Bulletin of the American Physical Society
64th Annual Meeting of the APS Division of Plasma Physics
Volume 67, Number 15
Monday–Friday, October 17–21, 2022; Spokane, Washington
Session NO05: Particle and Power Handling, Divertor Physics, and Plasma-Material InteractionsLive Streamed
|
Hide Abstracts |
Chair: Michele Romanelli, Tokamak Energy Ltd Room: Ballroom 111 B |
Wednesday, October 19, 2022 9:30AM - 9:42AM |
NO05.00001: Overview of recent DiMES and MiMES PMI experiments in DIII-D Dmitry L Rudakov, Tyler Abrams, Igor Bykov, Zana Popovic, Dan M Thomas, Gregory Sinclair, Shawn A Zamperini, John D Elder, Robert D Kolasinski, Jonathan D Coburn, William R Wampler, Jonathan G Watkins, Ulises Losada, Adam McLean, Jacob H Nichols, Svetlana Ratynskaia An overview of Plasma-Material Interactions (PMI) research at DIII-D tokamak over the last two years using the Divertor Material Evaluation Station (DiMES) and the Mid-plane Material Evaluation Sample (MiMES) is presented. Net/gross erosion and re-deposition of common and novel Plasma Facing Materials (PFMs) such as various grades of tungsten and Wf/W composites/alloys, SiC and other advanced materials are studied in the divertor. Comparisons of spectroscopy and post-mortem ion beam analysis are used to calibrate the S/XB coefficients for various spectral lines. DiMES is also used for studies of hydrogenic retention in various PFMs, dust production and transport, pellet launches, and effects of transients like edge-localized modes on PFMs and arc ignition. Experiments with controlled melting of W and Al (as proxy to Be) are performed/planned and results will be compared to MEMOS-U modeling. MiMES, located near the outboard mid-plane, is used primarily to house Collector Probes (CPs) in material migration experiments with 13C methane injection and studies of W leakage from a closed SAS-VW divertor. Comparisons of CP results with DIVIMP/MAFOT/3DLIM modeling will be presented. |
Wednesday, October 19, 2022 9:42AM - 9:54AM |
NO05.00002: Update of the Latest Operational Results from HIDRA and MEME Daniel Andruczyk, Rabel Rizkallah, Daniel O'Dea, Andrew J Shone, Brandon Kamiyama, Sam Smith, Rajesh Maingi, Charles E Kessel, Peter Buxton, Mikhail Gryaznevich, Davide Curreli, David N Ruzic Steady-state operation in HIDRA has been achieved with individual discharges lasting between 3600 to 5400 s. This is now standard operation for HIDRA. This is an important feature to have for future PMI and Li operations. A Li evaporation experimental campaign has been undertaken to understand He plasma-Li interactions and low recycling operation. The first H2 experiments have also been performed in HIDRA as part of a long-term study on steady state operation. It has been observed that the characteristic time-scale of recycling and/or plasma-chemistry is on the order of ~30 minutes. Studying the impact of the recycling and/or plasma-chemistry is essential before constructing a long pulse fusion demonstrator device. Li operation with hydrogen has achieved low recycling operation resulting in improved performance. Plasma parameters are measured using spectroscopy and a reciprocating Langmuir probe. As part of the domestic liquid metal program experiments on MAGNUM-PSI and HIDRA experiments have been undertaken to measure and characterize vapor shielding effects. In MEME a steady-state liquid Li loop has been built and is operating. Experiments are underway with bulk surface wetting experiments of PFC surfaces. This presentation will show the latest results from these experiments. |
Wednesday, October 19, 2022 9:54AM - 10:06AM |
