nanometers to 1 micron were recorded in single laser mode and from 50 to 380 microns in dual laser mode. The accuracy of the measurements will be compared with validation data and discussed in the context of image quality and noise levels. The characterization work indicates that coupling digital holography with the ET-arc for \textit{in situ} erosion measurement can achieve a successful result, and the experiment plan for an \textit{in situ} demonstration will be presented. [Preview Abstract] |
Tuesday, November 10, 2020 10:06AM - 10:18AM Live |
GO13.00003: In-situ measurements of surface structure and composition evolution during high-flux plasma exposure Robert Kolasinski, Chun-Shang Wong, Ryan Hood, Josh Whaley In-situ analysis of surfaces during high-flux plasma exposure represents a long-standing challenge in the study of plasma-material interactions. While post-mortem microscopy can provide excellent detail on how the material itself has been modified, in-situ techniques can provide information on dynamic effects. In this study, we demonstrate how spectroscopic ellipsometry and low energy ion beam analysis can be applied to real-time characterization of surfaces during low energy, high-flux He and D implantation. Here we report on three applications: (a) growth of He-induced nanostructure on W plasma-facing surfaces, (b) effects of N and B impurities on W surface morphology and composition, and (c) removal of oxides from super-permeable membranes used for pumping and separation of H isotopes from plasma exhaust. In-situ spectroscopic ellipsometry was found to be extremely sensitive to nm-scale changes surface morphology induced by plasma exposure, capturing the early stages of W nanostructure growth and providing insight into the erosion of thin oxides. Low energy ion beam analysis was used to complement these results by providing information on the composition of the first few nm of the exposed interface. [Preview Abstract] |
Tuesday, November 10, 2020 10:18AM - 10:30AM Live |
GO13.00004: Status and Progress of the Domestic Liquid Metal Plasma-Facing Component Design Program R. Maingi, A. Khodak, E. Kolemen, C.E. Kessel, J. Jun, J.D. Lore, B.A. Pint, D.L. Youchison, S. Smolentsev, D. Andruczyk, D. Curreli A new domestic liquid metal plasma-facing component (PFC) design program was initiated in FY2020. We are examining a flowing liquid lithium divertor design for a well-documented Fusion Nuclear Science Facility, which was the subject of a prior liquid metal PFC evaluation [C.E. Kessel et al., \textit{Fusion Sci. Techn}. \textbf{75} (2019) 886]. Presently we are evaluating the heat flux exhaust capability as a function of flow speed, initially at the maximum Li temperature about 450 $^{\mathrm{o}}$C, i.e. below the evaporative limit. The design calculations include liquid lithium magneto-hydrodynamic flow via computational fluid dynamics, and plasma response to the liquid lithium PFCs including SOLPS scrape-off layer and divertor plasma, and kinetic sheath calculations. This activity includes experiments on liquid lithium flow and material compatibility characteristics, such as wetting, dryout, corrosion, erosion, embrittlement, etc. Experimental facilities at ORNL, PPPL, and UI-UC are used for these studies. An overview of the PFC design calculations and the supporting experimental studies, some of which are used to validate the models, will be presented. [Preview Abstract] |
Tuesday, November 10, 2020 10:30AM - 10:42AM Live |
GO13.00005: Review of single effect experiments for the National LM PFC Program Daniel Andruczyk, Rabel Rizkallah, Daniel O'Dea, Davide Curreli, Rajesh Maingi, Chuck Kessel, Sergey Smolentsev In 2019 the DOE tasked PPPL, ORNL UIUC and UCLA to begin a National LM PFC program tasked with designing a flowing liquid metal divertor system for future fusion reactors. This was a follow on from the FESS LM PFC evaluation study led by Kessel \textit{et al., }[Fusion Sci. Tech. 75 (2019) 886]. Part of supporting the design effort, UIUC is performing single effect experiments on liquid lithium material compatibility (wetting, dryout, corrosion), flow characteristics of different concepts (fast, medium, slow, LiMIT, FLiLi, CPS) and vapor shielding (seeded environment, modeling). Though, initially not directly part of the national program, recent results performed on EAST at ASIPP in China with a LiMIT type limiter has been giving insight into PFC design. In July of this year vapor shielding experiments will be performed on MAGNUM-PSI at DIFFER. This presentation will give a brief overview of some of the results obtained so far and also will talk about future measurements that will take place in HIDRA and a new facility in Illinois, MEME, to support the ongoing design efforts. This work supported in part by the U.S. Department of Energy under grant number DOE DE-SC0020642. [Preview Abstract] |
