Bulletin of the American Physical Society
59th Annual Meeting of the APS Division of Plasma Physics
Volume 62, Number 12
Monday–Friday, October 23–27, 2017; Milwaukee, Wisconsin
Session JO4: Spherical Tokamaks, Other |
Hide Abstracts |
Chair: Kathreen Thome, General Atomics Room: 201AB |
Tuesday, October 24, 2017 2:00PM - 2:12PM |
JO4.00001: Overview of the Lockheed Martin Compact Fusion Reactor (CFR) Project. Thomas McGuire The Lockheed Martin Compact Fusion Reactor (CFR) Program endeavors to quickly develop a compact fusion power plant with favorable commercial economics and military utility. The CFR uses a diamagnetic, high beta, magnetically encapsulated, linear ring cusp plasma confinement scheme. Major project activities will be reviewed, including the T4B and T5 plasma heating experiments. The goal of the experiments is to demonstrate a suitable plasma target for heating experiments, to characterize the behavior of plasma sources in the CFR configuration and to then heat the plasma with neutral beams, with the plasma transitioning into the high Beta confinement regime. The design and preliminary results of the experiments will be presented, including discussion of predicted behavior, plasma sources, heating mechanisms, diagnostics suite and relevant numerical modeling. \copyright 2017 Lockheed Martin Corporation. All Rights Reserved. [Preview Abstract] |
Tuesday, October 24, 2017 2:12PM - 2:24PM |
JO4.00002: MAST Upgrade Status and Future Enhancements James Harrison The MAST Upgrade spherical tokamak has unique capabilities to address some of the key issues facing the development of fusion energy. Its main objectives are: 1) development of novel exhaust concepts, 2) contribution to the knowledge base for ITER and 3) to explore potential routes to smaller/cheaper fusion reactors. To fulfil these aims, it is equipped with 19 new poloidal field coils and closed divertors with Super-X capability. BT has been increased by 50{\%} and the pulse length and Ip have increased to 5s and 2MA respectively. Auxiliary heating is provided by on and off axis NBI. The gas fuelling system allows for injection from 10 poloidal locations. The divertors are diagnosed with probes, bolometers, Thomson scattering, IR, visible imaging and spectroscopy. Fast ion physics studies are enhanced with a new fast ion loss detector. Following the construction phase, further enhancements are underway including new diagnostics, a cryoplant to serve the cryopumps and 2 additional neutral beams to increase the heating power from 5 to 10MW. [Preview Abstract] |
Tuesday, October 24, 2017 2:24PM - 2:36PM |
JO4.00003: Non-Solenoidal Startup Research Directions on the Pegasus Toroidal Experiment R.J. Fonck, M.W. Bongard, B.T. Lewicki, J.A. Reusch, G.R. Winz The Pegasus research program has been focused on developing a physical understanding and predictive models for non-solenoidal tokamak plasma startup using Local Helicity Injection (LHI). LHI employs strong localized electron currents injected along magnetic field lines in the plasma edge that relax through magnetic turbulence to form a tokamak-like plasma. Pending approval, the Pegasus program will address a broader, more comprehensive examination of non-solenoidal tokamak startup techniques. New capabilities may include: increasing the toroidal field to 0.6 T to support critical scaling tests to near-NSTX-U field levels; deploying internal plasma diagnostics; installing a coaxial helicity injection (CHI) capability in the upper divertor region; and deploying a modest (200--400 kW) electron cyclotron RF capability. These efforts will address scaling of relevant physics to higher $B_{T} $, separate and comparative studies of helicity injection techniques, efficiency of handoff to consequent current sustainment techniques, and the use of ECH to synergistically improve the target plasma for consequent bootstrap and neutral beam current drive sustainment. This has an ultimate goal of validating techniques to produce a $\sim 1$ MA target plasma in NSTX-U and beyond. [Preview Abstract] |
Tuesday, October 24, 2017 2:36PM - 2:48PM |