NO05.00003: Ejection of ionic species from lithium and lithium compounds by low energy hydrogen and noble gas ion irradiation Shota Abe, Evan T Ostrowski, Anurag Maan, Predrag Krstic, Dick Majeski, Bruce E Koel We have conducted experimental measurements that quantify ejected lithium and impurity species in the form of positive and negative ions from lithium compounds during irradiation by hydrogen or noble gas ions. Lithium is a candidate material for plasma-facing components (PFCs) of magnetic confinement fusion devices. Ionic states of ejected species can play an important role in the plasma-material interaction as those are strongly affected by the sheath potential. The experiments include ions incident at ultralow energies (< 100 eV), which is relevant to tokamak divertor environments. Experiments detected Li as a positive ion species and O and H impurities as negative ion species. Ne and Ar ion irradiations, which are commonly used for glow discharge cleaning, were found to be an efficient O and H removal method. D ion irradiation works more efficiently for O and H removal than H ion irradiation. Oxidation of Li enhanced the ejection of Li positive ions. We observed different concentrations of ejected species from the surface at different temperatures (300 K and 460 K). The kinetic energy was found to be different for each ejected species, which suggests different ejection mechanisms. Quantified ejection yields for each species and ion irradiation will be presented. |
Wednesday, October 19, 2022 10:06AM - 10:18AM |
NO05.00004: Modeling a Lithium Vapor Box Divertor and Resulting Ion Flows on NSTX-U using SOLPS Eric D Emdee, Robert J Goldston Divertor detachment with medium-Z impurities seeded through gas puffing can entail radiating regions within the last closed flux surface. The lithium vapor box divertor seeks to detach via near-target low-Z lithium evaporation with the result that such a radiating region does not form. We show SOLPS-ITER predictions for the effect of a lithium vapor box divertor on NSTX-U. Past work has shown the lithium vapor box is capable of reducing 65 MW/m2 of perpendicular heat flux to below 5 MW/m2 for upstream lithium densities equal to a few percent of the electron density. Those results are tested for their sensitivity to choices of transport coefficients, recycling coefficients and puffing location. Even when transport coefficients are reduced to provide less particle flow from the core and higher heat flux at the target, sub-10 MW/m2 solutions are available to the lithium vapor box, as compared with an unmitigated perpendicular heat flux of 92 MW/m2. Private Flux Region (PFR) fuel puffing is seen to be more effective at reducing upstream lithium content while Common Flux Region (CFR) fuel puffing is seen to be more effective at heat flux reduction. The efficacy of both puffing locations is increased by increases to the divertor recycling coefficient. Increased recycling at walls upstream of the baffles improves the effect of the puffs, leading to cases with lower upstream lithium content for less heat flux. |
Wednesday, October 19, 2022 10:18AM - 10:30AM |
NO05.00005: Liquid Lithium Loop for Open-Surface Flow on PFCs Steven Stemmley, Cody Moynihan, David N Ruzic Interest in liquid lithium as a plasma facing component (PFC) has significantly increased as of recent. This is likely due to lithium’s gettering abilities which can provide access to low recycling regimes, as well as its heat flux handling capabilites. At the University of Illinois, extensive research has been focused on liquid lithium PFC geometries, particularly with thermoelectrically driven flow (TEMHD), which have been shown to handle near 10 MW m−2 of heat flux. A clean, flowing lithium surface will getter hydrogen and impurities from the plasma which then must be removed from the PFC. To provide a continuously refreshing surface, a loop system has been designed to provide lithium from an external reservoir, to the PFC, and removed. Components such as the lithium reservoir, liquid metal pumps, flow meters, and safety systems have been designed, constructed, and are being tested. Experimentally, the lithium will be pumped into the SLiDE device at the University of Illinois, where the lithium will flow across a PFC with a 3-D LiMIT style geometry in a variable magnetic field (up to 2000 G). Heat fluxes up to 15 MW m−2 will be applied to the lithium plate via a linear electron beam. Separately, deuterium plasma exposure will be tested to verify hydrogenic species extraction in a fully circulating scenario. Results from the various loop component testing will be presented. |
Wednesday, October 19, 2022 10:30AM - 10:42AM |