Tuesday, November 10, 2020 10:42AM - 10:54AM Live |
GO13.00006: Divertor Closure Effects on a Lithium Vapor Box Eric Emdee, Robert Goldston The SOL heat flux in a fusion reactor is predicted to be beyond what any solid, attached divertor could handle. Detached divertors generally succeed in reducing the heatflux to the target but have difficulty preventing gaseous impurities from reaching the main plasma. The goal is then to cause controlled divertor detachment; localizing radiation and impurities while maintaining large heat flux reductions. The lithium vapor box is a detached divertor design that seeks to control the detachment front via differential pumping. By evaporating lithium close to the divertor target and placing condensing surfaces between the X-point and the target, a vapor density gradient can be created resulting in natural feedback control that impurity gas puffing struggles to obtain. Here we present SOLPS-ITER simulations of NSTX-U with a hypothetical lithium vapor box divertor. In previous work, the addition of a fueling gas puff has been shown to drastically reduce upstream contamination via the ion frictional force acting on the lithium impurities. The effect of divertor closure on a lithium vapor box will be examined, as well as evaporation location. Lithium upstream fractions as well as redeposition of lithium vapor in the divertor region will be primary metrics of performance. [Preview Abstract] |
Tuesday, November 10, 2020 10:54AM - 11:06AM Live |
GO13.00007: Results from a lithium vapor box test stand at PPPL and B2.5-Eunomia simulations for a proposed Magnum-PSI lithium vapor box experiment J. A. Schwartz, R. J. Goldston The lithium vapor box is a concept for a divertor designed to handle the extreme heat fluxes generated in future fusion reactors. Within a slot lined with capillary-porous materials, Li vapor induces plasma detachment by cooling the divertor plasma until it volumetrically recombines. Strong differential pumping via condensation localizes the Li vapor. Two linear-geometry experiments study elements of the physics involved. A test stand at PPPL studies the evaporation, flow, and condensation of Li vapor in three 5\,cm long, 5\,cm diameter cylindrical boxes without plasma, as would occur between shots in near-term experiments. By using a Direct Simulation Monte Carlo code we were able to reproduce the measured value of mass transferred during the experiments to within~$\pm$15\%, and demonstrate the expected differential pumping. A second, proposed for the linear plasma divertor simulator Magnum-PSI, studies the interaction of a 4e20\,m$^{-3}$, 1.5\,eV, 1\,cm radius plasma beam with a 16\,cm long Li vapor cloud. In simulations with B2.5-Eunomia, a 12\,Pa vapor cloud from a 625$\,^\circ$C liquid Li surface reduces the plasma pressure at the target by a factor of 15 and the heat flux there from 3.7\,MW\,m$^{-2}$ to 0.13\,MW\,m$^{-2}$; the power is dissipated within the vapor box. [Preview Abstract] |
Tuesday, November 10, 2020 11:06AM - 11:18AM Live |
GO13.00008: Plasma-wall interaction on the SLiC spherical tokamak device with large-area, dynamic liquid lithium free surface Stephen Howard, A. Mossman, W. Zawalski, D. Froese General Fusion is developing a magnetized target fusion system in which a spherical tokamak plasma target is injected by a magnetized Marshall gun into a flux conserver consisting of a liquid lithium vortex. A compression system will then collapse the cavity to compress and heat the target plasma to fusion conditions. We have recently commissioned a subscale experiment called SLiC (Spector Lithium Configuration) as way to de-risk both the engineering and the confinement physics in the situation where a moderately hot magnetized plasma must interact with a large-area free surface of liquid lithium. SLiC is similar in design to the sequence of compact, high-performing spherical tokamak SPECTOR devices that have been in operation at GF since 2016. In the first phase of experiments, ST plasmas were injected into a well-instrumented solid metal flux conserver with an annular liquid lithium puddle on the bottom of the vessel that had an angular coverage of 7 to 28 degrees (spherical polar angle) depending on puddle fill depth. A second phase of experiments, aiming to approach hemispherical coverage of liquid lithium, has extended the angular coverage to 78 degrees (polar angle) starting at the outboard equator going downward. This is done by applying an early current pulse that propels the liquid puddle outward to~coat the flux conserver up to the equator in advance of the plasma formation. Interactions between the plasma and liquid free surface are studied with fast camera video and standard plasma diagnostics and used to validate corresponding MHD-CFD simulations. [Preview Abstract] |