JO4.00004: Non-solenoidal Startup with High-Field-Side Local Helicity Injection on the Pegasus ST J.M. Perry, G.M. Bodner, M.W. Bongard, M.G. Burke, R.J. Fonck, J.L. Pachicano, C. Pierren, N.J. Richner, C. Rodriguez Sanchez, D.J. Schlossberg, J.A. Reusch, J.D. Weberski Local Helicity Injection (LHI) is a non-solenoidal startup technique utilizing electron current injectors at the plasma edge to initiate a tokamak-like plasma at high$_{\mathrm{\thinspace }}I_{p} $. Recent experiments on Pegasus explore the inherent tradeoffs between high-field-side (HFS) injection in the lower divertor region and low-field-side (LFS) injection at the outboard midplane. Trade-offs include the relative current drive contributions of HI and poloidal induction, and the magnetic geometry required for relaxation to a tokamak-like state. HFS injection using a set of two increased-area injectors ($A_{inj} =4$ cm$^{\mathrm{2}}$, $V_{inj} \sim 1.5$ kV, and $I_{inj} \sim 8$ kA) in the lower divertor is demonstrated over the full range of toroidal field available on Pegasus ($B_{T0} \le 0.15$ T). Increased PMI on both the injectors and the lower divertor plates was observed during HFS injection, and was substantively mitigated through optimization of injector geometry and placement of local limiters to reduce scrape-off density in the divertor region. $I_{p} $ up to 200 kA is achieved with LHI as the dominant current drive, consistent with expectations from helicity balance. To date, experiments support $I_{p} $ increasing linearly with helicity injection rate. The high normalized current ($I_{N} \ge 10)$ attainable with LHI and the favorable stability of the ultra-low aspect ratio, low-$\ell_{i} $ LHI-driven plasmas allow access to high $\beta_{t} $---up to $100\% $, as indicated by kinetically-constrained equilibrium reconstructions. [Preview Abstract] |
Tuesday, October 24, 2017 2:48PM - 3:00PM |
JO4.00005: Status and Plans for NSTX-U Recovery R.J. Hawryluk, S. Gerhardt, J. Menard, C. Neumeyer The NSTX-U device experienced a series of technical problems; the most recent of which was the failure of one of the poloidal magnetic field coils, which has rendered the device inoperable and in need of significant repair. As a result of these incidents, the Laboratory performed a very comprehensive analysis of all of the systems on NSTX-U. Through an integrated system's analysis approach, this process identified which actions need to be taken to form a corrective action plan to ensure reliable and predictable operation. The actions required to address the deficiencies were reviewed by external experts who made recommendations on four high-level programmatic decisions regarding the inner poloidal field coils, limitations to the required bakeout temperature needed for conditioning of the vacuum vessel, divertor and wall protection tiles and coaxial helicity injection. The plans for addressing the recommendations from the external review panels will be presented. This research was sponsored by the U.S. Dept. of Energy under contract DE-AC02-09CH11466. [Preview Abstract] |
Tuesday, October 24, 2017 3:00PM - 3:12PM |
JO4.00006: A first look at resistive MHD stability differences between NSTX and NSTX-U high beta discharges L. A. Morton, R. J. La Haye, J. W. Berkery, J. E. Menard, N. M. Ferraro, D. P. Brennan, S. A. Sabbagh, L. F. Delgado-Aparicio, K. Tritz Comparison is made of the onset, growth rate and saturation of m/n $=$ 2/1 tearing modes in NSTX and NSTX-U high beta discharges. NSTX-U has stronger toroidal field, higher electron temperature (thus longer resistive diffusion time) and a larger aspect ratio (due to the expansion of the center stack). Experimental identification of the mode helicity, radial location, and width is accomplished by synergistically combining information from soft x-ray emission, Thomson scattering ($T_{e} $ profile), Charge Exchange Recombination ($T_{i} $ profile) and Mirnov diagnostics. Fitting the generalized Rutherford equation to the time-evolution of the island width allows evaluation of the different drive and stabilizing terms. Linear stability calculations have also been performed with M3D-C1.~ The possibility of a reduction in the stabilizing interchange effect due to curvature at somewhat larger aspect ratio in NSTX-U is one focus of the analysis. [Preview Abstract] |