NO05.00006: 3D Heat Flux and Melt Motion Analysis of Angled Tungsten Samples on DIII-D Jonathan D Coburn, Robert D Kolasinski, Dinh D Truong, Dmitry L Rudakov, Huiqian Wang, Jun Ren, Charlie Lasnier, Claudio Marini, Joshua D Sugar, Richard Nygren, Tyler Abrams, Jonathan G Watkins Dispersoid-strengthened tungsten (DSW) offers an avenue for increasing the recrystallization resistance of tungsten (W) alloys at higher temperatures. Thermal analysis is presented for a high heat flux exposure of angled (15°) DSW and ITER-grade W samples in the DIII-D tokamak. Recrystallization, cracking, and melting of both materials was observed as well as dispersoid evaporation. The exposure was performed using the Divertor Materials Evaluation System (DiMES) at DIII-D. Samples endured 9 H-mode discharges with ELMs. The inter-ELM heat flux ??⊥ strongly varied from ~11 – 24 MW/m2 on the angled surfaces. ELMs contributed up to 115 MW/m2 at ~40 Hz. Consequently, two of the samples nearest the strike point melted. The resulting melt motion was radially inward in the ?????? direction at ~20 mm/s. About 10 mm3, or ~50%, of the exposed geometry was displaced due to melt motion. EBSD microscopy shows uniform recrystallization throughout the sample depth. SMITER and SIERRA/Aria codes are used to estimate ??⊥, temperature evolution, melt formation, and ?????? melt motion. |
Wednesday, October 19, 2022 10:42AM - 10:54AM |
NO05.00007: A comprehensive molecular dynamics study of deuterium trapping and bubble formation on tungsten Enes Ercikan Understanding the trapping and de-trapping mechanisms of hydrogen isotopes on plasma facing materials (PFMs) of fusion devices is critical due to concerns of safety, fuel economy, and device efficiency. Recent computational and experimental studies focusing on deuterium bombardment on tungsten, current candidate of PFMs of tokamak design fusion reactors, show that nano-sized bubbles are formed in subsurface region of tungsten due to high concentration of deuterium. In this study, the deuterium bombardment with various energies on tungsten structures including monocrystalline and polycrystalline with grain boundary at various temperatures was simulated by molecular dynamics methods employing Large-scale Atomic/Molecular Parallel Simulator (LAMMPS) code. The Tersoff interatomic potential is employed to determine interaction between particles. The parameters of deuterium trapping such as trapping rate, implantation depth, and the mechanism of bubble formation are analyzed. Additionally, the D bubble formation is observed during high temperature (600K or higher) tungsten simulations. The bubble formation is being analyzed to study the effect of other parameters, e.g., crystalline structure, deuterium energy, etc. The goal is to identify reasonable operating windows for candidate PFMs to enhance lifetime. |
Wednesday, October 19, 2022 10:54AM - 11:06AM |
NO05.00008: Surface morphology changes observed in dispersion-strengthened tungsten alloys exposed to L- and H-mode plasmas in the DIII-D tokamak Hanna Schamis, Eric Lang, Gregory Sinclair, Adam McLean, Robert S Wilcox, Chase Hargrove, Camila Lopez Perez, Matthew S Parsons, Jean Paul Allain Dispersion-strengthened tungsten alloys are being investigated as candidate PFC materials due to their enhanced thermomechanical properties compared to pure tungsten. Samples of W-1.1 wt.% TiC, W-1.1 wt.% TaC, and W-1.1 wt.% ZrC were exposed to DIII-D divertor plasmas with the DiMES exposure probe to measure the impact of dispersoids on D retention, material sputtering and surface morphology. Scanning electron microscopy (SEM) analysis of the samples indicates that exposure to H-mode plasmas resulted in a higher density and size of blister-like surface features (up to ~10s µm) than exposure to L-mode plasmas (as small as ~150-200 nm). The H-mode plasmas had an edge localized mode (ELM) period of 10-20 ms, particle flux of ~4x1021 m-2s-1 and total fluence of ~1.6x1023 m-2 at the probe location, and the L-mode plasmas had a particle flux of ~7x1021 m-2s-1 and total fluence of ~1.2x1023 m-2. Graphite deposits were also observed in SEM micrographs; their presence will be further explored with energy-dispersive spectroscopy. To measure the erosion yield of tungsten and the second phase dispersoids, spectroscopy data was collected using visible emission spectroscopy for the W-I, Ti-I, Zr-I, and Ti-I spectral lines on each individual sample, the results of which will be presented. |