Tuesday, November 10, 2020 11:18AM - 11:30AM Live |
GO13.00009: Impact of transient dust injection on power load in impurity seeded divertor. Roman Smirnov, Sergei Krasheninnikov Dust can present significant challenges for operation of long pulse discharges in future fusion reactors. It is expected that large quantities of sputtered wall material can be deposited at the bottom of a reactor vessel. Formation of dust from such deposits and its subsequent mobilization by transient plasma events can present substantial intermittent source of plasma impurities. On the other hand, seeding of gaseous impurities is expected to be necessary for divertor power load mitigation. In the present work we investigate effects that transient dust sources can have on plasma and heat transport in a divertor with other seeded impurities. We conducted self-consistent time-dependent simulations of coupled dust and plasma dynamics following tungsten dust outburst in ITER-like divertor with seeded neon impurities using coupled DUSTT and UEDGE codes. The evolution of the edge, SOL, and divertor plasma conditions, following the injection of tungsten dust of various quantities and sizes at the outer strike point, was investigated. The estimates of the core plasma impurity fraction and the core impurity screening factor associated with the dust injection were also obtained. The simulations showed that injection of even modest amounts of tungsten dust into the divertor plasma can result in large transient increase of the divertor power load and intolerable levels of the core plasma contamination with tungsten impurities. [Preview Abstract] |
Tuesday, November 10, 2020 11:30AM - 11:42AM Live |
GO13.00010: Effects of Inverse Sheath Formation at Divertor Plates, Dust Grains, and Negative Ion Sources, on Tokamaks Michael Campanell, Grant Johnson Recent simulations [1] show that surfaces with strong electron emission form inverse sheaths, which differ from the conventional sheaths assumed in the tokamak literature [2]. We offer analytical calculations predicting that for the same SOL input power that leads to a tens-or-hundreds-of-eV target plasma in a conventional sheath operating regime, a few-eV or sub-eV target plasma will form if there is an inverse sheath, possibly facilitating detachment [3]. Other advantages of inverse sheaths would include reduction of ion impact energies (minimizing sputtering) and prevention of arcs. A recent paper by Masline et al. [4] offers the first fluid modelling of divertor plasmas with inverse target sheaths. Besides the target plates, inverse sheaths could form at hot tungsten dust grains, with implications on grain lifetime. Zhang et al. [5] found that inverse sheaths can also form under negative H emission and induce ion-ion plasmas in negative ion sources for tokamak neutral beam heating. [1] G. R. Johnson and M. D. Campanell, Plasma Phys. Rep. 45, 69 (2019) and PRL 122, 015003 (2019). [2] S. Takamura et al., CtPP 44, 126 (2004). [3] M.D. Campanell, PoP 27, 042511 (2020). [4] R. Masline et al., CtPP (2019) DOI: 10.1002/ctpp.201900097 [5] Z. Zhang et al., PSST 27, 06LT01 (2018). [Preview Abstract] |
Tuesday, November 10, 2020 11:42AM - 11:54AM Live |
GO13.00011: Tokamak fusion kidnapped by PSI Leonid Zakharov Plasma surface interaction (PSI) was always an important part of magnetic fiusion. Ater lack of success with $Q_{DT}=1$ on TFTR and JET in the 1990s, the role of PSI was dramatically enhanced. The tungsten divertor, with a non-sense requirement of cooling plasma down to 20 eV, became dominant in the research. In fact, the crazy sounding question ``Does tokamak fusion needs PSI ?'', has a definite answer: It does not. During last 2 decades PSI suggested nothing useful for progress but made tokamak hostage. The failure of 1990s was in insufficient energy confinement. The fusion relevant confinement requires high edge plasma temperature, which makes PSI inapplicable. The mean free path $L$ of plasma particles is $L_m\simeq100\cdot T^2_{keV}/n_{20}$. With $n_{20}\simeq0.1$ and $T_{keV}=1$, $L$ is already 1 km and is 4 km for 2 keV. The Scrape Off Layer becomes collisionless. Above 1 keV, SoL is not the PSI. It is a flux of independent particles in a self-established electric potential (which is much lower that 3-5 $T_e$ of PSI) with predictable and simple physics unlike the mess of PSI. Instead of PSI ``ideas'', the only acceptable material surface is 24/7-Flowing Liquid Lithium for low (0.5) recycling high performance regime compatible with burning plasma. [Preview Abstract] |
