Bulletin of the American Physical Society
63rd Annual Meeting of the APS Division of Plasma Physics
Volume 66, Number 13
Monday–Friday, November 8–12, 2021; Pittsburgh, PA
Session PO08: MFE: Plasma-Material InteractionsOn Demand
|
Hide Abstracts |
Chair: Tyler Abrams, General Atomics - San Diego Room: Rooms 317-318 |
Wednesday, November 10, 2021 2:00PM - 2:12PM |
PO08.00001: Overview of HIDRA and MEME Research with at the University of Illinois university E illinois, Rabel Rizkallah, Andrew J Shone, Daniel O'Dea, Samuel O Smith, Rajesh Maingi, Charles E Kessel, Sergey Smolentsev, Jean Paul Allain, Zachariah Koyn, Thomas W Morgan, Fabio Romano, David N Ruzic, Zhen Sun, Guizhong Zuo, Jiansheng Hu Fusion and liquid metal research has many different avenues at the Center for Plasma Material Interactions (CPMI). Two of those are through the National Liquid Metal PFC (NLMP) program which is a multi-institutional collaboration to develop a flowing liquid lithium divertor. This program is aimed at designing and developing a flowing liquid metal divertor for FNSF. The role of CPMI is to do many of the lithium and material compatibility studies related to flow, surface wetting, and surface damage as well as understanding the role vapor shielding plays in protecting the PFC surface and heat flux dissipation through modeling and experiments. Furthermore, HIDRA, which is a l = 2, m = 5 steady-state classical stellarator is now fully operational with routine plasma operation with discharge lengths between t = 60 – 1000 s. A material analysis test-stand has been developed to expose different materials to HIDRA’s plasma discharges, including with lithium. Some of the first Li exposures in helium discharges have yielded interesting results that may have implications in how helium can be removed by lithium. Finally, the LiMIT concept, which is a self-pumping PFC based on TEMHD, was run in EAST in January 2020. This was the first full scale deployment of LiMIT in a reactor and the first results have been obtained showing that the plate was able to operate well under the EAST operating conditions. This presentation will introduce important results in HIDRA, the NLMP program, LiMIT on EAST, and look at the future of liquid metal PFC’s. |
Wednesday, November 10, 2021 2:12PM - 2:24PM |
PO08.00002: Comprehensive study of the processes in H irradiation of the Li and Li-compound surfaces Predrag Krstic, Shota Abe, Evan Ostrowski, Bruce E Koel We performed computer simulations and experiments of processes for the impact of hydrogen (D, T) on lithium and lithium compounds Li2O, LiOH, and LiH surfaces. The probabilities of retention, reflection, and sputtering were computed in the range of impact energies of 5-100 eV, for various impact angles, and for amorphous and crystalline structures of the target surfaces. An unexpected and distinct dependence of the processes on the impact angle were obtained. The results were computed by molecular dynamics with the REAX bond-order force field, utilizing the Electronegativity Equalization Method, due to the imbalance in electronegativities of the multicomponent surfaces and impact atoms. This approach was partially benchmarked by quantum-classical molecular dynamics. The calculated probabilities are compared with controlled laboratory experimental results using lithium and lithium compound surfaces irradiated by low energy (< 100 eV) H and D ions. |
Wednesday, November 10, 2021 2:24PM - 2:36PM |
PO08.00003: Power Handling of LiMIT during Exposure to EAST plasma Daniel O'Dea, Rabel Rizkallah, Rajesh Maingi, Steven Stemmley, Zhen Sun, Guizhong Zuo, Jiansheng Hu, David N Ruzic, Daniel Andruczyk Liquid lithium PFCs offer an attractive solution to the issues faced by solid high Z PFCs such as: the sputtering of high Z impurities into the plasma core, fuzz formation due to helium implantation and thermal damage (melting, blistering and cracking.) By presenting a constantly refreshing surface a flowing lithium surface prevents damage to the PFC, via a reduction in sputtering of solid material and convective heat removal by the liquid, increasing component lifetime. The utilization of lithium presents several other benefits for fusion plasmas through impurity removal, helium ash pumping and a reduction in anomalous transport within the plasma. To further the development of lithium PFCs a LiMIT type limiter has been exposed to the EAST plasma, this work will present the experimental results from the campaign alongside a COMSOL model developed to calculate the conductive heat flux incident on LiMIT. During NBI heated shots in EAST the plate experienced a calculated peak conductive heat flux of 8MWm-2 further supplemented by the heat stored convectively in the lithium and the heat dissipated by the lithium vapor cloud, ultimately resulting in LiMIT withstanding a heat flux 10MWm-2 without sustaining any macroscopic damage. |