Wednesday, October 19, 2022 11:06AM - 11:18AM |
NO05.00009: Development of an absorbing first-wall interface: Deuterium trapping in Tantalum cold-spray coatings Mykola Ialovega, Danah Velez, Marcos X Navarro, Hwasung Yeom, Tyler Dabney, Arkadi Kreter, Kumar Sridharan, Jay K Anderson, Cary B Forest, Oliver Schmitz Effective control of hydrogen isotopes’ (HIs) partial pressure in the plasma edge in small fusion devices such as WHAM [1] is important to limit energy losses associated with the charge exchange events. We have investigated tantalum (Ta) cold-sprayed deposit [2,3] as a potential plasma-facing material capable of trapping significant amounts of HIs and to withstand the high particle loads at surface temperatures up to 1000 K. Ta was cold-sprayed on 316L stainless-steel substrates under five different spray conditions to achieve a range of coating microstructures. Surface and cross-sectional Scanning Electron Microscopy observations were used to evaluate the microstructure of the coatings. X-ray photoelectron spectroscopy confirmed a presence of a few nm-thick Ta2O5 oxide layer on the surface. 95 eV/D ion (deuterium) irradiation of the samples performed at the fluence of ~3×1025 D m-2 at the surface temperature Tsurf=523 K revealed the formation of D incorporated Ta-oxide at the surface. Tantalum deutoride (TaD) was observed in the bulk due to the incorporation of D in the Ta lattice, however no TaD, was observed at Tsurf=1000 K. Thermodesorption experiments performed on the ion irradiated samples and samples exposed to a D2 gas flow revealed enhanced D trapping in the cold-sprayed Ta compared to the bulk Ta when heated up to 1000 K. |
Wednesday, October 19, 2022 11:18AM - 11:30AM |
NO05.00010: Modeling and simulations of the dynamic and static D/T retention during DIII-D discharges Tatyana S Sizyuk, Tyler Abrams We used our ITMC-DYN package to simulate recent DIII-D experiments where W samples with different microstructure and damage rate were irradiated during ten discharges. We developed a new approach for modeling trapping sites accumulation depending on irradiation conditions and sample temperature. We predicted D accumulation during and between discharges and explained the spatial distribution of high-energy traps formation/growth. Carbon accumulated on W surface in DIII-D can affect D behavior in plasma facing materials (PFMs). We compared our modeling of impurity effects with recent results of lab experiments for W and WC. |
Wednesday, October 19, 2022 11:30AM - 11:42AM |
NO05.00011: A 0-dimensional simulation of particle balance between plasma and plasma facing components in the LHD Zhengnan Jiang, Gen Motojima, Suguru Masuzaki, Miyuki Yajima, Motoshi Goto We have developed a new 0-dimensional simulation for investigating the particle balance in the LHD plasma. The hydrogen barrier model (from K.Hanada 2015 J. Nucl. Mater. 463 1084) is adopted for evaluating the number of particles stored on the plasma-facing materials, including the first wall and the divertor tiles, and it is assumed that all the particle reactions only took place in the edge plasma. We attempted a calculation using experimental data, i.e, the electron density and temperature, and the neutral pressure, from an LHD discharge in nearly steady state, and it was confirmed that the steady state condition was actually reproduced. As a result, it was found that the first wall has a certain maximum ability of particle storage (4.8×1022 particles) against certain electron density (3×1019 m-3) and temperature (1.2keV), while the divertor tile has very little effect on the particle balance. |
Wednesday, October 19, 2022 11:42AM - 11:54AM |