Tuesday, November 10, 2020 11:54AM - 12:06PM Live |
GO13.00012: Effects of erosion/redeposition of mixed-material on deuterium recycling during L- and H-mode plasmas in DIII-D. Tatyana Sizyuk, Ahmed Hassanein, Jeffrey Brooks, Tyler Abrams, Houyang Guo The effects of plasma parameters on material mixing and deuterium recycling were studied using self-consistent integrated models implemented in ITMC-DYN and REDEP/WBC computer packages benchmarked with results of RBS and TDS sample analyses. We simulated W erosion and redeposition and calibrated our results with RBS data for two types of discharges in DIII-D using the DiMES probe. The probe surface was initially covered by C with several W spots on the surface. Using the reconstructed edge plasma characteristics from OEDGE/DIVIMP, our simulations successfully explained the factor of 10 difference in W net erosion between high-temperature, low-density and low-temperature, high-density divertor plasma conditions and showed significant W redeposition on 15 mm spots, up to 75{\%}, while only 15{\%} of W was redeposited on 1 mm spots. This analysis was further extended to time-dependent simulations of the D, C, and W ion interactions (i.e., erosion/redeposition) on the evolving surfaces. Our simulations calculated the C/D relative concentrations in plasma near the divertor surface and in co-deposited layers. Small amount of C impurities in plasma (3{\%}) can significantly increase D retention (2 times) due to the formation of C/W surface layer compared to pure W. [Preview Abstract] |
Tuesday, November 10, 2020 12:06PM - 12:18PM |
GO13.00013: Heat Handling Capabilities of LiMIT under Fusion Relevant Heat Fluxes Daniel O'Dea, Rajesh Maingi, Zhen Sun, Alfonso De Castro, Rabel Rizkallah, Steven Stemmley, Daniel Andruczyk Liquid Metal (LM) PFCs provide an alternative to current solid PFCs. Flowing LM components present a constantly refreshing surface to the plasma enabling particle handling, heat removal and potential T/D recovery for re-injection. The current popular choice for these PFCs is Li due to its low recycling and low-Z. Low recycling greatly reduces the proportion of neutral H and impurities sputtered into the plasma improving the energy confinement. To further research into this field a LiMIT-type limiter was deployed in EAST allowing for the temperature response of the plate under fusion relevant heat fluxes (\textgreater 1MW/m$^{\mathrm{2}})$ to be examined. Temperature profiles on the plate are measured by TCs embedded beneath the surface of the plate, these measurements combined with post-mortem analysis of the plate provide insights into the plate performance during plasma exposure. The LiMIT concept utilizes TEMHD to produce lithium flow along the surface and is cooled by He gas. To supplement the experiment, heat transfer simulations are being done in COMSOL in an attempt to elucidate the major mechanisms controlling the heat transfer upon the plate and aide improvements in the heat handling capabilities of future LM PFCs. [Preview Abstract] |
Tuesday, November 10, 2020 12:18PM - 12:30PM Not Participating |
GO13.00014: Influence of tin-lithium alloy composition on basic liquid plasma facing material characteristics Alfonso de Castro Calles, Cody Moynihan, Steven Stemmley, Daniel Andruczyk, David Ruzic Lithium is considered the most promising Liquid Metal (LM) option to be used as a PFC material due to its unique characteristics as a plasma-material boundary. However, there are intrinsic drawbacks (high vapor pressure and hydrogen retention) motivating the investigation of tin-lithium (SnLi) alloys that may have greater benefits but also ameliorating the problems. In this work, studies to determine feasible alloy candidates for flowing LM PFCs are explored. Different key issues have been studied, including topics such as wettability, compatibility of the alloy with material substrate and stability/homogeneity of the liquid surface depending on the alloy composition, focusing on the behavior of eutectic mixtures whose long-term stability would be in principle favored. These trials will elucidate if liquid surface instabilities and/or droplet ejection, that may cause a drawback when using just tin based LM components, are also present with the alloy. Material characterization techniques such as 3D, SEM as well as EDXS among others, combined with image processing tools have been utilized. The results of this investigations focusing on future perspectives and possible optimum scenarios for the eventual testing of the alloy will be addressed. [Preview Abstract] |
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