Wednesday, November 10, 2021 2:36PM - 2:48PM |
PO08.00004: SOLPS-ITER calculations of lithium vapor box mitigation of qpeaktarget = 96 MW/m2 in NSTX-U Eric D Emdee, Robert J Goldston NSTX-U can provide a valuable test facility for reactor-relevant heat fluxes. Goldston-Eich scaling of the H-mode SOL power width [1,2] for NSTX-U parameters (Pin = 10MW, frad=0.3, Ip=2MA, BT=1T) gives a prediction of qpeaktarget ~ 96 MW/m2 for extremely low flux expansion geometry. The lithium vapor box seeks to detach the plasma via dense lithium vapor localized near the target through evaporation and condensation [3]. Extensive modeling has been done on a simple lithium evaporation concept, proving its viability in detaching a lower heat flux L-mode discharge at a viable upstream lithium concentration [4]. Here the design of the vapor box is examined for high-power NSTX-U H-modes using a higher degree of divertor closure than previously simulated by SOLPS. Configurations with and without a baffle are compared using upstream lithium concentration, divertor heat flux reduction, and upstream plasma temperature as primary metrics of performance. The heat flux to the target is seen to be reduced to levels that should be acceptable for capillary porous material, ≤ 15 MW/m2. The next set of studies will examine the role of deuterium gas puffing in reducing upstream lithium concentration from the values of ~5% currently observed. |
Wednesday, November 10, 2021 2:48PM - 3:00PM |
PO08.00005: Transient Model of Fuzzy Tungsten Growth, Annealing, and Erosion in Fusion Environments Adam M Darr, Allen L Garner, Ahmed Hassanein Under fusion conditions, tungsten can develop a porous (fuzzy) surface microstructure [1]. This is concerning for most metallic divertor materials, especially during Edge Localized Mode (ELM) events with high particle bombardment [2,3]. If this fuzzy layer breaks off, it could introduce significant high Z impurities into the bulk plasma, potentially quenching the fusion reaction [4]. Several models have been proposed to model fuzz growth and erosion [4-6]. In this study, we combine elements of several prior theories to model transient growth, annealing, and erosion processes in fuzzy tungsten using a zero-dimensional kinetic model. We find steady-state conditions for fuzz layer thickness – including novel pseudo-steady-states induced by ELMs – and compare to previous models. Extensions to higher dimensionalities and tungsten alloys will be addressed. |
Wednesday, November 10, 2021 3:00PM - 3:12PM |
PO08.00006: Experimental methods to characterize surface changes of plasma facing materials exposed to ITER-like relevant fusion conditions Tyler E Ray, Yaoxing Wu, Ahmed Hassanein Inside a tokamak fusion reactor there will inevitably be wall degradation caused by plasma interactions. Recent studies have shown growth of helium nano tendrils, often called fuzz, on the surface of tungsten plasma facing components (PFCs) under low energy ion irradiation within a high temperature environment. Fuzz structures can decrease the thermal conductivity and hardness, but more importantly can easily sputter and erode introducing impurities into the core plasma, quenching the plasma. Unique experiments performed at CMUXE are used to quantify the fuzz depth, tendril size, and erosion by using ion beams and lasers. Different fuzz depths and densities are formed on tungsten surfaces when varying He/D ion irradiation energy and exposure time. SEM analysis provides insight on tendril size and fuzz depth while Spectrograph and Faraday cup are used to characterize the surface remotely. Experiments are performed under conditions relevant to ITER like devices including simultaneous high temperature D-He irradiation combined with laser irradiation to simulate ELMs. The goal of this research is to further advance the understanding of plasma interactions with PFCs regarding conditions to minimize fuzz structures and identify better materials and plasma conditions for divertor materials. |