Tuesday, October 24, 2017 3:12PM - 3:24PM |
JO4.00007: Simulation of the internal kink-like mode driven by the toroidal rotation in spherical tokamak G.Z. Hao, W.W. Heidbrink, D. Liu, Y.Q. Liu, M. Podesta, E. Fredrickson, D. Darrow, N. Crocker, K. Tritz Based on the L-mode discharge of NSTX, the linear simulation indicates that the internal kink-like mode can be driven by the toroidal rotation when it exceeds 25{\%} of Alfven velocity at magnetic axis. The predicted critical value of rotation is close to the experimental rotation at the onset of the mode. The mode frequency agrees well with the measured value (\textasciitilde 35 kHz). The simulated mode structure agrees with the measurement from reflectometer diagnostic which monitors the major radius larger than 120cm for the studied case. Furthermore, in simulation, the triggering of the mode is robust and insensitive to uncertainty in the reconstructed equilibrium. The preliminary analysis of soft-x ray data suggests that the mode perturbation initially occurs in the core and moves outside during the frequency chirping process of the mode. A comparison between the simulated soft-x ray and the experiment measurement in the core region will be presented. This work is supported by the US DOE under Grant Nos. DE-AC02-09CH11466, DE-FG02-06ER54867, and DE-FG03-02ER54681 [Preview Abstract] |
Tuesday, October 24, 2017 3:24PM - 3:36PM |
JO4.00008: Electron Scale Turbulence and Transport in an NSTX H-mode Plasma Using a Synthetic Diagnostic for High-k Scattering Measurements Juan Ruiz Ruiz, Walter Guttenfelder, Nuno Loureiro, Yang Ren, Anne White Turbulent fluctuations on the electron gyro-radius length scale are thought to cause anomalous transport of electron energy in spherical tokamaks such as NSTX and MAST [1, 2] in some parametric regimes [3]. In NSTX, electron-scale turbulence is studied through a combination of experimental measurements from a high-k scattering system [4] and gyrokinetic simulations. Until now most comparisons between experiment and simulation of electron scale turbulence have been qualitative, with recent work expanding to more quantitative comparisons via synthetic diagnostic development [5,6]. In this new work, we propose two alternate, complementary ways to perform a synthetic diagnostic using the gyrokinetic code GYRO. The first approach builds on previous work [5,6] and is based on the traditional selection of wavenumbers using a wavenumber filter, for which a new wavenumber mapping was implemented for general axisymmetric geometry. A second alternate approach selects wavenumbers in real-space to compute the power spectra. These approaches are complementary, and recent results from both synthetic diagnostic approaches applied to NSTX plasmas will be presented. [1] Kaye NF 2007, [2] Valovic NF 2011, [3] Guttenfelder PoP 2013, [4] Smith RSI 2008, [5] Poli PoP 2010, [6] Poli APS 2010. [Preview Abstract] |
Tuesday, October 24, 2017 3:36PM - 3:48PM |
JO4.00009: Divertor-localized fluctuations in NSTX-U L-mode discharges Filippo Scotti, V.A. Soukhanovskii, S. Zweben, J. Myra, D. Baver, S.A. Sabbagh The 3-D structure of divertor turbulence is characterized in NSTX-U by means of fast camera imaging. Edge and divertor turbulence can be important in determining the heat flux width in fusion devices. Field-aligned filaments are found on the divertor legs via imaging of C III and D-$\alpha$ emission in NBI-heated diverted L-mode discharges, similar to observations in Alcator C-Mod and MAST. These flute-like fluctuations of up to 10-20$\%$ in RMS/mean are radially localized around the separatrix and limited to the region below the X-point. Poloidal and parallel correlation lengths are a few cm (10-50$\rho_i$) and several meters, respectively. For the outer leg filaments, poloidal correlation lengths decrease along the leg away from the strike point and typical effective toroidal mode numbers are in the range of 10-20. Opposite toroidal rotation is observed for inner (co-current rotation) and outer leg (counter-current rotation) filaments with apparent poloidal propagation of $\sim$1 km/s. The poloidal motion of outer leg filaments is opposite to the one typically observed for NSTX upstream blobs in the scrape-off layer. The shape, dynamics and absence of correlation with upstream turbulence suggest that these fluctuations are generated and localized in the divertor region. [Preview Abstract] |