NO05.00012: Occurrence and control of X‐point radiators and marfes in ASDEX Upgrade: experimental validation of an analytic model Ulrich Stroth, Matthias Bernert, Dominik Brida, Marco Cavedon, Ralph Dux, Tilmann Lunt, Ou Pan, Marco Wischmeier, Anja Gude, Felix Klossek, Mark Maraschek, Bernhard Sieglin Cold, dense and strongly radiating plasma volumes play an important role for power exhaust and the protection of the plasma facing components in fusion devices. A recently developed analytical model [1] deals with the occurrence of such a plasma volume known as X‐point radiator (XPR) [2], where a key role is attributed to neutral deuterium fueling. Furthermore, a criterion is derived, predicting marfes [3] when the XPR becomes MHD unstable. The main results from the model are consistent with experimental observations and with results from recent SOLPS-ITER simulations [4]. The model explains why stable XPRs are more commonly observed in metal‐wall tokamaks with impurity seeding. The model provides simple expressions which can be used for the extrapolation to future devices. |
Wednesday, October 19, 2022 11:54AM - 12:06PM |
NO05.00013: GITR Monte Carlo Predictive Simulations of the DIII-D SAS Divertor Alyssa L Hayes, Harry Hughes, Tim Younkin, Jerome Guterl, Gregory Sinclair, Zachary J Bergstrom, Jeremy D Lore, Jon T Drobny, Davide Curreli, Brian D Wirth The Global Impurity Transport Code (GITR) is a 3D fully gyrokinetic Monte Carlo transport code that can track impurity particle motion in great detail, capturing full 3D gyro-orbits, prompt re-deposition, and long-range migration distributions. |
Wednesday, October 19, 2022 12:06PM - 12:18PM |
NO05.00014: The effects of alternative divertor configurations on upstream SOL turbulence in TCV Nicola Offeddu, Woonghee Han, Christian Theiler, Theodore Golfinopoulos, James L Terry, Earl S Marmar, Curdin Wuthrich, Diego Sales de Oliveira, Davide Galassi In attached and detached lower single-null L-mode plasma discharges at TCV, we observed [1] turbulent SOL filaments to mostly populate the resistive X-point regime. In the far and, in certain conditions, near SOL, upstream filaments extend beyond the X-point and into the diverter region. This raised the question if, and to what extent, the divertor magnetic geometry influences upstream filament dynamics, and how this could be used in the design of a fusion reactor. We therefore extend this study to X-point Target (XPT) and SnowFlake (SF) divertor configurations. In SF-diverted discharges, we find that the position of the secondary X-point affects filament cross-field size and radial velocity, in the measure of its effects on the parallel connection length. When the latter is increased by ~50-100% filament size and velocity also increase by up to 30%. Filaments in XPT-diverted plasmas appear to be less sensitive to the geometry, consistent with the parallel distance from the midplane to the secondary X-point, which is 2x larger in XPT-diverted discharges than in SF. The implications of these findings for the prospects of alternative divertor solutions will be discussed. |
Wednesday, October 19, 2022 12:18PM - 12:30PM |
NO05.00015: Assessment of light impurity content at the WEST divertor during L-mode discharges based on an integrated multi-diagnostic workflow. Alex GROSJEAN, David C Donovan, Sean R Kosslow, Curtis Johnson, Jake B Maeker, Nicolas Fedorczak, James P Gunn, Christophe Guillemaut, C.Christopher Klepper, E.A. Unterberg, Jacob H Nichols WEST is an actively cooled, long-pulse tokamak with nearly all plasma-facing components (PFC) made of tungsten (W). For long-pulse operation, the W impurity content and transport to the core plasma are critical concerns that require further investigation to improve plasma performance and PFC durability. Assessing the light impurity (B, C, N, O) content at the divertor is a key metric as it is the main contributor to tungsten sputtering at the divertor in L-mode. |
Follow Us |
Engage
Become an APS Member |
My APS
Renew Membership |
Information for |
About APSThe American Physical Society (APS) is a non-profit membership organization working to advance the knowledge of physics. |
© 2024 American Physical Society
| All rights reserved | Terms of Use
| Contact Us
Headquarters
1 Physics Ellipse, College Park, MD 20740-3844
(301) 209-3200
Editorial Office
100 Motor Pkwy, Suite 110, Hauppauge, NY 11788
(631) 591-4000
Office of Public Affairs
529 14th St NW, Suite 1050, Washington, D.C. 20045-2001
(202) 662-8700