Wednesday, November 10, 2021 3:12PM - 3:24PM |
PO08.00007: ERO2.0 Benchmarking of Material Mixing for Local and Global Beryllium Transport in JET Marcos X Navarro, Juri Romazanov, Andreas Kirschner, Oliver Schmitz Plasma-material interactions (PMI) is a challenging topic because it affects the lifetime of wall components and plasma performance. The fully 3D ERO 2.0 Monte Carlo code is capable of modeling global impurity transport, which defines the erosion/deposition patterns in a reactor. A material mixing model was implemented in the ERO2.0 code and successfully benchmarked against its predecessor, ERO, for local impurity transport along the JET tungsten divertor tile 5. The case for benchmark corresponds to inter-ELM plasma conditions, with experimentally fitted exponentially decaying plasma profiles along the private flux region (PFR) and the scrape off layer. Because ERO was restricted to local studies, certain assumptions like the constant level of background Be was assumed. With the ERO2.0 improvements, it is possible to see how a Be source in the main chamber affects the Be deposition in the W divertor and the Be/W mixed layer formation. Two cases are considered: (1) a fixed beryllium background concentration and (2) beryllium sources injected at the outer mid-plane limiters to mimic the main wall beryllium source and providing the beryllium flux to the divertor. Global modeling successfully recreates beryllium buildup in the PFR along tile 5 as expected from the local benchmark. |
Wednesday, November 10, 2021 3:24PM - 3:36PM |
PO08.00008: Beryllium Melt Motion and Splashing from First Wall in ITER: OpenFOAM CFD Modeling Cheng Zhang, Gennady Miloshevsky Beryllium (Be) material possesses low density, high strength, and high thermal conductivity. Therefore, it is planned to be utilized as a first wall in ITER. However, the Be first wall panels can undergo melting and melt splashing during the transient events of high magnitude that can lead to the surface erosion. Be droplets which ejected from the melt surface can affect drastically the efficiency of ITER fusion reactor. Therefore, it is important to analyze the physics mechanisms governing the Be melt splashing under the relevant ITER conditions. The OpenFOAM CFD model was used to investigate the effects of Be vapor on the melt motion and splashing. The influence of vapor pressure, mass density and velocity changing with the temperature at the surface on the Be melt motion and droplet ejection is studied. The results provide an insight into the development of melt structures and waves at the vapor-melt interface and physical mechanisms of melt entrainment into the motion within a thin boundary layer. |
Wednesday, November 10, 2021 3:36PM - 3:48PM |
PO08.00009: New ITER divertor design using carbon insert on the tungsten plate to mitigate ELMs and secondary radiation effects on nearby components. Valeryi Sizyuk, Ahmed Hassanein Significant heat loads are expected during ELMs on divertor plate in ITER and future tokamaks normal operation. Such high power loads at the strike point (SP) can cause surface vaporization, ionization, and development of secondary plasma from the divertor materials. The secondary plasma converts and redistributes the incident core plasma energy into intense photon radiation and scattered core particle fluxes from this dense secondary plasma (more than 3 orders of magnitude denser than the core plasma). In this study, we compared the power load redistribution on nearby components for two divertor designs: current ITER design with tungsten divertor (high-Z case) and tungsten divertor with carbon insert at the SP (low-Z case). We simulated the divertor response during ELMs using our integrated HEIGHTS 3D package. These include self-consistent modeling of the interaction of incoming core plasma particles with the initial solid divertor material, energy deposition processes, MHD of secondary plasma evolution, incident core particles collisions and scattering from the dense secondary plasma, secondary plasma photon radiation, and the resulting heat loads on nearby components. Our simulations showed that carbon insert at the SP can significantly reduce the overall expected damage on W dome structure, reflector plates, and prevent W vaporization and potential core plasma contamination. |