Tuesday, October 24, 2017 3:48PM - 4:00PM |
JO4.00010: Characterization of boronized graphite in NSTX-U and its effect on plasma performance Felipe Bedoya, Jean Paul Allain, Robert Kaita, Charles Skinner Plasma Facing Components (PFC) conditioning can have a crucial influence in plasma performance in tokamak machines. The National Spherical Torus Experiment (NSTX-U) used boronization as the main wall conditioning technique during the FY16 experimental campaign. The Materials Analysis Particle Probe (MAPP), a characterization facility, was used to investigate the surface of ATJ graphite exposed to boronization and plasma in the tokamak using X-ray Photoelectron Spectroscopy (XPS). The measurements showed that plasma induced oxidation plays a critical role in the chemical evolution of the surfaces and as a consequence in plasma performance. Additionally, ex-vessel in-situ laboratory experiments and post-mortem studies of extracted NSTX-U tiles were performed to complement the observations made with MAPP, including controlled D irradiations and XPS depth profiles. These three methodologies show congruent results where D exposures increase the oxygen concentration between 20-30{\%}, highlighting the influence of these two species on the chemistry of the samples. [Preview Abstract] |
Tuesday, October 24, 2017 4:00PM - 4:12PM |
JO4.00011: Elemental and topographical imaging of microscopic variations in deposition on NSTX-U and DIII-D samples C.H. Skinner, R. Kaita, B.E. Koel, C.P. Chrobak, W.R. Wampler Tokamak plasma facing components (PFCs) have surface roughness that can cause microscopic spatial variations in erosion and deposition and hence influence material migration. Previous RBS measurements showed indirect evidence for this but the spatial (~0.5mm) resolution was insufficient for direct imaging. We will present elemental images at sub-micron resolution of deposition on NSTX-U and DiMES samples that show strong microscopic variations and correlate this with 3D topographical maps of surface irregularities. The elemental imaging is performed with a Scanning Auger Microprobe (SAM) that measures element-specific Auger electrons excited by an SEM electron beam. 3D topographical maps of the samples are performed with a Leica DCM 3D confocal light microscope and compared to the elemental deposition pattern. The initial results appear consistent with erosion at the downstream edges of the surface pores exposed to the incident ion flux, whereas the deeper regions are shadowed and serve as deposition traps. [Preview Abstract] |
Tuesday, October 24, 2017 4:12PM - 4:24PM |
JO4.00012: Design and Modeling of a Liquid Lithium LiMIT Loop Matthew Szott, Michael Christenson, Steven Stemmley, Chisung Ahn, Daniel Andruczyk, David Ruzic The use of flowing liquid lithium in plasma facing components has been shown to reduce erosion and thermal stress damage, prolong device lifetime, decrease edge recycling, reduce impurities, and increase plasma performance, all while providing a clean and self-healing surface. The Liquid Metal Infused Trench (LiMIT) system has proven the concept of controlled thermoelectric magnetohydrodynamic-driven lithium flow for use in fusion relevant conditions, through tests at UIUC, HT-7, and Magnum PSI. As the use of liquid lithium in fusion devices progresses, emphasis must now be placed on full systems integration of flowing liquid metal concepts. The LiMIT system will be upgraded to include a full liquid lithium loop, which will pump lithium into the fusion device, utilize TEMHD to drive lithium through the vessel, and remove lithium for filtration and degassing. Flow control concepts recently developed at UIUC -- including wetting control, dryout control, and flow velocity control -- will be tested in conjunction in order to demonstrate a robust system. Lithium loop system requirements, designs, and modeling work will be presented, along with plans for installation and testing on the HIDRA device at UIUC. [Preview Abstract] |