Wednesday, November 10, 2021 3:48PM - 4:00PM |
PO08.00010: Development of rugged hydrogen sensors for measuring charge-exchange neutral flux at the wall and divertor Ryan T Hood, Robert D Kolasinski, Dinh Truong, Josh A Whaley, Dmitry L Rudakov, Tyler Abrams, Jonathan G Watkins, Albert A Talin Rugged Pd-MIS style hydrogen sensors have recently been demonstrated in the Divertor Materials Evaluation System (DiMES) at the DIII-D tokamak in upper and lower single null configurations. This pair of sensors utilized 7 nm thick gold overlayers to reduce sensitivity to low energy particles as well as thick palladium layers to protect against damage from high energy particles. Both sensors responded similarly to upper single null discharges until saturation occurred after approximately 3 shots. This is consistent with a thermal distribution of charge-exchange neutrals with a temperature ~1 keV and fluence ~ 1015 D/cm2 per shot. Results are also presented from laboratory testing and calibration performed with mass-separated 340 eV and 1500 eV deuterium (D) ion beams with a total dose > 1016 D/cm2 onto the sensors, which had a 1.7 mm diameter active area. Characterization of charge-exchange neutral flux in tokamaks is important to understanding wall erosion and redeposition in future devices. |
Wednesday, November 10, 2021 4:00PM - 4:12PM |
PO08.00011: New insight on deuterium diffusion, trapping, and desorption in plasma-facing materials during and after irradiation Tatyana S Sizyuk, Ahmed Hassanein, Jeffrey Brooks, Tyler Abrams Integrating Monte Carlo collisional processes with deterministic methods for the thermal processes, we analyzed hydrogen isotope behavior in tungsten and tungsten-based compounds, an important issue for the performance and safety of future fusion reactors. Deuterium retention was analyzed in its dependence on flux, fluence, impurities, and material temperature. We benchmarked our new models against laboratory experiments and DIII-D recent data regarding D trapping sites/defects evolution as well as ion/atom collisions, diffusion, retention, and desorption during and after exposure of D plasma and multispecies D/C plasma fluxes. Comparing modeling and experiments at various irradiation conditions and delay time between exposure and thermal desorption spectroscopy, we predicted the retained and the solute (free) D distribution in material. These simulations explained, for the first time, the effect of material temperature on spatial distribution of trapped D concentration and the four-fold increase in D accumulation at higher (400-500K) W temperatures compared with 300K. Substantial free D in the sub-surface layer of warm target can lead to stress-induced vacancies and voids formation. Low diffusivity at 300K leads to D surface accumulation and significant desorption during and after irradiation. |
Wednesday, November 10, 2021 4:12PM - 4:24PM |
PO08.00012: Modeling carbon pellet injection in DIII-D divertor for advanced material studies. Roman Smirnov, Igor Bykov, Jerome Guterl, Eva Kostadinova, Dmitriy M Orlov, Dmitry L Rudakov Injections of spherical carbon pellets into DIII-D L-mode plasma discharges were performed in order to investigate pellet dynamics and ablation in the plasma as part of the 2021 Frontiers in Science campaign. Reliable pellet ablation and transport models in high heat flux plasma environment are needed for aerospace and fusion applications. Several types of pellets, from 2 mm to 3.6 mm diameter porous carbon and 0.7 mm glassy carbon spheres, were injected using spring loaded injectors installed in DIII-D Divertor Material Evaluation System (DiMES). Pellet trajectories were tracked using fast imaging cameras in visible and IR light. The DIII-D spectroscopy diagnostics were used to monitor plasma carbon content in the core. The obtained experimental pellet trajectories and ablation rates were compared with simulations performed using DUSTT-UEDGE codes showing that the pellets experienced unexpected radial acceleration. Possible physical mechanism of the acceleration is discussed. The comparison allowed improving the models used in the codes for large pellets in order to reproduce the observed dynamics. |
Wednesday, November 10, 2021 4:24PM - 4:36PM |
PO08.00013: Hypervelocity impact in stellar media: Frontiers in Plasma Science experiments at DIII-D. Dmitriy M Orlov, Eva Kostadinova, Igor Bykov, Roman Smirnov, Dmitry L Rudakov A study of carbon ablation at high heat flux relevant to hypervelocity spacecraft entries was performed at the DIII-D as part of the Frontiers in Science campaign. Exploration missions to the Solar System's gaseous giants and hyperbolic re-entries into the Earth's atmosphere require spacecraft that can withstand high velocity (>10 km/s) and extreme heat flux (>10 MW/m2). Testing and modeling material performance in this regime is challenging due to inadequate ground testing facilities. Conditions in DIII-D L-mode edge plasma reproduce flow velocity and high heat flux comparable to those experienced during the Galileo probe's entry into the atmosphere of Jupiter. Three types of samples were used for the experiments: stationary ATJ graphite rods protruding from the floor of the vessel, 1-mm-diameter porous carbon spheres, and 700-micron-diameter glassy carbon spheres injected from the floor into the scrape-off layer and edge plasma. Experimental results for the pellet trajectories and ablation rates are compared to the UEDGE-DUSTT simulations. The resulting mass loss rates as a function of calculated heat flux to the surface of the samples are found to agree with the semi-empirical engineering models. |
Wednesday, November 10, 2021 4:36PM - 4:48PM |
PO08.00014: Effect of Thermal Cycling on Morphology Changes and Deuterium Retention in Helium Exposed Tungsten Mykola Ialovega, Elodie Bernard, Celine Martin, Regis Bisson, Christian Grisolia, Thierry Angot Near-surface morphology modification of tungsten (W) plasma-facing material (PFM) under intense particle and heat fluxes raises concerns about material property conservation and tritium retention in the ITER divertor. The accumulation of helium (He) in W induces formation of cavities, so-called helium bubbles, which may alter the retention property of the metal [1]. |
Wednesday, November 10, 2021 4:48PM - 5:00PM |
PO08.00015: Plasma-wall interaction studies in the WEST experiment and wall conditioning with the Impurity Powder Dropper Grant Bodner, Alberto Gallo, Jerome Bucalossi, Clarisse Bourdelle, Sebastijan Brezinsek, Matthieu De Combarieu, Ahmed Diallo, Corinne Desgranges, Annika Ekedahl, Christophe Guillemaut, Remy Guirlet, Jamie P Gunn, Nicolas Fedorczak, C.Christopher Klepper, Thierry Loarer, Robert A Lunsford, Patrick Maget, Philippe Moreau, A. Nagy, Francis-Pierre Pellissier, Emmanuelle Tsitrone, E.A. Unterberg The Tungsten (W) Environment in Steady-state Tokamak (WEST) experiment is the only European tokamak equipped with superconducting toroidal magnetic field coils and metallic plasma-facing components (PFCs). The first two campaigns were devoted to commissioning without boronization. From the third campaign, wall conditioning through deuterium (D2), He, and diborane (B2D6) glow discharges was routinely performed, opening the operational space. Improved wall conditions allowed a long pulse scenario (~1 minute) and divertor peak heat load up to 5.5 MW m-2 in L-mode, essential steps towards a steady state 10 MW m-2 heat load to test ITER-like W PFCs. To further reduce impurity content and radiative losses, successful real-time wall conditioning using an Impurity Powder Dropper (IPD) was recently performed. As boron (B) powder (with grain sizes < 150 um) was injected for up to 16.5 s during L-mode plasma discharges, spectroscopic measurements showed a reduction of D and O signals at the outboard limiter and the lower divertor, suggesting reduced fuel recycling and improved light impurity screening. Furthermore, in the pre-drop phase, a promising reduction of the total radiated power and no sign of enhanced W sputtering were observed as the cumulated amount of dropped B was increased. |
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