Tuesday, October 24, 2017 4:24PM - 4:36PM |
JO4.00013: On the Development of Hydrogen Isotope Extraction Technologies for a Full LiMIT-Style PFC Liquid Lithium Loop Michael Christenson, Matthew Szott, Steven Stemmley, Jeremy Mettler, John Wendeborn, Cody Moynihan, Chisung Ahn, Daniel Andruczyk, David Ruzic Lithium has proven over numerous studies to improve core confinement, allowing access to operational regimes previously unattainable when using solid, high-Z divertor and limiter modules in magnetic confinement devices. Lithium readily absorbs fuel species, and while this is advantageous, it is also detrimental with regards to tritium inventory and safety concerns. As such, extraction technologies for the recovery of hydrogenic isotopes captured by lithium require development and testing in the context of a larger lithium loop recycling system. Proposed reclamation technologies at the University of Illinois at Urbana-Champaign (UIUC) will take advantage of the thermophysical properties of the lithium-hydrogen-lithium hydride system as the driving force for recovery. Previous work done at UIUC indicates that hydrogen release from pure lithium hydride reaches a maximum of 7 x 10$^{\mathrm{18}}$ s$^{\mathrm{-1}}$ at 665 \textdegree C. While this recovery rate is appreciable, reactor-scale scenarios will require isotope recycling to happen on an even faster timescale. The ratio of isotope dissolution to hydride precipitate formation must therefore be determined, along with the energy needed to recoup trapped hydrogen isotopes. Extraction technologies for use with a LiMIT-style loop system will be discussed and results will be presented. [Preview Abstract] |
Tuesday, October 24, 2017 4:36PM - 4:48PM |
JO4.00014: HIDRA-MAT: A Material Analysis Tool for Fusion Devices Daniel Andruczyk, Rabel Rizkallah, Felipe Bedoya, Aveek Kapat, Hanna Schamis, Jean Paul Allain The former WEGA stellarator which is now operating as HIDRA at the University of Illinois will be almost exclusively used to study the intimate relationship between the plasma interacting with surfaces of different materials. A Material Analysis Tool (HIDRA-MAT) is being designed and will be built based on the successful Material Analysis and Particle Probe (MAPP) which is currently used on NSTX-U at PPPL. This will be an \textit{in-situ} material diagnostic probe, meaning that all analysis can be done without breaking vacuum. This allows surface changes to be studied in real-time. HIDRA-MAT will consist of several \textit{in-situ} diagnostics including Langmuir probes (LP), Thermal Desorption Spectroscopy (TDS), X-ray Photo Spectroscopy (XPS) and Ion Scattering Spectroscopy (ISS). This presentation will outline the HIDRA-MAT diagnostic and initial design, as well as its integration into the HIDRA system. [Preview Abstract] |
Tuesday, October 24, 2017 4:48PM - 5:00PM |
JO4.00015: Behavior of axisymmetric density fluctuations in TCV Gabriele Merlo, Frank Jenko, Stephan Brunner, Stefano Coda, Zhouji Huang, Laurent Villard, Tobias Goerler, Alejandro B. Navarro, Daniel Told Axisymmetric density fluctuations, either with a radially coherent or dispersive nature, are routinely observed in the TCV tokamak and experimentally interpreted as Geodesic Acoustic Modes (GAMs). We use local and global GENE simulations to investigate their behavior. With a simplified physical model, neglecting impurities and using heavy electrons, simulations reproduce the observed behavior. Simulations allow to conclude that the modification of the safety factor q alone cannot explain the transition between these two different fluctuation regimes, which thus appear as a consequence of variations of other parameters, including collisionality and finite machine size effects. The behavior of the radially coherent GAM is further investigated with high-realism GENE simulations. With this set-up, local simulations reproduce the experimental transport level at different radii while matching the observed GAM frequency at the location where the mode peaks. Global high-realism runs, aiming at reproducing the radial extent of the fluctuations, will be discussed as well. [Preview Abstract] |
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